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Sample records for uranium-graphite critical facility

  1. Criticality calculations for a critical assembly, graphite moderate, using 20% enriched uranium

    International Nuclear Information System (INIS)

    Almeida Ferreira, A.C. de; Hukai, R.Y.

    1975-01-01

    The construction of a Zero Power Reactor (ZPR) at the Instituto de Energia Atomica in order to measure the neutron characteristics (parameters) of HTGR reactors is proposed. The necessary quantity fissile uranium for these measurements has been calculed. Criticality studies of graphite moderated critical assemblies containing thorium have been made and the critical mass of each of several typical commercial HTGR compositions has been calculated using computer codes HAMMER and CITATION. Assemblies investigated contained a central cylindrical core region, simulating a typical commercial HTGR composition, a uranium-graphite driver region and a outer pure graphite reflector region. It is concluded that a 10Kg inventory of fissile uranium will be required for a program of measurements utilizing each of the several calculated assemblies

  2. Integral measurements of lattice properties in the natural uranium-graphite critical facility Marius; Mesures globales de reseaux a graphite dans l'empilement critique marius

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    A systematic study of natural uranium-graphite lattices has been undertaken in the critical facility MARIUS, which was built in 1959 in Marcoule. Integral measurement of lattice properties are carried out by the progressive replacement method. This report describes the experimental methods, the analysis of the experiments and the results obtained for lattices with pitches ranging from 192 to 317 mm and fuel elements with cross sections ranging from 6 to 20 cm{sup 2}. The principles of correlation of the results are also outlined. Additional experimental results are also given, pertaining to the determination of the anisotropy, of both the axial and the radial migration areas, and of the age in graphite. (author) [French] L'empilement critique MARIUS, construit en 1959 a Marcoule, a ete utilise pour l'etude systematique des reseaux a graphite-uranium naturel. Les mesures globales de reseaux sont faites par la methode de remplacement progressif. On decrit ici les methodes experimentales utilisees pour ces mesures globales, les principes du depouillement et les resultats obtenus pour des pas de 192 a 317 mm et des combustibles de 6 a 20 cm{sup 2} d'uranium naturel. On donne d'autre part le principe de correlation des mesures. Un certain nombre de resultats experimentaux complementaires sont donnes, en permettant de determiner l'anisotropie, les aires de migration axiale et radiale, l'age dans le graphite. (auteur)

  3. Critical experiments on enriched uranium graphite moderated cores

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko; Akino, Fujiyoshi; Kitadate, Kenji; Kurokawa, Ryosuke

    1978-07-01

    A variety of 20 % enriched uranium loaded and graphite-moderated cores consisting of the different lattice cells in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments systematically. In the present report, the experimental results for homogeneously or heterogeneously fuel loaded cores and for simulation core of the experimental reactor for a multi-purpose high temperature reactor are filed so as to be utilized for evaluating the accuracy of core design calculation for the experimental reactor. The filed experimental data are composed of critical masses of uranium, kinetic parameters, reactivity worths of the experimental control rods and power distributions in the cores with those rods. Theoretical analyses are made for the experimental data by adopting a simple ''homogenized cylindrical core model'' using the nuclear data of ENDF/B-III, which treats the neutron behaviour after smearing the lattice cell structure. It is made clear from a comparison between the measurement and the calculation that the group constants and fundamental methods of calculations, based on this theoretical model, are valid for the homogeneously fuel loaded cores, but not for both of the heterogeneously fuel loaded cores and the core for simulation of the experimental reactor. Then, it is pointed out that consideration to semi-homogeneous property of the lattice cells for reactor neutrons is essential for high temperature graphite-moderated reactors using dispersion fuel elements of graphite and uranium. (author)

  4. SRTC criticality technical review: Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility

    International Nuclear Information System (INIS)

    Rathbun, R.

    1993-01-01

    Separate review of NMP-NCS-930058, open-quotes Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility (U), August 17, 1993,close quotes was requested of SRTC Applied Physics Group. The NCSE is a criticality assessment to determine waste container uranium limits in the Uranium Solidification Facility's Waste Handling Facility. The NCSE under review concludes that the NDA room remains in a critically safe configuration for all normal and single credible abnormal conditions. The ability to make this conclusion is highly dependent on array limitation and inclusion of physical barriers between 2x2x1 arrays of boxes containing materials contaminated with uranium. After a thorough review of the NCSE and independent calculations, this reviewer agrees with that conclusion

  5. Calculation of the fissile mass of a graphite moderated critical assembly using 93% enriched uranium

    International Nuclear Information System (INIS)

    Correa, F.; Marzo, M.A.S.; Collussi, I.; Ferreira, A.C.A.

    1976-01-01

    The critical mass of uranium has been calculated for a graphite moderated set fueled with 93% enriched uranium to be mounted on the Instituto de Energia Atomica split table Zero Power Reactor. The core composition was optimized to permit the maximum number of configurations to be studied. Analysis of three core compositions shows that 8 Kg of uranium enriched to 93% - U-235 (by weight) and 100 Kg of thorium would be sufficient for criticality experiments [pt

  6. Exponential experiment on a uranium-graphite lattice; Experience exponentielle sur un reseau uranium-graphite

    Energy Technology Data Exchange (ETDEWEB)

    Leroy, J; Martelly, J

    1958-12-01

    An exponential experiment on a natural uranium and graphite lattice is described. The critical buckling for a cubic pile made with this lattice is B{sup 2} = 0.726 {+-} 0.011 m{sup -2} and the anisotropy {alpha} = (B{sub L}{sup 2}/B{sub T}{sup 2}) is 0.987 {+-} 0.006. The behavior of the neutron density in the lattice near the reflector and sources is discussed in detail. (authors) [French] Une experience exponentielle sur reseau a uranium naturel et graphite est decrite. Le Laplacien de la pile cubique nue, critique, constituee avec un tel reseau serait: B{sup 2} = 0.726 {+-} 0.011 m{sup -2} et l'anisotropie {alpha} (B{sub L}{sup 2}/B{sub T}{sup 2}) est egale a 0.987 {+-} 0.006. La perturbation apportee a la densite de neutrons dans le reseau par le voisinage du reflecteur est discutee en detail. (auteurs)

  7. Channel uranium-graphite reactor mounting

    International Nuclear Information System (INIS)

    Polushkin, K.K.; Kuznetsov, A.G.; Zheleznyakov, B.N.

    1981-01-01

    According to theoretical principles of general engineering technology the engineering experience of construction-mounting works at the NPP with channel uranium-graphite reactors is systematized. Main parameters and structural features of the 1000 MW channel uranium-graphite reactors are considered. The succession of mounting operations, premounting equipment and pipelines preparation and mounting works technique are described. The most efficient methods of fitting, welding and machining of reactor elements are recommended. Main problems of technical control service are discussed. A typical netted diagram of main equipment of channel uranium-graphite reactors mounting is given

  8. The potential for criticality following disposal of uranium at low-level waste facilities: Uranium blended with soil

    Energy Technology Data Exchange (ETDEWEB)

    Toran, L.E.; Hopper, C.M.; Naney, M.T. [and others

    1997-06-01

    The purpose of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) facilities can be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of experts in hydrology, geology, geochemistry, soil chemistry, and criticality safety was formed to develop achievable scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM), and to use these scenarios to aid in evaluating the potential for nuclear criticality. The team`s approach was to perform simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some achievable scenarios for uranium migration and concentration increase at LLW disposal facilities, (2) model groundwater transport and subsequent concentration increase via sorption or precipitation of uranium, and (3) evaluate the potential for nuclear criticality resulting from potential increases in uranium concentration over disposal limits. The analysis of SNM was restricted to {sup 235}U in the present scope of work. The outcome of the work indicates that criticality is possible given established regulatory limits on SNM disposal. However, a review based on actual disposal records of an existing site operation indicates that the potential for criticality is not a concern under current burial practices.

  9. The potential for criticality following disposal of uranium at low-level waste facilities: Uranium blended with soil

    International Nuclear Information System (INIS)

    Toran, L.E.; Hopper, C.M.; Naney, M.T.

    1997-06-01

    The purpose of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) facilities can be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of experts in hydrology, geology, geochemistry, soil chemistry, and criticality safety was formed to develop achievable scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM), and to use these scenarios to aid in evaluating the potential for nuclear criticality. The team's approach was to perform simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some achievable scenarios for uranium migration and concentration increase at LLW disposal facilities, (2) model groundwater transport and subsequent concentration increase via sorption or precipitation of uranium, and (3) evaluate the potential for nuclear criticality resulting from potential increases in uranium concentration over disposal limits. The analysis of SNM was restricted to 235 U in the present scope of work. The outcome of the work indicates that criticality is possible given established regulatory limits on SNM disposal. However, a review based on actual disposal records of an existing site operation indicates that the potential for criticality is not a concern under current burial practices

  10. The Potential for Criticality Following Disposal of Uranium at Low-Level-Waste Facilities. Containerized Disposal

    International Nuclear Information System (INIS)

    Colten-Bradley, V.A.; Hopper, C.M.; Parks, C.V.; Toran, L.E.

    1999-01-01

    The purpose of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) facilities can be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of experts in hydrology, geology, geochemistry, soil chemistry, and criticality safety was formed to develop and test some reasonable scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM) and to use these scenarios to aid in evaluating the potential for nuclear criticality. The team's approach was to perform simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some possible scenarios for uranium migration and concentration increase at LLW disposal facilities, (2) model groundwater transport and subsequent concentration increase via precipitation of uranium, and (3) evaluate the potential for nuclear criticality resulting from potential increase in uranium concentration over disposal limits. The analysis of SNM was restricted to 235 U in the present scope of work. The work documented in this report indicates that the potential for a criticality safety concern to arise in an LLW facility is extremely remote, but not impossible. Theoretically, conditions that lead to a potential criticality safety concern might arise. However, study of the hydrogeochemical mechanisms, the associated time frames, and the factors required for an actual criticality event indicate that proper emplacement of the SNM at the site can eliminate practical concerns relative to the occurrence and possible consequences of a criticality event

  11. Uranium Oxide Aerosol Transport in Porous Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, Jeremy; Gerlach, David C.; Scheele, Randall D.; Stewart, Mark L.; Reid, Bruce D.; Gauglitz, Phillip A.; Bagaasen, Larry M.; Brown, Charles C.; Iovin, Cristian; Delegard, Calvin H.; Zelenyuk, Alla; Buck, Edgar C.; Riley, Brian J.; Burns, Carolyn A.

    2012-01-23

    The objective of this paper is to investigate the transport of uranium oxide particles that may be present in carbon dioxide (CO2) gas coolant, into the graphite blocks of gas-cooled, graphite moderated reactors. The transport of uranium oxide in the coolant system, and subsequent deposition of this material in the graphite, of such reactors is of interest because it has the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). The GIRM is a technology that has been developed to validate the declared operation of graphite moderated reactors. GIRM exploits isotopic ratio changes that occur in the impurity elements present in the graphite to infer cumulative exposure and hence the reactor’s lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.

  12. Long-term criticality control in radioactive waste disposal facilities using depleted uranium

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1997-01-01

    Plant photosynthesis has created a unique planetary-wide geochemistry - an oxidizing atmosphere with oxidizing surface waters on a planetary body with chemically reducing conditions near or at some distance below the surface. Uranium is four orders of magnitude more soluble under chemically oxidizing conditions than it is under chemically reducing conditions. Thus, uranium tends to leach from surface rock and disposal sites, move with groundwater, and concentrate where chemically reducing conditions appear. Earth's geochemistry concentrates uranium and can separate uranium from all other elements except oxygen, hydrogen (in water), and silicon (silicates, etc). Fissile isotopes include 235 U, 233 U, and many higher actinides that eventually decay to one of these two uranium isotopes. The potential for nuclear criticality exists if the precipitated uranium from disposal sites has a significant fissile enrichment, mass, and volume. The earth's geochemistry suggests that isotopic dilution of fissile materials in waste with 238 U is a preferred strategy to prevent long-term nuclear criticality in and beyond the boundaries of waste disposal facilities because the 238 U does not separate from the fissile uranium isotopes. Geological, laboratory, and theoretical data indicate that the potential for nuclear criticality can be minimized by diluting fissile materials with- 238 U to 1 wt % 235 U equivalent

  13. Nuclear criticality safety assessment of the Consolidated Edison Uranium-Solidification Program Facility

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1984-01-01

    A nuclear criticality assessment of the Consolidated Edison Uranium-Solidification Program facility confirms that all operations involved in the process may be conducted with an acceptable margin of subcriticality. Normal operation presents no concern since subcriticality is maintained by design. Several recommendations are presented to prevent, or mitigate the consequences of, any abnormal events that might occur in the various portions of the process. These measures would also serve to reduce to a minimum the administrative controls required to prevent criticality

  14. Experience of on-site disposal of production uranium-graphite nuclear reactor.

    Science.gov (United States)

    Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

    2018-04-01

    The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

  15. Construction of new critical experiment facilities in JAERI

    International Nuclear Information System (INIS)

    Takeshita, Isao; Itahashi, Takayuki; Ogawa, Kazuhiko; Tonoike, Kotaro; Matsumura, Tatsuro; Miyoshi, Yoshinori; Nakajima, Ken; Izawa, Naoki

    1995-01-01

    Japan Atomic Energy Research Institute (JAERI) has promoted the experiment research program on criticality safety since early in 1980s and two types of new critical facilities, Static Experiment Critical Facility (STACY) and Transient Experiment Critical Facility (TRACY) were completed on 1994 in Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF) of JAERI Tokai Research Establishment. STACY was designed so as to obtain critical mass data of low enriched uranium and plutonium solution which is extensively handled in LWR fuel reprocessing plant. TRACY is the critical facility where critical accident phenomenon is demonstrated with low enriched uranium nitrate solution. For criticality safety experiments with both facilities, the Fuel Treatment System is attached to them, where composition and concentration of uranium and plutonium nitrate solutions are widely varied so as to obtain experiments data covering fuel solution conditions in reprocessing plant. Design performances of both critical facilities were confirmed through mock-up tests of important components and cold function tests. Hot function test has started since January of 1995 and some of the results on STACY are to be reported. (author)

  16. Construction of STACY (Static Experiment Critical Facility)

    International Nuclear Information System (INIS)

    Murakami, Kiyonobu; Onodera, Seiji; Hirose, Hideyuki

    1998-08-01

    Two critical assemblies, STACY (Static Experiment Critical Facility) and TRACY (Transient Experiment Critical Facility), were constructed in NUCEF (Nuclear Fuel Cycle Safety Engineering Research Facility) to promote researches on the criticality safety at a reprocessing facility. STACY aims at providing critical data of uranium nitrate solution, plutonium nitrate solution and their mixture while varying concentration of solution fuel, core tank shape and size and neutron reflecting condition. STACY achieved first criticality in February 1995, and passed the licensing inspection by STA (Science and Technology Agency of Japan) in May. After that a series of critical experiments commenced with 10 w/o enriched uranium solution. This report describes the outline of STACY at the end of FY 1996. (author)

  17. Uranium and thorium abundances in some graphite-bearing precambrian rocks of India and implications

    International Nuclear Information System (INIS)

    Pandey, U.K.; Krishnamurthy, P.

    1995-01-01

    Graphite schists from parts of Gujarat in the Aravalli supergroup show maximum contents of uranium (70-95 ppm), hosted mainly in the graphites, whereas such schists from the Tamil Nadu granulite terrain contain distinctly lower amounts of uranium (7-9 ppm). Graphite-bearing hornblende gneiss and calc-granulites from Madurai, Tamil Nadu, contain higher amounts of uranium (12-28 ppm) than the schists, and uranium is mainly hosted by the magnetite and allanite occurring as independent grains with flaky graphite and also as inclusions within quartz. Khondalites from Andhra Pradesh are depleted in uranium (0.9-1.3 ppm) compared to Th (17.5-20.2 ppm). Except for the khondalites, which have high Th/U ratio (13.5-22.4), all the other samples have very low Th/U ratios (0.10-0.80) compared to the crustal average (3-4). Such variations among similar rock types, may in part be related to uranium and thorium abundances inherited from parental rocks, modified later by hydrothermal and/or metasomatic processes. Graphites from such rock types can provide both in situ and migrant reductants for hosting a variety of uranium and other metallic deposits. (author). 12 refs., 1 tab., 1 fig

  18. SRTC criticality safety technical review: Nuclear criticality safety evaluation 94-02, uranium solidification facility pencil tank module spacing

    International Nuclear Information System (INIS)

    Rathbun, R.

    1994-01-01

    Review of NMP-NCS-94-0087, ''Nuclear Criticality Safety Evaluation 94-02: Uranium Solidification Facility Pencil Tank Module Spacing (U), April 18, 1994,'' was requested of the SRTC Applied Physics Group. The NCSE is a criticality assessment to show that the USF process module spacing, as given in Non-Conformance Report SHM-0045, remains safe for operation. The NCSE under review concludes that the module spacing as given in Non-Conformance Report SHM-0045 remains in a critically safe configuration for all normal and single credible abnormal conditions. After a thorough review of the NCSE, this reviewer agrees with that conclusion

  19. Chapter 3: Exponential experiments on graphite-moderated lattices fuelled with near-natural uranium metal rods

    International Nuclear Information System (INIS)

    McCulloch, D.B.; Clarke, W.G.; Ashworth, F.P.O.; Hoskins, T.A.

    1963-01-01

    Exponential experiments have been carried out on graphite lattices fuelled by 1.2 in. diameter uranium metal rods at three near-natural U 235 compositions, 0.6 Co, 1.3 Co and 1.6 Co. The results, together with those already existing from earlier exponential or critical measurements on these and similar natural uranium rods, have been correlated with the theory of Syrett (1961) and also with the modified form of this theory given in Vol.1, Ch. 7. (author)

  20. Experiments with HEU (93.14 wt.%) metal annuli with internal graphite cylinder

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiaobo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wehmann, Udo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mihalczo, John T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, only three experimental configurations are described here. They are internal graphite reflected metal uranium assemblies with three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches). These experiments can be found in Reference 1 and in their associated logbook

  1. STACY and TRACY: nuclear criticality experimental facilities under construction

    International Nuclear Information System (INIS)

    Kobayashi, I.; Takeshita, I.; Yanagisawa, H.; Tsujino, T.

    1992-01-01

    Japan Atomic Energy Research Institute is constructing a Nuclear Fuel Cycle Safety Engineering Research Facility, NUCEF, where the following research themes essential for evaluating safety problems relating to back-end technology in nuclear fuel cycle facilities will be studied: nuclear criticality safety research; research on advanced reprocessing processes and partitioning; and research on transuranic waste treatment and disposal. To perform nuclear criticality safety research related to the reprocessing of light water reactor spent fuels, two criticality experimental facilities, STACY and TRACY, are under construction. STACY (Static Criticality Facility) will be used for the study of criticality conditions of solution fuels, uranium, plutonium and their mixtures. TRACY (Transient Criticality Facility) will be used to investigate criticality accident phenomena with uranium solutions. The construction progress and experimental programmes are described in this Paper. (author)

  2. Fabrication of uranium carbide/beryllium carbide/graphite experimental-fuel-element specimens

    International Nuclear Information System (INIS)

    Muenzer, W.A.

    1978-01-01

    A method has been developed for fabricating uranium carbide/beryllium carbide/graphite fuel-element specimens for reactor-core-meltdown studies. The method involves milling and blending the raw materials and densifying the resulting blend by conventional graphite-die hot-pressing techniques. It can be used to fabricate specimens with good physical integrity and material dispersion, with densities of greater than 90% of the theoretical density, and with a uranium carbide particle size of less than 10 μm

  3. Features of spherical uranium-graphite HTGR fuel elements control

    International Nuclear Information System (INIS)

    Kreindlin, I.I.; Oleynikov, P.P.; Shtan, A.S.

    1985-01-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described

  4. Features of spherical uranium-graphite HTGR fuel elements control

    Energy Technology Data Exchange (ETDEWEB)

    Kreindlin, I I; Oleynikov, P P; Shtan, A S

    1985-07-01

    Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described.

  5. Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

    Science.gov (United States)

    Pavliuk, A. O.; Zagumennov, V. S.; Kotlyarevskiy, S. G.; Bespala, E. V.

    2018-01-01

    The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.

  6. Critical experiment study on uranyl nitrate solution experiment facility

    International Nuclear Information System (INIS)

    Zhu Qingfu; Shi Yongqian; Wang Jinrong

    2005-01-01

    The Uranyl Nitrate Solution Experiment Facility was constructed for the research on nuclear criticality safety. In this paper, the configuration of the facility is introduced; a series of critical experiments on uranyl nitrate solution is described later, which were performed for various uranium concentrations under different conditions, i.e. with or without neutron absorbers in the core and with or without water-reflector outside the core. Critical volume and the minimum 235U critical mass for different uranium concentrations are presented. Finally, theoretical analysis is made on the experimental results. (authors)

  7. Isotopic composition of uranium in U3O8 by neutron induced reactions utilizing thermal neutrons from critical facility and high resolution gamma-ray spectrometry

    International Nuclear Information System (INIS)

    Acharya, R.; Pujari, P.K.; Goel, Lokesh

    2015-01-01

    Uranium in oxide and metal forms is used as fuel material in nuclear power reactors. For chemical quality control, it is necessary to know the isotopic composition (IC) of uranium i.e., 235 U to 238 U atom ratio as well as 235 U atom % in addition to its total concentration. Uranium samples can be directly assayed by passive gamma ray spectrometry for obtaining IC by utilizing 185 keV (γ-ray abundance 57.2%) of 235 U and 1001 keV (γ-ray abundance 0.837%) of 234m Pa (decay product of 238 U). However, due to low abundance of 1001 keV, often it is not practiced to obtain IC by this method as it gives higher uncertainty even if higher mass of sample and counting time are used. IC of uranium can be determined using activity ratio of neutron induced fission product of 235 U to activation product of 238 U ( 239 Np). In the present work, authors have demonstrated methodologies for determination of IC of U as well as 235 U atom% in natural ( 235 U 0.715%) and low enriched uranium (LEU, 3-20 atom % of 235 U) samples of uranium oxide (U 3 O 8 ) by utilizing ratio of counts at 185 keV γ-ray or γ-rays of fission products with respect to 277 keV of 239 Np. Natural and enriched samples (about 25 mg) were neutron irradiated for 4 hours in graphite reflector position of AHWR Critical Facility (CF) using highly thermalized (>99.9% thermal component) neutron flux (∼10 7 cm -2 s -1 )

  8. Criticality safety studies for plutonium–uranium metal fuel pin fabrication facility

    International Nuclear Information System (INIS)

    Stephen, Neethu Hanna; Reddy, C.P.

    2013-01-01

    Highlights: ► Criticality safety limits for PUMP-F facility is identified. ► The fissile mass which can be handled safely during alloy preparation is 10.5 kg. ► The number of fuel slugs which can be handled safely during injection casting is 53. ► The number of fuel slugs which can be handled safely after fuel fabrication is 71. - Abstract: This study focuses on the criticality safety during the fabrication of fast reactor metal fuel pins comprising of the fuel type U–15Pu, U–19Pu and U–19Pu–6Zr in the Plutonium–Uranium Metal fuel Pin fabrication Facility (PUMP-F). Maximum amount of fissile mass which can be handled safely during master alloy preparation, Injection casting and fuel slug preparation following fuel pin fabrication were identified and fixed based on this study. In the induction melting furnace, the fissile mass can be limited to 10.5 kg. During fuel slug preparation and fuel pin fabrication, fuel slugs and pins were arranged in hexagonal and square lattices to identify the most reactive configuration. The number of fuel slugs which can be handled safely after injection casting can be fixed to be 53, whereas after fuel fabrication it is 71

  9. Criticality analysis in uranium enrichment plant

    International Nuclear Information System (INIS)

    Okamoto, Tsuyoshi; Kiyose, Ryohei

    1977-01-01

    In a large scale uranium enrichment plant, uranium inventory in cascade rooms is not very large in quantity, but the facilities dealing with the largest quantity of uranium in that process are the UF 6 gas supply system and the blending system for controlling the product concentration. When UF 6 spills out of these systems, the enriched uranium is accumulated, and the danger of criticality accident is feared. If a NaF trap is placed at the forestage of waste gas treatment system, plenty of UF 6 and HF are adsorbed together in the NaF trap. Thus, here is the necessity of checking the safety against criticality. Various assumptions were made to perform the computation surveying the criticality of the system composed of UF 6 and HF adsorbed on NaF traps with WIMS code (transport analysis). The minimum critical radius resulted in about 53 cm in case of 3.5% enriched fuel for light water reactors. The optimum volume ratio of fissile material in the double salt UF 6 .2NaF and NaF.HF is about 40 vol. %. While, criticality survey computation was also made for the annular NaF trap having the central cooling tube, and it was found that the effect of cooling tube radius did not decrease the multiplication factor up to the cooling tube radius of about 5 cm. (Wakatsuki, Y.)

  10. Very high temperature gas-cooled reactor critical facility for Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Ishihara, Noriyuki

    1985-01-01

    The outline of the critical facility, its construction, the results of the basic studies and experiments on the graphite material, and the results obtained from the test conducted on the overall functions of the critical facility were reported. With the completion of the critical facility, it has been made possible to demonstrate the establishment of the manufacturing techniques and product-quality guarantee for extremely pure isotropic graphite in addition to the reliability of the structural design and analytical techniques for the main unit of the critical facility. It is expected that the present facility will prove instrumental in the verification of the nuclear safety of the very high temperature gas-cooled nuclear reactor and in the acquisition of experimental data on the reactor physics pertaining to the improvement of the reactor characteristics. The tasks which remain to be accomplished hereafter are the improvements of the performance and quality features with regard to the oxidization of graphite, the heat-resisting structural materials, and the welded structures. (Kubozono, M.)

  11. Developments in natural uranium - graphite reactors; Developpement des reacteurs a graphite et uranium naturel

    Energy Technology Data Exchange (ETDEWEB)

    Bourgeois, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Saitcevsky, B [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

  12. Comparative analysis of calculations and experiment for uranium-graphite lattices with natural and slightly-enriched uranium

    International Nuclear Information System (INIS)

    Khrennikov, N.N.; Shchukin, A.V.

    1988-01-01

    Three sets of experiments carried out at different times and in different laboratories on measuring the material parameter for uranium-graphite lattices using natural and slightly enriched uranium are analyzed. Comparison with the calculations by the TRIFOGR and MCU (the Monte Carlo method) codes reveals resonable agreement between the calculation and experiment (of the order of 0.4% in K eff ). 17 refs.; 3 tabs

  13. Developments in natural uranium - graphite reactors; Developpement des reacteurs a graphite et uranium naturel

    Energy Technology Data Exchange (ETDEWEB)

    Bourgeois, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Saitcevsky, B. [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

  14. Experimental performance and results of the critical pebble bed facility KAHTER

    Energy Technology Data Exchange (ETDEWEB)

    Krings, F. J.; Drueke, V.; Kirch, N.; Neef, R. D.

    1974-10-15

    The paper provides a description and results of critical experiments performed in KAHTER fueled with pebbles containing coated particles of HEU/Th oxide with a ratio of uranium-to-thorium of 1.1:5. Core loadings with varying amounts of fuel and solid graphite pebbles were tested with fuel-to-graphite pebble ratios of 3:1, 1:1, and 1:3. Tests included criticality for various fuel loadings with all control rods removed, control rod worths for reflector-mounted control as single rods and in a bank and control worths for a central control rod, reaction rates by flux wire activations (Dy, Mn, In, Au, and U-235) and detector measurements (BF3 and fission chamber), simulated xenon stability testing using the motions of a Cf-252 source and Cd-absorber observed by an externally-located BF3 detector, and the reactivity worth of a Hf burnable absorber. For calculations of the room-temperature zero-power critical experiment, the values for nitrogen and hydrogen contents of the graphite were taken from previous experiments in CESAR.

  15. LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

    International Nuclear Information System (INIS)

    Brogli, R.; Bucher, K.H.; Chawla, R.; Foskolos, K.; Luchsinger, H.; Mathews, D.; Sarlos, G.; Seiler, R.

    1990-01-01

    New critical experiments in the framework of an IAEA Coordinated Research Program on 'Validation of Safety Related Reactor Physics Calculations for Low Enriched HTRs' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. (author)

  16. LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

    Energy Technology Data Exchange (ETDEWEB)

    Brogli, R; Bucher, K H; Chawla, R; Foskolos, K; Luchsinger, H; Mathews, D; Sarlos, G; Seiler, R [Paul Scherrer Institute, Laboratory for Reactor Physics and System Technology Wuerenlingen and Villigen, Villigen PSI (Switzerland)

    1990-07-01

    New critical experiments in the framework of an IAEA Coordinated Research Program on 'Validation of Safety Related Reactor Physics Calculations for Low Enriched HTRs' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. (author)

  17. Diffusion of uranium in H-451 graphite at 900 to 14000C

    International Nuclear Information System (INIS)

    Tallent, O.K.; Wichner, R.P.; Towns, R.L.

    1983-03-01

    In this study, the diffusion of uranium (as a stand-in for plutonium) was investigated under conditions approximating those of the primary coolant loop in a High Temperature Gas-Cooled Reactor (HTGR). Profiles were obtained for uranium penetration in H-451 graphite (from the Great Lakes Carbon Company) at temperatures ranging from 900 to 1400 0 C. Diffusion coefficients are established for UO 2 and UC 2

  18. Criticality management of Tokai reprocessing facility

    Energy Technology Data Exchange (ETDEWEB)

    Nojiri, Ichiro [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2001-01-01

    In fuel cycle centers a number of equipment and vessels of various types and of complex design are used in several processes, i.e. dissolution of spent fuels, separation and storage of uranium and plutonium from fission products and transuranium elements. For each processes, Monte Carlo codes are frequently applied to manage the fuel criticality. Safety design depends largely on specific features of each facilities. The present report describes status of criticality management for main processes in Tokai Reprocessing Facility, JNC, and the criticality conditions specifically existing there. The guiding principle throughout consists of mass control, volume control, design (form) control, concentration control, and control due to employment of neutron poisons. (S. Ohno)

  19. COGEMA's UMF [Uranium Management Facility

    International Nuclear Information System (INIS)

    Lamorlette, G.; Bertrand, J.P.

    1988-01-01

    The French government-owned corporation, COGEMA, is responsible for the nuclear fuel cycle. This paper describes the activities at COGEMA's Pierrelatte facility, especially its Uranium Management Facility. UF6 handling and storage is described for natural, enriched, depleted, and reprocessed uranium. UF6 quality control specifications, sampling, and analysis (halocarbon and volatile fluorides, isotopic analysis, uranium assay, and impurities) are described. In addition, the paper discusses the filling and cleaning of containers and security at UMF

  20. Developments in natural uranium - graphite reactors

    International Nuclear Information System (INIS)

    Bourgeois, J.

    1964-01-01

    The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

  1. Fine structure and spectral index measurements in natural uranium - graphite lattices; Mesures fines dans des reseaux a graphite

    Energy Technology Data Exchange (ETDEWEB)

    Cogne, F; Journet, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The experiments described in this report have been carried out for the most part in the critical facility MARIUS, and a few during the start up of the EDF-1 power reactor. The first part deals with the fine structure measurements made in various lattices and with their analysis. Integration over the neutron spectrum of the mono-kinetic disadvantage factor derived by the A.B.H method yields results in good agreement with the experiments. The second part deals with spectral indexes measurements (Pu/U, In/Mn) made at room temperature in MARIUS. Comparison are made of experiments with calculations using various thermalization models. Experiments carried out at higher temperatures in EDF-1 are also described. (authors) [French] Les mesures decrites dans ce rapport ont ete faites pour la plupart dans l'empilement critique MARIUS sur des reseaux a graphite-uranium naturel. Une premiere partie traite des mesures de structure fine faites dans differents reseaux et de leur interpretation. On montre en particulier qu'une integration sur le spectre d'un calcul monocinetique type A.B.H. rend bien compte des experiences. Dans une deuxieme partie, on donne les resultats de mesures d'indices de spectre Pu/U et In/Mn faites sur des reseaux froids a MARIUS et leur comparaison avec les differents modeles de calculs de thermalisation. On donne egalement les resultats de quelques mesures en temperature effectuees lors du demarrage du reacteur EDF-1. (auteurs)

  2. Study of graphite reactivity worth on well-defined cores assembled on LR-0 reactor

    International Nuclear Information System (INIS)

    Košťál, Michal; Rypar, Vojtěch; Milčák, Ján; Juříček, Vlastimil; Losa, Evžen; Forget, Benoit; Harper, Sterling

    2016-01-01

    Highlights: • A light water critical facility for graphite reactivity worth measurements. • Comparison of calculated and measured k eff . • Effect of graphite description on k eff . - Abstract: Graphite is an often-used moderating material on the basis of its good moderating power and very low absorption cross section. This small absorption cross section permits the use of natural or low-enriched uranium in graphite moderated reactors. Graphite is now being considered as the moderator for Fluoride-salt-cooled High Temperature Reactors (FHR). The critical moderator level was measured for various graphite block configurations in an experimental dry assembly of the LR-0 reactor. Comparisons with experiments were performed between Monte Carlo simulation tools for which satisfactory agreement was obtained with the exception of some systematic discrepancies. The larger discrepancies were observed when using the ENDF/B-VII.0 library. To decrease the uncertainties, based on conservative assumptions, relative comparisons were done. The results provided by the different nuclear data libraries are within 3 sigma interval of experimental uncertainties. It has been determined that differences between the results of calculations are caused by variations in the (n,n), (n,n′), (n,g) reactions and also by various angular distributions, while the (n,g) cross section variations play only a minor role for these configurations.

  3. Graphite Isotope Ratio Method Development Report: Irradiation Test Demonstration of Uranium as a Low Fluence Indicator

    International Nuclear Information System (INIS)

    Reid, B.D.; Gerlach, D.C.; Love, E.F.; McNeece, J.P.; Livingston, J.V.; Greenwood, L.R.; Petersen, S.L.; Morgan, W.C.

    1999-01-01

    This report describes an irradiation test designed to investigate the suitability of uranium as a graphite isotope ratio method (GIRM) low fluence indicator. GIRM is a demonstrated concept that gives a graphite-moderated reactor's lifetime production based on measuring changes in the isotopic ratio of elements known to exist in trace quantities within reactor-grade graphite. Appendix I of this report provides a tutorial on the GIRM concept

  4. Decommissioning of U.S. uranium production facilities

    Energy Technology Data Exchange (ETDEWEB)

    1995-02-01

    From 1980 to 1993, the domestic production of uranium declined from almost 44 million pounds U{sub 3}O{sub 8} to about 3 million pounds. This retrenchment of the U.S. uranium industry resulted in the permanent closing of many uranium-producing facilities. Current low uranium prices, excess world supply, and low expectations for future uranium demand indicate that it is unlikely existing plants will be reopened. Because of this situation, these facilities eventually will have to be decommissioned. The Uranium Mill Tailings and Radiation Control Act of 1978 (UMTRCA) vests the U.S. Environmental Protection Agency (EPA) with overall responsibility for establishing environmental standards for decommissioning of uranium production facilities. UMTRCA also gave the U.S. Nuclear Regulatory Commission (NRC) the responsibility for licensing and regulating uranium production and related activities, including decommissioning. Because there are many issues associated with decommissioning-environmental, political, and financial-this report will concentrate on the answers to three questions: (1) What is required? (2) How is the process implemented? (3) What are the costs? Regulatory control is exercised principally through the NRC licensing process. Before receiving a license to construct and operate an uranium producing facility, the applicant is required to present a decommissioning plan to the NRC. Once the plan is approved, the licensee must post a surety to guarantee that funds will be available to execute the plan and reclaim the site. This report by the Energy Information Administration (EIA) represents the most comprehensive study on this topic by analyzing data on 33 (out of 43) uranium production facilities located in Colorado, Nebraska, New Mexico, South Dakota, Texas, Utah, and Washington.

  5. Decommissioning of U.S. uranium production facilities

    International Nuclear Information System (INIS)

    1995-02-01

    From 1980 to 1993, the domestic production of uranium declined from almost 44 million pounds U 3 O 8 to about 3 million pounds. This retrenchment of the U.S. uranium industry resulted in the permanent closing of many uranium-producing facilities. Current low uranium prices, excess world supply, and low expectations for future uranium demand indicate that it is unlikely existing plants will be reopened. Because of this situation, these facilities eventually will have to be decommissioned. The Uranium Mill Tailings and Radiation Control Act of 1978 (UMTRCA) vests the U.S. Environmental Protection Agency (EPA) with overall responsibility for establishing environmental standards for decommissioning of uranium production facilities. UMTRCA also gave the U.S. Nuclear Regulatory Commission (NRC) the responsibility for licensing and regulating uranium production and related activities, including decommissioning. Because there are many issues associated with decommissioning-environmental, political, and financial-this report will concentrate on the answers to three questions: (1) What is required? (2) How is the process implemented? (3) What are the costs? Regulatory control is exercised principally through the NRC licensing process. Before receiving a license to construct and operate an uranium producing facility, the applicant is required to present a decommissioning plan to the NRC. Once the plan is approved, the licensee must post a surety to guarantee that funds will be available to execute the plan and reclaim the site. This report by the Energy Information Administration (EIA) represents the most comprehensive study on this topic by analyzing data on 33 (out of 43) uranium production facilities located in Colorado, Nebraska, New Mexico, South Dakota, Texas, Utah, and Washington

  6. Graphite reactor physics; Physique des piles a graphite

    Energy Technology Data Exchange (ETDEWEB)

    Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Noc, B [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The study of graphite-natural uranium power reactor physics, undertaken ten years ago when the Marcoule piles were built, has continued to keep in step with the development of this type of pile. From 1960 onwards the critical facility Marius has been available for a systematic study of the properties of lattices as a function of their pitch, of fuel geometry and of the diameter of cooling channels. This study has covered a very wide field: lattice pitch varying from 19 to 38 cm. uranium rods and tubes of cross-sections from 6 to 35 cm{sup 2}, channels with diameters between 70 and 140 mm. The lattice calculation methods could thus be checked and where necessary adapted. The running of the Marcoule piles and the experiments carried out on them during the last few years have supplied valuable information on the overall evolution of the neutronic properties of the fuel as a function of irradiation. More detailed experiments have also been performed in Marius with plutonium-containing fuels (irradiated or synthetic fuels), and will be undertaken at the beginning of 1965 at high temperature in the critical facility Cesar, which is just being completed at Cadarache. Spent fuel analyses complement these results and help in their interpretation. The thermalization and spectra theories developed in France can thus be verified over the whole valid temperature range. The efficiency of control rods as a function of their dimensions, the materials of which they are made and the lattices surrounding them has been measured in Marius, and the results compared with calculation on the one hand and with the measurements carried out in EDF 1 on the other. Studies on the control proper of graphite piles were concerned essentially with the risks of spatial instability and the means of detecting and controlling them, and with flux distortions caused by the control rods. (authors) [French] Entreprise il y a dix ans a l'occasion de la construction des piles de Marcoule, l'etude de la

  7. Safeguarding uranium enrichment facilities. Review and analysis of the status of safeguards technology for uranium enrichment facilities

    International Nuclear Information System (INIS)

    1977-09-01

    The objective of this paper is to examine critically the diversion potential at uranium enrichment facilities and to outline a basic safeguards strategy which counters all identified hazards as completely as possible yet with a minimum of non-essential redundancy. Where existing technology does not appear to be adequate for effective safeguards, the limitations are examined, and suggestions for further R and D effort are made. Parts of this report are generally applicable to all currently known enrichment processes, while other parts are specifically directed toward facilities based on the gas centrifuge process. It is hoped that additional sections discussing a safeguards strategy for gas diffusion facilities can be added later. It should be emphasized that this is a technical report, and does not reflect any legal positions. The safeguards strategy and subsequent inspection procedures are intended as guidelines, not as negotiating positions

  8. Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France; Quelques aspects economiques de la filiere uranium naturel - Graphite - gaz. Etat actuel et tendance des couts en France

    Energy Technology Data Exchange (ETDEWEB)

    Gaussens, J; Tanguy, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Leo, B [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The first part of this report defines the economic advantages of natural uranium fuels, which are as follows: the restricted number and relatively simple fabrication processes of the fuel elements, the low cost per kWh of the finished product and the reasonable capital investments involved in this type of fuel cycle as compared to that of enriched uranium. All these factors combine to reduce the arbitrary nature of cost estimates, which is particularly marked in the case of enriched uranium due to the complexity of its cycle and the uncertainties of plutonium prices). Finally, the wide availability of yellowcake, as opposed to the present day virtual monopoly of isotope separation, and the low cost of natural uranium stockpiling, offer appreciable guarantees in the way of security of supply and economic and political independence as compared with the use of enriched uranium. As far as overall capital investments are concerned, it is shown that, although graphite-gas reactor costs are higher than those of light water reactors in certain capacity ranges, the situation becomes far less clear when we start taking into account, in the interest of national independence, the cost of nuclear fuel production equipment in the case of each of these types of reactor. Finally, the marginal cost of the power capacity of a graphite-gas reactor is low and its technological limitations have receded (owing particularly to the use of prestressed concrete). It is a well known fact that the trend is now towards larger power station units, which means that the rentability of natural uranium graphite reactors as compared to other types of reactors will become more and more pronounced. The second section aims at presenting a realistic short and medium term view of the fuel, running, and investment costs of French natural uranium graphite gas, reactors. Finally, the economic goals which this type of reactor can reach in the very near future are given. It is thus shown that considerable

  9. Use of Savannah River Site facilities for blend down of highly enriched uranium

    International Nuclear Information System (INIS)

    Bickford, W.E.; McKibben, J.M.

    1994-02-01

    Westinghouse Savannah River Company was asked to assess the use of existing Savannah River Site (SRS) facilities for the conversion of highly enriched uranium (HEU) to low enriched uranium (LEU). The purpose was to eliminate the weapons potential for such material. Blending HEU with existing supplies of depleted uranium (DU) would produce material with less than 5% U-235 content for use in commercial nuclear reactors. The request indicated that as much as 500 to 1,000 MT of HEU would be available for conversion over a 20-year period. Existing facilities at the SRS are capable of producing LEU in the form of uranium trioxide (UO 3 ) powder, uranyl nitrate [UO 2 (NO 3 ) 2 ] solution, or metal. Additional processing, and additional facilities, would be required to convert the LEU to uranium dioxide (UO 2 ) or uranium hexafluoride (UF 3 ), the normal inputs for commercial fuel fabrication. This study's scope does not include the cost for new conversion facilities. However, the low estimated cost per kilogram of blending HEU to LEU in SRS facilities indicates that even with fees for any additional conversion to UO 2 or UF 6 , blend-down would still provide a product significantly below the spot market price for LEU from traditional enrichment services. The body of the report develops a number of possible facility/process combinations for SRS. The primary conclusion of this study is that SRS has facilities available that are capable of satisfying the goals of a national program to blend HEU to below 5% U-235. This preliminary assessment concludes that several facility/process options appear cost-effective. Finally, SRS is a secure DOE site with all requisite security and safeguard programs, personnel skills, nuclear criticality safety controls, accountability programs, and supporting infrastructure to handle large quantities of special nuclear materials (SNM)

  10. Decommissioning: A critical component of the design for uranium tailings management facilities

    International Nuclear Information System (INIS)

    Clifton, W.A.; Barsi, R.G.; Misfeldt, G.A.

    2000-01-01

    Uranium was discovered in the Beaverlodge area of northern Saskatchewan in 1934 with the first major mill beginning operation in 1953. Little attention was paid to tailings quality or tailings management practices. With the onset of the modem uranium operations beginning in the late 1970's, it was repeatedly evident, that the public had significant concerns, particularly with respect to tailings management, that must be addressed if the developments were to be allowed to proceed. Primary considerations related to environmental protection, public safety and an assurance of the ongoing sustainable development of the region. Integrating the decommissioning of a mine/mill site into development planning from the very outset has proven to be a critical component that has contributed to the ongoing success of the Saskatchewan uranium operations. This paper will provide a case study of the evolution of the uranium tailings management technology utilized in Saskatchewan. It documents the evolution of tailings management processes and the characteristics of tailings produced by successive mines in northern Saskatchewan. It also discusses the evolution of technologies applied to management of uranium mill tailings and demonstrates how progressively increasing levels of environmental protection have been achieved during the last 47 years of uranium mill operation. The paper also shows that the planned and progressive decommissioning of an operational site is the key to: Minimizing environmental impacts; Satisfying public and regulatory concerns; Minimizing operational and decommissioning costs; Minimizing corporate liability; and Shifting public resistance to public support. (author)

  11. Decommissioning: A critical component of the design for uranium tailings management facilities

    International Nuclear Information System (INIS)

    Clifton, A.W.; Barsi, R.G.; Misfeldt, G.A.

    2002-01-01

    Uranium was discovered in the Beaverlodge area of northern Saskatchewan in 1934 with the first major mill beginning operation in 1953. Little attention was paid to tailings quality or tailings management practices. With the onset of the modern uranium operations beginning in the late 1970's, it was repeatedly evident, that the public had significant concerns, particularly with respect to tailings management, that must be addressed if the developments were to be allowed to proceed. Primary considerations related to environmental protection, public safety and an assurance of the ongoing sustainable development of the region. Integrating the decommissioning of a mine/mill site into development planning from the very outset has proven to be a critical component that has contributed to the ongoing success of the Saskatchewan uranium operations. This paper will provide a case study of the evolution of the uranium tailings management technology utilized in Saskatchewan. It documents the evolution of tailings management processes and the characteristics of tailings produced by successive mines in northern Saskatchewan. It also discusses the evolution of technologies applied to management of uranium mill tailings and demonstrates how progressively increasing levels of environmental protection have been achieved during the last 47 years of uranium mill operation. The paper also shows that the planned and progressive decommissioning of an operational site is the key to: Minimizing environmental impacts; Satisfying public and regulatory concerns; Minimizing operational and decommissioning costs; Minimizing corporate liability; and Shifting public resistance to public support. (author)

  12. Buckling and reaction rate experiments in plutonium/uranium metal fuelled, graphite moderated lattices at temperatures up to 400 deg. C. Part I: Experimental techniques and results

    Energy Technology Data Exchange (ETDEWEB)

    Carter, D H; Clarke, W G; Gibson, M; Hobday, R; Hunt, C; Marshall, J; Puckett, B J; Symons, C R; Wass, T [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1964-07-15

    This report presents experimental measurements of bucklings, flux fine structure and fission rate distributions in graphite moderated lattices fuelled with plutonium/uranium metal at temperatures up to 400 deg. C in the sub-critical assemblies SCORPIO I and SCORPIO II. The experimental techniques employed are described in some detail. The accuracy of the experimental measurements appears to be adequate for testing methods of calculation being developed for the calculation of reactivity and temperature coefficient of reactivity for power reactors containing plutonium and uranium. (author) 26 refs, 17 tabs, 17 figs

  13. CRITICALITY ANALYSIS OF URANIUM STORAGE FACILITY WITH FORMATION RACKS

    Directory of Open Access Journals (Sweden)

    Sri Kuntjoro

    2017-03-01

    ANALISIS KRITIKALITAS DI FASILITAS PENYIMPANAN BAHAN URANIUM DENGAN FORMASI PENGATURAN RAK. Bahan uranium dibutuhkan untuk produksi bahan bakar reaktor penelitian dan radioisotop. Bahan uranium sebelum digunakan terlebih dahulu disimpan pada fasilitas penyimpanan. Salah satu prasyarat fasilitas penyimpanan bahan uranium adalah fasilitas tersebut harus dalam kondisi sub-kritis. Bila kondisi kritis terjadi mengakibatkan proses fissi pada bahan uranium tidak terkendali, sehingga akan menimbulkan suhu yang sangat tinggi. Tujuan dari penelitian ini adalah untuk menganalisa kondisi kritikalitas dari fasilitas penyimpanan bahan uranium yang berada di PT. INUKI (Persero untuk menjamin fasilitas tersebut dalam kondisi sub-kritis. Analisis kritikalitas dilakukan menggunakan program MCNP-5 untuk mengetahui tingkat kritikalitas dari tiga fasilitas penyimpanan bahan uranium untuk kondisi awal dan kondisi setelah ditambahkan rak penyimpanan. Untuk fasilitas penyimpanan 1 dan 2 dibuat tiga skenario pengaturan container pada rak penyimpanan, sedangkan pada fasilitas penyimpanan 3 dilakukan 1 skenario.  Hasil ini menunjukkan seluruh fasilitas penyimpanan pada kondisi awal dan setelah ditambah rak penyimpanan dalam kondisi sub-kritis (k-eff<1. Hasil tersebut selanjutnya dipergunakan sebagai dasar untuk menyusun manejemen pengelolaan bahan uranium. Selain itu juga digunakan sebagai dasar untuk pembuatan ijin dari badan pengawas (BAPETEN. Kata Kunci : kritikalitas, fasilitas penyimpanan berbahan uranium,  k-eff

  14. Neutron study of fast neutron reactor systems by exponential experiments on Harmonie - Graphite program HUG-PHUG - Oxide program PHRIXOS - Uranium program UK

    International Nuclear Information System (INIS)

    Desprets, Alain.

    1977-12-01

    Exponential experiments allow to obtain the fundamental characteristics of a lattice (material buckling, reaction rate ratios) more economically than critical experiments. This report describes the experimental techniques and the methods of analysis used for this type of experiments. The results obtained with three programs performed with the source reactor HARMONIE are given: graphite-lattices program (3 U-fueled and 3 Pu-fueled lattices); oxide-fuel program (4 PuO 2 -UO 2 lattices); pure uranium program (one lattice). Some of these lattices were also studied in critical experiments. The coherence of the results obtained by the two types of experiments is established [fr

  15. Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France

    International Nuclear Information System (INIS)

    Gaussens, J.; Tanguy, P.

    1964-01-01

    The first part of this report defines the economic advantages of natural uranium fuels, which are as follows: the restricted number and relatively simple fabrication processes of the fuel elements, the low cost per kWh of the finished product and the reasonable capital investments involved in this type of fuel cycle as compared to that of enriched uranium. All these factors combine to reduce the arbitrary nature of cost estimates, which is particularly marked in the case of enriched uranium due to the complexity of its cycle and the uncertainties of plutonium prices). Finally, the wide availability of yellowcake, as opposed to the present day virtual monopoly of isotope separation, and the low cost of natural uranium stockpiling, offer appreciable guarantees in the way of security of supply and economic and political independence as compared with the use of enriched uranium. As far as overall capital investments are concerned, it is shown that, although graphite-gas reactor costs are higher than those of light water reactors in certain capacity ranges, the situation becomes far less clear when we start taking into account, in the interest of national independence, the cost of nuclear fuel production equipment in the case of each of these types of reactor. Finally, the marginal cost of the power capacity of a graphite-gas reactor is low and its technological limitations have receded (owing particularly to the use of prestressed concrete). It is a well known fact that the trend is now towards larger power station units, which means that the rentability of natural uranium graphite reactors as compared to other types of reactors will become more and more pronounced. The second section aims at presenting a realistic short and medium term view of the fuel, running, and investment costs of French natural uranium graphite gas, reactors. Finally, the economic goals which this type of reactor can reach in the very near future are given. It is thus shown that considerable

  16. Dosage of boron traces in graphite, uranium and beryllium oxide; Dosage de traces de bore dans le graphite, l'uranium et l'oxyde de beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Coursier, J [Ecole Nationale Superieure de Physique et Chimie Industrielles, 75 - Paris (France); Hure, J; Platzer, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    The problem of the dosage of the boron in the materials serving to the construction of nuclear reactors arises of the following way: to determine to about 0,1 ppm close to the quantities of boron of the order of tenth ppm. We have chosen the colorimetric analysis with curcumin as method of dosage. To reach the indicated contents, it is necessary to do a previous separation of the boron and the materials of basis, either by extraction of tetraphenylarsonium fluoborate in the case of the boron dosage in uranium and the beryllium oxide, either by the use of a cations exchanger resin of in the case of graphite. (M.B.) [French] Le probleme du dosage du bore dans les materiaux servant a la construction de reacteurs nucleaires se pose de la facon suivante: determiner a environ 0,1 ppm pres des quantites de bore de l'ordre de quelques dixiemes de ppm. Nous avons choisit la colorimetrie a la curcumine comme methode de dosage. Pour atteindre les teneurs indiquees, il est necessaire d'effectuer une separation prealable du bore et des materiaux de base, soit par extraction du fluoborate de tetraphenylarsonium dans le cas du dosage de bore dans l'uranium et l'oxyde de beryllium, soit par l'utilisation d'une resine echangeuse de cations dans le cas du graphite. (M.B.)

  17. Several issues of uranium geology exploration facilities decommissioning

    International Nuclear Information System (INIS)

    Zhang Lu; Lu Caixia; Sheng Qing; Zhuang Jingqi; Xie Shujun; Liao Yunxuan

    2013-01-01

    The environmental protection completion acceptance review work of uranium geology exploration facilities 'llth five-year plan' decommissioned and remediation projects is introduced. Some questions related to norms and standards for uranium geology exploration facilities decommissioning and remediation, scheme of decommissioning and remediation, process inspection and acceptance of project and so on are discussed, and corresponding countermeasures and suggestions are put forward, Some references can be provided for the later development of uranium geological exploration facility '12th five-year plan' decommissioning and remediation projects. (authors)

  18. Analysis of RA-8 critical facility core in some configurations

    International Nuclear Information System (INIS)

    Abbate, Maximo J.; Sbaffoni, Maria M.

    2000-01-01

    The RA-8 critical facility was designated and built to be used in the experimental plan of the 'CAREM' Project but is, in itself, very versatile and adequate to perform many types of other experiments. The present paper includes calculated estimates of some critical configurations and comparisons with experimental results obtained during its start up. Results for Core 1 with homogeneous arrangement of rods containing 1.8 % enriched uranium, showed very good agreement. In fact, an experimentally critical configuration was reached with 1.300 rods and calculated values were: 1.310 using the WIMS code and 1.148 from the CONDOR code. Moreover, it was verified that the estimated number of 3.4% enriched uranium rods to be fabricated is enough to build a heterogeneous core or even a homogeneous core with this enrichment. The replacement of 3.4 % enriched uranium by 3.6 % will not present problems related with the original plan. (author)

  19. Criticality Benchmark Analysis of Water-Reflected Uranium Oxyfluoride Slabs

    International Nuclear Information System (INIS)

    Marshall, Margaret A.; Bess, John D.

    2009-01-01

    A series of twelve experiments were conducted in the mid 1950's at the Oak Ridge National Laboratory Critical Experiments Facility to determine the critical conditions of a semi-infinite water-reflected slab of aqueous uranium oxyfluoride (UO2F2). A different slab thickness was used for each experiment. Results from the twelve experiment recorded in the laboratory notebook were published in Reference 1. Seven of the twelve experiments were determined to be acceptable benchmark experiments for the inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. This evaluation will not only be available to handbook users for the validation of computer codes and integral cross-section data, but also for the reevaluation of experimental data used in the ANSI/ANS-8.1 standard. This evaluation is important as part of the technical basis of the subcritical slab limits in ANSI/ANS-8.1. The original publication of the experimental results was used for the determination of bias and bias uncertainties for subcritical slab limits, as documented by Hugh Clark's paper 'Subcritical Limits for Uranium-235 Systems'.

  20. 77 FR 33782 - License Amendment To Construct and Operate New In Situ Leach Uranium Recovery Facility; Uranium...

    Science.gov (United States)

    2012-06-07

    ... and Operate New In Situ Leach Uranium Recovery Facility; Uranium One Americas; Ludeman AGENCY: Nuclear... provided the first time that a document is referenced. The Ludeman facility In Situ Leach Uranium Recovery... request to amend Source Material License SUA-1341 to construct and operate a new in situ leach uranium...

  1. Monte Carlo calculations of fast effects in uranium graphite lattices

    International Nuclear Information System (INIS)

    Beardwood, J.E.; Tyror, J.G.

    1962-12-01

    Details are given of the results of a series of computations of fast neutron effects in natural uranium metal/graphite cells. The computations were performed using the Monte Carlo code SPEC. It is shown that neutron capture in U238 is conveniently discussed in terms of a capture escape probability ζ as well as the conventional probability p. The latter is associated with the slowing down flux and has the classical exponential dependence on fuel-to-moderator volume ratio whilst the former is identified with the component of neutron flux above 1/E. (author)

  2. The nucleation and growth of uranium on the basal plane of graphite studied by scanning tunneling microscopy

    International Nuclear Information System (INIS)

    Tench, R.J.

    1992-11-01

    For the first time, nanometer scale uranium clusters were created on the basal plane of highly oriented pyrolytic graphite by laser ablation under ultra-high vacuum conditions. The physical and chemical properties of these clusters were investigated by scanning tunneling microscopy (STM) as well as standard surface science techniques. Auger electron and X-ray photoelectron spectroscopies found the uranium deposit to be free of contamination and showed that no carbide had formed with the underlying graphite. Clusters with sizes ranging from 42 Angstrom 2 to 630 Angstrom 2 were observed upon initial room temperature deposition. Surface diffusion of uranium was observed after annealing the substrate above 800 K, as evidenced by the decreased number density and the increased size of the clusters. Preferential depletion of clusters on terraces near step edges as a result of annealing was observed. The activation energy for diffusion deduced from these measurements was found to be 15 Kcal/mole. Novel formation of ordered uranium thin films was observed for coverages greater than two monolayers after annealing above 900 K. These ordered films displayed islands with hexagonally faceted edges rising in uniform step heights characteristic of the unit cell of the P-phase of uranium. In addition, atomic resolution STM images of these ordered films indicated the formation of the β-phase of uranium. The chemical properties of these surfaces were investigated and it was shown that these uranium films had a reduced oxidation rate in air as compared to bulk metal and that STM imaging in air induced a polarity-dependent enhancement of the oxidation rate

  3. 10 CFR 70.23a - Hearing required for uranium enrichment facility.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Hearing required for uranium enrichment facility. 70.23a... MATERIAL License Applications § 70.23a Hearing required for uranium enrichment facility. The Commission... license for construction and operation of a uranium enrichment facility. The Commission will publish...

  4. Proserpine - plutonium 239 - Proserpine - uranium 235 - comparison of experimental results; Proserpine - plutonium 239 - proserpine - uranium 235 - comparaison de resultats experimentaux

    Energy Technology Data Exchange (ETDEWEB)

    Brunet, J P; Caizergues, R; Clouet D' Orval, Ch; Kremser, J; Moret-Bailly, J; Verriere, Ph [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The Proserpine homogeneous reactor is constituted by a tank, 25 cm dia, 30 cm high, surrounded by a composite reflector made of beryllium oxide and graphite. In this tank can be made critical plutonium or 90 per cent enriched uranium solutions, the fissile substances being in the form of a dissolved salt. In varying the concentration of the solution, critical masses were studied as a function of the level of the liquid in the tank. The minimum critical mass is 256 {+-} 2 grs for plutonium and 409 {+-} 3 grs for uranium 235. In the range of the critical concentrations which were studied, the neutronic properties of fissionable solutions of plutonium and enriched uranium were compared for identical geometries. (authors) [French] Proserpine est un reacteur homogene comportant une cuve de diametre 25 cm, de hauteur 30 cm, entouree d'un reflecteur composite d'oxyde de beryllium et de graphite. On y a rendu critiques des solutions de plutonium ou d'uranium enrichi a 90 pour cent, le produit fissile se trouvant sous la forme d'un sel dissous. En faisant varier la concentration de la solution, on a etudie les masses critiques en fonction de la hauteur du liquide dans la cuve. La masse- critique minimum est, pour le plutonium de 256 {+-} 2 g, pour l'uranium 235 de 409 {+-} 3 g. Dans la gamme des concentrations critiques etudiees, on a compare, dans des conditions de geometrie identique, les proprietes neutroniques des solutions fissiles de plutonium et d'uranium enrichi. (auteurs)

  5. Organic tissues, graphite, and hydrocarbons in host rocks of the Rum Jungle Uranium Field, northern Australia

    Science.gov (United States)

    Foster, C.B.; Robbins, E.I.; Bone, Y.

    1990-01-01

    The Rum Jungle Uranium field consists of at least six early Proterozoic deposits that have been mined either for uranium and/or the associated base and precious metals. Organic matter in the host rocks of the Whites Formation and Coomalie Dolomite is now predominantly graphite, consistent with the metamorphic history of these rocks. For nine samples, the mean total organic carbon content is high (3.9 wt%) and ranged from 0.33 to 10.44 wt%. Palynological extracts from the host rocks include black, filamentous, stellate (Eoastrion-like), and spherical morphotypes, which are typical of early Proterozoic microbiota. The colour, abundance, and shapes of these morphotypes reflect the thermal history, organic richness, and probable lacustrine biofacies of the host rocks. Routine analysis of rock thin sections and of palynological residues shows that mineral grains in some of the host rocks are coated with graphitized organic matter. The grain coating is presumed to result from ultimate thermal degradation of a petroleum phase that existed prior to metamorphism. Hydrocarbons are, however, still present in fluid inclusions within carbonates of the Coomalie Dolomite and lower Whites Formation. The fluid inclusions fluoresce dull orange in blue-light excitation and their hydrocarbon content is confirmed by gas chromatography of whole-rock extracts. Preliminary analysis of the oil suggests that it is migrated, and because it has escaped graphitization through metamorphism it is probably not of early Proterozoic age. The presence of live oil is consistent with fluid inclusion data that suggest subsequent, low-temperature brine migration through the rocks. The present observations support earlier suggestions that organic matter in the host formations trapped uranium to form protore. Subsequent fluid migrations probably brought additional uranium and other metals to these formations, and the organic matter provided a reducing environment for entrapment. ?? 1990.

  6. Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

    International Nuclear Information System (INIS)

    Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed

    2013-01-01

    The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed

  7. Synthesis of graphite intercalation compound of group VI metals and uranium hexafluorides

    International Nuclear Information System (INIS)

    Fukui, Toshihiro; Hagiwara, Rika; Ema, Keiko; Ito, Yasuhiko

    1993-01-01

    Systematic investigations were made on the synthesis of graphite intercalation compounds of group VI transition metals (W and Mo) and uranium hexafluorides. The reactions were performed by interacting liquid or gaseous metal hexafluorides with or without elemental fluorine at ambient temperature. The degree of intercalation of these metal fluorides depends on the formation enthalpy of fluorometallate anion from the original metal hexafluoride, as has been found for other intercalation reactions of metal fluorides. (author)

  8. Chapter 8: Exponential experiments on graphite moderated lattices fuelled by natural uranium tubes containing cylindrical graphite cores

    International Nuclear Information System (INIS)

    McCulloch, D.B.; Hoskins, T.A.

    1963-01-01

    Experiments have been carried out using a fuel element comprising a 2.75 in. o.d./2.40 in. i.d. natural uranium tube containing a graphite core of diameter 2.0 in. Values of material buckling and migration area asymmetry for lattices at 7 in., 8 in. and 8/2 in. pitch have been obtained, and correlated with the theory of Syrett (1961) to derive an effective resonance integral for the cored element. By comparison with the resonance integral for the same fuel tube without a core, a value for the constant 'γ' of the theory of Stace (1959) is obtained. (author)

  9. Challenges of ICRP 60 for uranium refining and conversion facilities

    International Nuclear Information System (INIS)

    Takala, J.M.

    1998-01-01

    Cameco Corporation operates high-grade uranium mines in northern Saskatchewan and uranium refining and conversion facilities in Ontario. The dose limits for these and all other nuclear facilities in Canada are 50 mSv per year and 4 WLM per year, which are applied separately. However, the upcoming incorporation of the recommendations in ICRP 60 into the Canadian regulations will result in several important changes. In addition to a more restrictive dose limit, the new regulations will require that all radiation exposures be combined into a single index of exposure. Meeting the new lower dose limits of 50 mSv per year and 100 mSv per 5 years will not be a major problem at Cameco facilities. However, the incorporation of long-lived radioactive dust exposures into the dose calculation will be a major challenge. This will cause the most difficulty at the uranium refining and conversion facilities where much of the process involves handling a variety of uranium compounds in the form of a dry powder. At the uranium conversion facilities the control of exposure to airborne uranium is achieved through a combination of lung counting, urinalysis, and fixed area monitors. To progress from a system of exposure control to dose estimation to individual workers will require some major changes. (author)

  10. Criticality of mixtures of plutonium and high enriched uranium

    International Nuclear Information System (INIS)

    Grolleau, E.; Lein, M.; Leka, G.; Maidou, B.; Klenov, P.

    2003-01-01

    This paper presents a criticality evaluation of moderated homogeneous plutonium-uranium mixtures. The fissile media studied are homogeneous mixtures of plutonium and high enriched uranium in two chemical forms: aqueous mixtures of metal and mixtures of nitrate solutions. The enrichment of uranium considered are 93.2wt.% 235 U and 100wt.% 235 U. The 240 Pu content in plutonium varies from 0wt.% 240 Pu to 12wt.% 240 Pu. The critical parameters (radii and masses of a 20 cm water reflected sphere) are calculated with the French criticality safety package CRISTAL V0. The comparison of the calculated critical parameters as a function of the moderator-to-fuel atomic ratio shows significant ranges in which high enriched uranium systems, as well as plutonium-uranium mixtures, are more reactive than plutonium systems. (author)

  11. Several issues on the decommissioning of uranium mining/milling facilities

    International Nuclear Information System (INIS)

    Xu Lechang; Xu Jianxin; Gao Shangxiong

    2007-01-01

    Several issues on the decommissioning of uranium mining/milling facilities are discussed at the national and international level of decommissioning, including radiation, monitoring, dose evaluation, covering, water treatment and stabilization of uranium tailings impoundment, etc. Some suggestions are made: drawing international lessons on decommissioning of uranium mining/milling facilities; enhancing monitoring and database construction in decommissioning management; stressing utilization of measured dose data; using the experience of other countries for reference on covering designs for uranium tailings impoundment and water treatment; strengthening decommissioning management, etc. (authors)

  12. Uranium isotopic signatures measured in samples of dirt collected at two former uranium facilities

    International Nuclear Information System (INIS)

    Meyers, L.A.; Stalcup, A.M.; LaMont, S.P.; Spitz, H.B.

    2014-01-01

    Nuclear forensics is a multidisciplinary science that uses a variety of analytical methods and tools to explore the physical, chemical, and isotopic characteristics of nuclear and radiological materials. These characteristics, when evaluated alone or in combination, become signatures that may reveal how and when the material was fabricated. The signatures contained in samples of dirt collected at two different uranium metal processing facilities in the United States were evaluated to determine uranium isotopic composition and compare results with processes that were conducted at these sites. One site refined uranium and fabricated uranium metal ingots for fuel and targets and the other site rolled hot forged uranium and other metals into dimensional rods. Unique signatures were found that are consistent with the activities and processes conducted at each facility and establish confidence in using these characteristics to reveal the provenance of other materials that exhibit similar signatures. (author)

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

    International Nuclear Information System (INIS)

    Kanamori, Masashi; Yanagibashi, Katsumi; Okamoto, Naritoshi

    2002-12-01

    On September 30th 1999, the criticality accident occurred at JCO's uranium conversion building in Tokai. The accident occurred during reconversion from U 3 O 8 to uranium nitrate solution (UNH) with uranium enriched 18.8% and about 60 kgU. JCO contacted with JNC to supply UNH that is fuel material for the experimental fast breeder reactor 'JOYO'. JNC has contracted with JCO that had started nuclear fuel material processing business following a definite policy of Japanese government and developed SUMITOMO ADU PROCESS'. JNC made the first contract with JCO in 1985 and has made a contact every year. There had never been a problem in their products. JNC inspected products based on contract. JNC discharge our duty as customer inspecting products based on contract. As for safety control, JCO had taken licensing safety review and had been permitted to be 'a processing facility'. Therefore JNC understood that JCO produced following this license. 'The Uranium Processing Plant Criticality Accident Investigation' showed that JCO had been taking a different method from the permit and violating the license. However JNC had never been explained about that and JCO's operation procedures had never described about that. Therefore the Criticality Accident couldn't be avoided. This report describes the relationship of JNC and JCO in the uranium reconversion contract for JOYO, atomic development policy of Japanese government, process to the order and the contents of contract. (author)

  14. Derivation of release limits for a typical uranium mining and milling facility

    International Nuclear Information System (INIS)

    1985-09-01

    This report develops guidelines for calculating derived release limits (DRLs) for releases of each radionuclide belonging to the uranium-238 and thorium-232 decay chains to atmosphere, surface water and groundwater from uranium mining and milling operations in Canada. DRLs are defined as calculated limits on releases from the facility that result in radiation exposures through all environmental pathways equal to the annual effective dose equivalent limit of 0.005 Sv for stochastic effects or the annual dose equivalent limit of 0.05 Sv for non-stochastic effects in the critical group. By definition, DRLs apply to controllable radionuclide emissions which occur during the operational phase of mine/mill facilities. The report develops a steady-state environmental transfer model to determine environmental dilution and dispersion in atmosphere, surface water and groundwater between the sources at the mine and mill and the critical group receptor. Exposure pathways incorporated in the model include external exposure from immersion in the airborne plume, immersion in water, contaminated ground and contaminated shoreline sediments. Internal exposure pathways include inhalation of contaminated air and ingestion of contaminated water and terrestrial and aquatic foods

  15. A modern depleted uranium manufacturing facility

    International Nuclear Information System (INIS)

    Zagula, T.A.

    1995-07-01

    The Specific Manufacturing Capabilities (SMC) Project located at the Idaho National Engineering Laboratory (INEL) and operated by Lockheed Martin Idaho Technologies Co. (LMIT) for the Department of Energy (DOE) manufactures depleted uranium for use in the U.S. Army MIA2 Abrams Heavy Tank Armor Program. Since 1986, SMC has fabricated more than 12 million pounds of depleted uranium (DU) products in a multitude of shapes and sizes with varying metallurgical properties while maintaining security, environmental, health and safety requirements. During initial facility design in the early 1980's, emphasis on employee safety, radiation control and environmental consciousness was gaining momentum throughout the DOE complex. This fact coupled with security and production requirements forced design efforts to focus on incorporating automation, local containment and computerized material accountability at all work stations. The result was a fully automated production facility engineered to manufacture DU armor packages with virtually no human contact while maintaining security, traceability and quality requirements. This hands off approach to handling depleted uranium resulted in minimal radiation exposures and employee injuries. Construction of the manufacturing facility was complete in early 1986 with the first armor package certified in October 1986. Rolling facility construction was completed in 1987 with the first certified plate produced in the fall of 1988. Since 1988 the rolling and manufacturing facilities have delivered more than 2600 armor packages on schedule with 100% final product quality acceptance. During this period there was an annual average of only 2.2 lost time incidents and a single individual maximum radiation exposure of 150 mrem. SMC is an example of designing and operating a facility that meets regulatory requirements with respect to national security, radiation control and personnel safety while achieving production schedules and product quality

  16. Benchmarking of HEU mental annuli critical assemblies with internally reflected graphite cylinder

    Directory of Open Access Journals (Sweden)

    Xiaobo Liu

    2017-01-01

    Full Text Available Three experimental configurations of critical assemblies, performed in 1963 at the Oak Ridge Critical Experiment Facility, which are assembled using three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches metal annuli with internally reflected graphite cylinder are evaluated and benchmarked. The experimental uncertainties which are 0.00057, 0.00058 and 0.00057 respectively, and biases to the benchmark models which are − 0.00286, − 0.00242 and − 0.00168 respectively, were determined, and the experimental benchmark keff results were obtained for both detailed and simplified models. The calculation results for both detailed and simplified models using MCNP6-1.0 and ENDF/B-VII.1 agree well to the benchmark experimental results within difference less than 0.2%. The benchmarking results were accepted for the inclusion of ICSBEP Handbook.

  17. Proserpine - plutonium 239 - Proserpine - uranium 235 - comparison of experimental results

    International Nuclear Information System (INIS)

    Brunet, J.P.; Caizergues, R.; Clouet D'Orval, Ch.; Kremser, J.; Moret-Bailly, J.; Verriere, Ph.

    1964-01-01

    The Proserpine homogeneous reactor is constituted by a tank, 25 cm dia, 30 cm high, surrounded by a composite reflector made of beryllium oxide and graphite. In this tank can be made critical plutonium or 90 per cent enriched uranium solutions, the fissile substances being in the form of a dissolved salt. In varying the concentration of the solution, critical masses were studied as a function of the level of the liquid in the tank. The minimum critical mass is 256 ± 2 grs for plutonium and 409 ± 3 grs for uranium 235. In the range of the critical concentrations which were studied, the neutronic properties of fissionable solutions of plutonium and enriched uranium were compared for identical geometries. (authors) [fr

  18. Dosage of boron traces in graphite, uranium and beryllium oxide

    International Nuclear Information System (INIS)

    Coursier, J.; Hure, J.; Platzer, R.

    1955-01-01

    The problem of the dosage of the boron in the materials serving to the construction of nuclear reactors arises of the following way: to determine to about 0,1 ppm close to the quantities of boron of the order of tenth ppm. We have chosen the colorimetric analysis with curcumin as method of dosage. To reach the indicated contents, it is necessary to do a previous separation of the boron and the materials of basis, either by extraction of tetraphenylarsonium fluoborate in the case of the boron dosage in uranium and the beryllium oxide, either by the use of a cations exchanger resin of in the case of graphite. (M.B.) [fr

  19. 10 CFR 40.33 - Issuance of a license for a uranium enrichment facility.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Issuance of a license for a uranium enrichment facility... License Applications § 40.33 Issuance of a license for a uranium enrichment facility. (a) The Commission... the licensing of the construction and operation of a uranium enrichment facility. The Commission will...

  20. Fire criticality probability analysis for 300 Area N Reactor fuel fabrication and storage facility. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.E.

    1995-02-08

    Uranium fuel assemblies and other uranium associated with the shutdown N Reactor are stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility). The 3712 Building, where the majority of the fuel assemblies and other uranium is stored, is where there could be a potential for a criticality bounding case. The purpose of this study is to evaluate the probability of potential fires in the Facility 3712 Building that could lead to criticality. This study has been done to support the criticality update. For criticality to occur, the wooden fuel assembly containers would have to burn such that the fuel inside would slump into a critical geometry configuration, a sufficient number of containers burn to form an infinite wide configuration, and sufficient water (about a 17 inch depth) be placed onto the slump. To obtain the appropriate geometric configuration, enough fuel assembly containers to form an infinite array on the floor would have to be stacked at least three high. Administrative controls require the stacks to be limited to two high for 1.25 wt% enriched finished fuel. This is not sufficient to allow for a critical mass regardless of the fire and accompanying water moderation. It should be noted that 0.95 wt% enriched fuel and billets are stacked higher than only two high. In this analysis, two initiating events will be considered. The first is a random fire that is hot enough and sufficiently long enough to burn away the containers and fuel separators such that the fuel can establish a critical mass. The second is a seismically induced fire with the same results.

  1. On the separation of so-called non-volatile uranium fission products of uranium using the conversion of neutron-irradiated uranium dioxide and graphite

    International Nuclear Information System (INIS)

    Elhardt, W.

    1979-01-01

    The investigations are continued in the following work which arose from the concept of separating uranium fission products from uranium. This is achieved in that due to the lattice conversions occurring during the course of solid chemical reactions, fission products can easily pass from the uranium-contained solid to a second solid. The investigations carried out primarily concern the release behaviour of cerium and neodymium in the temperature region of 1200 to 1700 0 C. UO 2 + graphite, both in powder form, are selected as suitable reaction system having the preconditions needed for the lattice conversion for the release effect. The target aimed at from the practical aspect for the improved release of lanthanoids is achieved by an isobar test course - changing temperature from 1200 to 1500 0 C at constant pressure, with a cerium release of 75-80% and a neodynium release of 80-90% (maximum at 1400 0 C). The concepts on the mechanism of the fission product release are related to transport processes in crystal lattices, as well as chemical solid reactions and evaporation processes on the surface of UC 2 grains. (orig./RB) [de

  2. Criticality analyses of regions containing uranium in the earth history

    International Nuclear Information System (INIS)

    Ravnik, M.

    2005-01-01

    Investigations of necessary conditions for a self-sustained chain reaction in the Earth inner regions hypothetically containing uranium is presented for the time interval from Earth formation to present time. It is determined that criticality was theoretically possible up to 2.5 Ga before present if uranium concentrated in pure form. In the early geological history (4 Ga before present) the self-sustained criticality could occur even if uranium was diluted up to 1:20 by the average core material or 1:60 by the average mantle material. If other metallic materials of similar density as uranium (e.g., Au, W) or similar atomic weight (e.g., Th) concentrated from the primordial mixture in equal proportion as uranium, criticality was not possible for any period in Earth history provided that the basic material contained no light nuclides (H, C). Criticality in the Earth inner regions could have established only if uranium concentrated from the basic material more effectively than elements of similar density or atomic number. (orig.)

  3. Guide to the bioassay of uranium at uranium mine-mill facilities

    International Nuclear Information System (INIS)

    1981-01-01

    As a result of occupational exposure, uranium may be taken into the body by inhalation, ingestion or absorption through skin wounds. The organs at risk are the lung, kidney, and bones. Analysis of urine samples for uranium is recommended on a regular monthly basis, before and after a rest period, and it is suggested that a worker be removed from a working area if a level above 300 μg/l is found before a rest period, or 150 μg/l after a rest period. Background information on the development of a bioassay program is given, and a recommended program for uranium mine and mill facilities is included. (L.L.)

  4. Description of an engineering-scale facility for uranium fluorination studies

    International Nuclear Information System (INIS)

    Yagi, Eiji; Saito, Shinichi; Horiuchi, Masato

    1976-03-01

    In the research program of power reactor fuel reprocessing by fluoride volatility process, the engineering facility was constructed to establish the techniques of handling kilogram quantities of fluorine and uranium hexafluoride and to obtain engineering data on the uranium fluidized-bed oxidation and fluorination. This facility is designed for a capacity of 5 kg per batch. Descriptions on the facility and equipment are given, including design philosophy, safety and its analysis. (auth.)

  5. Radiation protection of workers from uranium mines and of the public living nearby uranium mining and milling facilities

    International Nuclear Information System (INIS)

    Kiselev, Mikhail; Romanov, Vladimir; Shandala, Nataliya; Gneusheva, Galina; Titov, Alex; Novikova, Natalia; Smith, Graham

    2008-01-01

    As part of the program of nuclear power development, the Russia Federation plans to increase uranium production and to improve supply from existing uranium mining and milling facilities. Moreover, development of new uranium ore deposits is also envisaged. A corollary of these developments is the placing of a high priority on environmental and human health protection Special attention should be paid to assurance of health protection both of workers and of the public living nearby such facilities. This paper reviews the status and development of understanding of facilities in the Russian Federation from a regulatory perspective. (author)

  6. Uranium Production Safety Assessment Team. UPSAT. An international peer review service for uranium production facilities

    International Nuclear Information System (INIS)

    1996-01-01

    The IAEA Uranium Production Safety Assessment Team (UPSAT) programme is designed to assist Member States to improve the safe operation of uranium production facilities. This programme facilitates the exchange of knowledge and experience between team members and industry personnel. An UPSAT mission is an international expert review, conducted outside of any regulatory framework. The programme is implemented in the spirit of voluntary co-operation to contribute to the enhancement of operational safety and practices where it is most effective, at the facility itself. An UPSAT review supplements other facility and regulatory efforts which may have the same objective

  7. Remotely operated facility for in situ solidification of fissile uranium

    International Nuclear Information System (INIS)

    McGinnis, C.P.; Collins, E.D.; Patton, B.D.

    1986-01-01

    A heavily shielded, remotely operated facility, located within the Radiochemical processing Plant at Oak Ridge National Laboratory (ORNL), has been designed and is being operated to convert approx.1000 kg of fissile uranium (containing approx.75% 235 U, approx.10% 233 U, and approx.140 ppM 232 U) from a nitrate solution (130 g of uranium per L) to a solid oxide form. This project, the Consolidated Edison Uranium Solidification Program (CEUSP), is being carried out in order to prepare a stable uranium form for longterm storage. This paper describes the solidification process selected, the equipment and facilities required, the experimental work performed to ensure successful operation, some problems that were solved, and the initial operations

  8. Potential criticality accident at the General Electric Nuclear Fuel and Component Manufacturing Facility, May 29, 1991

    International Nuclear Information System (INIS)

    1991-08-01

    At the General Electric Nuclear Fuel and Component Manufacturing facility, located near Wilmington, North Carolina, on May 28 and 29, 1991, approximately 150 kilograms of uranium were inadvertently transferred from safe process tanks to an unsafe tank located at the waste treatment facility, thus creating the potential for a localized criticality safety problem. The excess uranium was ultimately safely recovered when the tank contents were centrifuged to remove the uranium-bearing material. Subsequently, the US Nuclear Regulatory Commission dispatched an Incident Investigation Team to determine what happened, to identify probable causes, and to make appropriate findings and conclusions. This report describes the incident, the methodology used by the team in its investigation, and presents the team's findings and conclusions. 48 figs., 8 tabs

  9. Restoration activities in uranium mining and milling facilities in Spain

    International Nuclear Information System (INIS)

    Garcia Quiros, J.M.

    1997-01-01

    From the end of the 80's up to now, several tasks have been carried out in Spain on restoration in the field of uranium mining and milling, significant among them being Andujar Uranium Mill (FUA) closure and La Haba closure. Also, a study has been carried out on restoration of inoperative and abandoned uranium mine sites. At present, detailed plans are being worked out for the project on the closure of the Elefante plant. All activities have been developed in the common framework of national standards and regulations which are generally in compliance with the standards, regulations and recommendations of international organizations. This paper describes briefly the standards and the criteria applied to the restoration tasks at various sites of the uranium mining and milling facilities in Spain. The restoration activities have different characteristics La Haba facility is an isolated and conventional facility to produce uranium concentrate; in the case of old and abandoned uranium mines the intervention criteria is more relevant than the activities to be carried out; the closure (the first phase of licensing) and restoration activities of Elefante plant have to be developed taking into account that it is sited within the area of Quercus plant which is currently in operation. (author)

  10. International safeguards for critical facilities

    International Nuclear Information System (INIS)

    Ney, J.F.; Todd, J.L.

    1979-01-01

    A study was undertaken to investigate various approaches to provide international safeguards for critical facilities and to select an optimized system. Only high-inventory critical facilities were considered. The goal of the study was to detect and confirm the protracted or abrupt diversion of 8kg of plutonium or 25kg of the uranium isotope 235 within approximately a week of the diversion. The general safeguards alternatives considered were (1) continuous inspections by resident inspectors, with varying degrees of comprehensiveness, (2) periodic inspections by regional inspectors at varying time intervals, (3) unattended containment/surveillance measures, and (4) various combinations of the above. It was concluded that a practical and effective international safeguards system can be achieved by employing a method of continuously monitoring facility activities which could lead to diversion. This is in addition to the routine inspections typical of current international safeguards. Monitoring detects inventory discrepancies and violations of agreed-upon procedural restrictions, as well as unauthorized removal of Special Nuclear Materials (SNM). A special inventory is used following detection to confirm any suspected diversion. Comparison of 28 safeguards options led to the selection of a system for further development which uses a combination of surveillance and inspection by resident IAEA personnel, containment/surveillance by unattended equipment, and routine inventory sampling. A development programme is described which is intended to demonstrate the feasibility of several containment and surveillance measures proposed in the study. Included are a personnel portal and an instrument/material pass-through as well as associated recording and tamper-protection features. (author)

  11. Minimum critical masses for uranium at the Portsmouth Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    Tayloe, R.W. Jr.; Davis, T.C.

    1994-06-01

    This report presents a tabulation of safe masses and minimum critical masses for uranium (U). These minimum critical mass and safe mass tables were obtained by interpolating between the values reported in the literature to obtain values as a function of enrichment within the 1.5 percent to 100 percent range. Equivalent mass values for uranium-235 (U 235 ), uranium hexafluoride (UF 6 ), and uranyl fluoride (UO 2 F 2 ) have been generated from the safe mass and minimum critical masses for uranium

  12. Criticality of moderated and undermoderated low-enriched uranium oxide systems

    International Nuclear Information System (INIS)

    Goebel, G.R.

    1980-06-01

    Uranium oxide was enriched to 4.46 wt % 235 U compacted to a density of 4.68 g/cm 3 . The uranium oxide was packed into cubical aluminum cans and water added to the oxide until an H/U atomic ratio of 0.77 was achieved. A 5 x 5 x 5 array of uranium oxide cans for the experiments were used when no plastic moderator material was placed between cans. High enriched uranium drivers were used to achieve criticality. Criticality was achieved for smaller arrays without a driver when 24.5 mm plastic moderator material was placed between the cans. Twelve critical experiments are reported, six in each reflector

  13. NARCISS critical stand experiments for studying the nuclear safety in accident water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoj, N.N.; Glushkov, E.S.; Bubelev, V.G.

    2005-01-01

    A brief description of the Topaz-2 SNPS designed under scientific supervision of RRC KI in Russia, and of the NARCISS critical facility, is given. At the NARCISS critical facility, neutronic peculiarities and nuclear safety issues of the Topaz-2 system reactor were studied experimentally. This work is devoted to a detailed description of experiments on investigation of criticality safety in accident water immersion og highly enriched uranium dioxide fuel elements, performed at the NARCISS facility. The experiments were carried out at water-moderated critical assemblies with varying height, number, and spacing of fuel elements. The results obtained in the critical experiments, computational models of the investigated critical configurations, and comparison of the computational and experimental results are given [ru

  14. Control and prevention of seepage from uranium mill waste disposal facilities

    International Nuclear Information System (INIS)

    Williams, R.E.

    1978-01-01

    This paper constitutes an analysis of the technologies which are available for the prevention of movement of waste waters out of uranium mill waste disposal facilities via sub-surface routes. Hydrogeologic criteria for potential uranium mill waste disposal sites and mathematical modeling of contaminant migration in ground water are presented. Methods for prevention of seepage from uranium mill waste disposal facilities are investigated: liners, clay seals, synthetic polymeric membranes (PVC, polyethylene, chlorinated polyethylene, hypalon, butyl rubber, neoprene, elasticized polyolefin)

  15. Facility effluent monitoring plan for the uranium trioxide facility

    International Nuclear Information System (INIS)

    Thompson, R.J.; Sontag, S.

    1991-11-01

    A facility effluent monitoring plant is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan is the first annual report. It shall ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated as a minimum every three years. The UO 3 Plant is located in the south-central portion of the 200 West Area of the Hanford Site. The plant consists of two primary processing buildings and several ancillary facilities. The purpose of the UO 3 Plant is to receive uranyl nitrate hexahydrate (UNH) from the Plutonium-Uranium Extraction (PUREX) Plant, concentrate it, convert the UNH to uranium trioxide (UO 3 ) powder by calcination and package it for offsite shipment. The UO 3 Plant has been placed in a standby mode. There are two liquid discharges, and three gaseous exhaust stacks, and seven building exhausters that are active during standby conditions

  16. Plans and equipment for criticality measurements on plutonium-uranium nitrate solutions

    International Nuclear Information System (INIS)

    Lloyd, R.C.; Clayton, E.D.; Durst, B.M.

    1982-01-01

    Data from critical experiments are required on the criticality of plutonium-uranium nitrate solutions to accurately establish criticality control limits for use in processing and handling of breeder type fuels. Since the fuel must be processed both safely and economically, it is necessary that criticality considerations be based on accurate experimental data. Previous experiments have been reported on plutonium-uranium solutions with Pu weight ratios extending up to some 38 wt %. No data have been presented, however, for plutonium-uranium nitrate solutions beyond this Pu weight ratio. The current research emphasis is on the procurement of criticality data for plutonium-uranium mixtures up to 60 wt % Pu that will serve as the basis for handling criticality problems subsequently encountered in the development of technology for the breeder community. Such data also will provide necessary benchmarks for data testing and analysis on integral criticality experiments for verification of the analytical techniques used in support of criticality control. Experiments are currently being performed with plutonium-uranium nitrate solutions in stainless steel cylindrical vessels and an expandable slab tank system. A schematic of the experimental systems is presented

  17. 77 FR 60482 - Regulatory Guide 5.67, Material Control and Accounting for Uranium Enrichment Facilities...

    Science.gov (United States)

    2012-10-03

    ... Accounting for Uranium Enrichment Facilities Authorized To Produce Special Nuclear Material of Low Strategic... Accounting for Uranium Enrichment Facilities Authorized to Produce Special Nuclear Material of Low Strategic... and is applicable to the Paducah GDP and other uranium enrichment facilities that have been licensed...

  18. Decommissioning of nuclear facilities involving operations with uranium and thorium

    International Nuclear Information System (INIS)

    Shum, E.Y.; Neuder, S.M.

    1990-01-01

    When a licensed nuclear facility ceases operation, the U.S. Nuclear Regulatory Commission (NRC) ensures that the facility and its site are decontaminated to acceptable levels so they may safely be released for unrestricted public use. Because specific environmental standards or broad federal guidelines governing release of residual radioactive contamination have not been issued, NRC has developed ad hoc cleanup criteria for decommissioning nuclear facilities that involved uranium and thorium. Cleanup criteria include decontamination of buildings, equipment, and land. We will address cleanup criteria and their rationale; procedures for decommissioning uranium/thorium facilities; radiological survey designs and procedures; radiological monitoring and measurement; and cost-effectiveness to demonstrate compliance

  19. Study of the thermal drop at the uranium-can interface for fuel elements in gas-graphite reactors; Etude de la chute thermique au contact uranium-gaine pour des elements combustibles de reacteur de la filiere graphite-gaz

    Energy Technology Data Exchange (ETDEWEB)

    Faussat, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Levenes, G; Michel, M [Societe Industrielle de Combustible Nucleaire (France)

    1964-07-01

    The report reviews the tests now under way at the CEA, for determining the thermal contact resistance at the uranium-can interface for fuel elements used in gas-graphite type reactors. These are laboratory tests carried out with equipment based on the principle of a heat flow across a stack of test pieces having planar contact surfaces. The following points emerge from this work: - for a metallic uranium element canned in magnesium, of the type G-2 or EDF-2, a value of 0.2 deg C/W/cm{sup 2} seems reasonable for can temperatures of 400 deg C and above. - this value is independent of the micro-geometric state of the uranium surface in a range of roughness which easily includes those observed on tubes and rods produced industrially. - for the internal cans of elements cooled internally and externally, the value of the contact resistance for temperatures of under 400 deg C as a function of the stresses in the can has not yet been measured exactly. (authors) [French] Le rapport fait le point des essais actuellement en cours au CEA pour determiner la resistance thermique de contact uranium-gaine pour des reacteurs de la filiere graphite-gaz. Ces essais sont effectues en laboratoire sur des appareils bases sur le principe d'une circulation de flux de chaleur a travers un empilement d'eprouvettes dont les faces en contact sont planes. De l'etude, il ressort essentiellement que: - pour un element a uranium metallique et gaine de magnesium type G-2 ou EdF-2, on peut admettre la valeur de 0,2 deg C/W/cm{sup 2} pour des temperatures de gaines de 400 deg C et plus. - cette valeur ne depend pas de l'etat de surface microgeometrique de l'uranium pour un domaine de rugosites couvrant largement celles que l'on observe sur des tubes et barreaux fabriques en serie. - pour les gaines internes d'elements a refroidissement interne et externe la valeur de la resistance de contact reste a preciser pour les temperatures inferieures a 400 deg C, en fonction des contraintes existant dans les

  20. Reference materials for nondestructive assay of special nuclear material. Volume 1. Uranium oxide plus graphite powder

    International Nuclear Information System (INIS)

    Sprinkle, J.K.; Likes, R.N.; Parker, J.L.; Smith, H.A.

    1983-10-01

    This manual describes the fabrication of reference materials for use in gamma-ray-based nondestructive assay of low-density uranium-bearing samples. The sample containers are 2-l bottles. The reference materials consist of small amounts of UO 2 spread throughout a graphite matrix. The 235 U content ranges from 0 to 100 g. The manual also describes the far-field assay procedure used with low-resolution detectors

  1. Decommissioning and disposal of foreign uranium mine and mill facilities

    International Nuclear Information System (INIS)

    Pan Yingjie; Xue Jianxin; Yuan Baixiang; Xu Lechang

    2012-01-01

    Disposal techniques in decommissioning of foreign uranium mine and mill facilities are systematically discussed, including covering of uranium tailing impoundment, drainaging and consolidation of uranium tailing, and treatment of mining waste water and polluted groundwater, and the costs associated with disposal are analyzed. The necessity of strengthening the decommissioning disposal technology research and international exchanges and cooperation is emphasized. (authors)

  2. Criticality calculations for homogeneous mixtures of uranium and plutonium

    International Nuclear Information System (INIS)

    Spiegelberg, R. de S.H.

    1981-05-01

    Critical parameters were calculated using the one-dimensional multigroup transport theory. Calculations have been performed for water mixture of uranium metal and uranium oxides and plutonium nitrates to determine the dimensions of simple critical geometries (sphere and cylinder). The results of the calculations were plotted showing critical parameters (volume, radius or critical mass). The critical values obtained in Handbuch zur Kritikalitat were used to compare with critical parameters. A sensitivity study for the influences of mesh space size, multigroup structure and order of the S sub(n) approximation on the critical radius was carried out. The GAMTEC-II code was used to generate multigroup cross sections data. Critical radius were calculated using the one-dimensional multigroup transport code DTF-IV. (Author) [pt

  3. Physics experiments in graphite lattices (1962); Experiences sur les reseaux a graphite (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1962-07-01

    A review is made of the various experimental methods used to determine the physics of graphite, natural uranium lattices: integral lattice experiments; both absolute and differential, effective cross section measurements, both by activation methods and by analysis of irradiated fuels, fine structure measurements. A number of experimental results are also given. (authors) [French] On decrit les differentes methodes experimentales utilisees pour determiner les parametres physiques de reseaux a uranium-graphite. Il s'agit d'experiences globales: mesures absolues et relatives de laplaciens, mesures de sections efficaces effectives par activation et par analyses de combustibles irradies, mesures de structures fines. Un certain nombre de resultats experimentaux sont communiques. (auteurs)

  4. The regulation of uranium refineries and conversion facilities in Canada

    International Nuclear Information System (INIS)

    Didyk, J.P.

    1986-04-01

    The nuclear regulatory process as it applies to uranium refineries and conversion facilities in Canada is reviewed. In the early 1980s, Eldorado Resources Limited proposed to construct and operate new facilities for refining yellowcake and for the production of uranium hexafluoride (UF 6 ). These projects were subject to regulation by the Atomic Energy Control Board (AECB). A description of the AECB's comprehensive licensing process covering all stages of siting, construction, operation and eventual decommissioning of nuclear facilities is traced as it was applied to the Eldorado projects. The AECB's concern with occupational health and safety, with public health and safety and with the protection of the environment in so far as it affects public health and safety is emphasized. Some regulatory difficulties encountered during the project's development which led to opening up the licensing process to public input and closer coordination of regulatory activities with other provincial and federal regulatory agencies are described. The Board's regulatory operational compliance program for uranium refineries and conversion facilities is summarized

  5. Production yields of noble-gas isotopes from ISOLDE UC$_{x}$/graphite targets

    CERN Document Server

    Bergmann, U C; Catherall, R; Cederkäll, J; Diget, C A; Fraile-Prieto, L M; Franchoo, S; Fynbo, H O U; Gausemel, H; Georg, U; Giles, T; Hagebø, E; Jeppesen, H B; Jonsson, O C; Köster, U; Lettry, Jacques; Nilsson, T; Peräjärvi, K; Ravn, H L; Riisager, K; Weissman, L; Äystö, J

    2003-01-01

    Yields of He, Ne, Ar, Kr and Xe isotopic chains were measured from UC$_{x}$/graphite and ThC$_{x}$/graphite targets at the PSB-ISOLDE facility at CERN using isobaric selectivity achieved by the combination of a plasma-discharge ion source with a water-cooled transfer line. %The measured half-lives allowed %to calculate the decay losses of neutron-rich isotopes in the %target and ion-source system, and thus to obtain information on the in-target %productions from the measured yields. The delay times measured for a UC$_x$/graphite target allow for an extrapolation to the expected yields of very neutron-rich noble gas isotopes, in particular for the ``NuPECC reference elements'' Ar and Kr, at the next-generation radioactive ion-beam facility EURISOL. \\end{abstract} \\begin{keyword} % keywords here, in the form: keyword \\sep keyword radioactive ion beams \\sep release \\sep ion yields \\sep ISOL (Isotope Separation On-Line) \\sep uranium and thorium carbide targets. % PACS codes here, in the form: \\PACS code \\sep code...

  6. Benchmark calculations for critical experiments at FKBN-M facility with uranium-plutonium-polyethylene systems using JENDL-3.2 and MVP Monte-Carlo code

    International Nuclear Information System (INIS)

    Obara, Toru; Morozov, A.G.; Kevrolev, V.V.; Kuznetsov, V.V.; Treschalin, S.A.; Lukin, A.V.; Terekhin, V.A.; Sokolov, Yu.A.; Kravchenko, V.G.

    2000-01-01

    Benchmark calculations were performed for critical experiments at FKBN-M facility in RFNC-VNIITF, Russia using JENDL-3.2 nuclear data library and continuous energy Monte-Carlo code MVP. The fissile materials were high-enriched uranium and plutonium. Polyethylene was used as moderator. The neutron spectrum was changed by changing the geometry. Calculation results by MVP showed some errors. Discussion was made by reaction rates and η values obtained by MVP. It showed the possibility that cross sections of U-235 had different trend of error in fast and thermal energy region respectively. It also showed the possibility of some error of cross section of Pu-239 in high energy region. (author)

  7. SRTC criticality safety technical review: Nuclear Criticality Safety Evaluation 93-04 enriched uranium receipt

    International Nuclear Information System (INIS)

    Rathbun, R.

    1993-01-01

    Review of NMP-NCS-930087, open-quotes Nuclear Criticality Safety Evaluation 93-04 Enriched Uranium Receipt (U), July 30, 1993, close quotes was requested of SRTC (Savannah River Technology Center) Applied Physics Group. The NCSE is a criticality assessment to determine the mass limit for Engineered Low Level Trench (ELLT) waste uranium burial. The intent is to bury uranium in pits that would be separated by a specified amount of undisturbed soil. The scope of the technical review, documented in this report, consisted of (1) an independent check of the methods and models employed, (2) independent HRXN/KENO-V.a calculations of alternate configurations, (3) application of ANSI/ANS 8.1, and (4) verification of WSRC Nuclear Criticality Safety Manual procedures. The NCSE under review concludes that a 500 gram limit per burial position is acceptable to ensure the burial site remains in a critically safe configuration for all normal and single credible abnormal conditions. This reviewer agrees with that conclusion

  8. Buckling and reaction rate measurements in graphite moderated lattices fuelled with plutonium-uranium oxide clusters at temperatures up to 400 deg. C

    International Nuclear Information System (INIS)

    Carter, D.H.; Gibson, M.; King, D.C.; Marshall, J.; Puckett, B.J.; Richards, A.E.; Wass, T.; Wilson, D.J.

    1965-07-01

    The Report describes a series of experiments carried out in SCORPIO I and II on sub-critical graphite moderated lattices fuelled with 21-rod clusters of PuO 2 /UO 2 fuel. Three fuel batches with nominal plutonium: uranium ratios of 0.25%, 0.8% and 1.2% were investigated at temperatures between 20 deg. C and 400 deg. C. Because of the limited amounts of the three fuels, exponential measurements were made in 2-zone stacks, the outer regions of which were loaded with suitably matched 'reference fuel'. Fine structure distributions in the lattice cell were obtained with manganese and indium foils. Pu239/U235 fission ratios were determined both by fission chambers and by fission-product counting techniques. (author)

  9. Nuclear criticality safety parameter evaluation for uranium metallic alloy

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Andrea; Abe, Alfredo, E-mail: andreasdpz@hotmail.com, E-mail: abye@uol.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Energia Nuclear

    2013-07-01

    Nuclear criticality safety during fuel fabrication process, transport and storage of fissile and fissionable materials requires criticality safety analysis. Normally the analysis involves computer calculations and safety parameters determination. There are many different Criticality Safety Handbooks where such safety parameters for several different fissile mixtures are presented. The handbooks have been published to provide data and safety principles for the design, safety evaluation and licensing of operations, transport and storage of fissile and fissionable materials. The data often comprise not only critical values, but also subcritical limits and safe parameters obtained for specific conditions using criticality safety calculation codes such as SCALE system. Although many data are available for different fissile and fissionable materials, compounds, mixtures, different enrichment level, there are a lack of information regarding a uranium metal alloy, specifically UMo and UNbZr. Nowadays uranium metal alloy as fuel have been investigated under RERTR program as possible candidate to became a new fuel for research reactor due to high density. This work aim to evaluate a set of criticality safety parameters for uranium metal alloy using SCALE system and MCNP Monte Carlo code. (author)

  10. 10 CFR 140.13b - Amount of liability insurance required for uranium enrichment facilities.

    Science.gov (United States)

    2010-01-01

    ... enrichment facilities. 140.13b Section 140.13b Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL... required for uranium enrichment facilities. Each holder of a license issued under Parts 40 or 70 of this chapter for a uranium enrichment facility that involves the use of source material or special nuclear...

  11. Secondary limits of exposure in facilities handling uranium

    International Nuclear Information System (INIS)

    Raghavayya, M.

    1999-08-01

    Annual limits of exposure and intake for radiation workers in nuclear installations have been recommended by the International Commission on Radiological Protection and the same have been adopted by the Indian Atomic Energy Regulatory Board for all the radionuclides of interest. The prescribed limits cannot be directly used for day to day radiation protection work. Hence secondary limits have to be derived for routine applications. The modeling steps may be simple in some situations and more complicated in some others. The limits recommended are for individual radionuclides. But in facilities handling natural or enriched uranium the radionuclides (isotopes of uranium and its decay products) generally occur together in specific ratios. Derivation of secondary limits has to take this into consideration. The present document is an attempt at deriving the secondary limits required for routine application in facilities handling uranium (Mine, mill, refineries and fuel fabrication etc.). Secondary limits of exposure have been derived in this document for air borne activity, activity in water, surface contamination and internal exposures. (author)

  12. Chapter 5: Exponential experiments on natural uranium graphite moderated systems. II: Correlation of results with the method of Syrett (1961)

    International Nuclear Information System (INIS)

    Brown, G.; Moore, P.G.F.; Richmond, R.

    1963-01-01

    The results are given of exponential experiments on graphite moderated systems with fuel elements consisting of single rods and tubes of natural uranium metal. A correlation is given with the method of calculation proposed by Syrett (1961) and new consistent values of neutron yield and effective resonance integral are derived. (author)

  13. The Bare Critical Assembly of Natural Uranium and Heavy Water

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

    1958-07-01

    The first reactor built in Yugoslavia was the bare zero energy heavy water and natural uranium assembly at the Boris Kidric Institute of Nuclear Sciences, Belgrade. The reactor went critical on April 29, 1958. The possession of four tons of natural uranium metal and the temporary availability of seven tons of heavy water encouraged the staff of the Institute to build a critical assembly. A critical assembly was chosen, rather than high flux reactor, because the heavy water was available only temporarily. Besides, a 10 MW, enriched uranium, research reactor is being built at the same Institute and should be ready for operation late this year. It was supposed that the zero energy reactor would provide experience in carrying out critical experiments, operational experience with nuclear reactors, and the possibility for an extensive program in reactor physics. (author)

  14. Criticality safety considerations for MSRE fuel drain tank uranium aggregation

    International Nuclear Information System (INIS)

    Hollenbach, D.F.; Hopper, C.M.

    1997-01-01

    This paper presents the results of a preliminary criticality safety study of some potential effects of uranium reduction and aggregation in the Molten Salt Reactor Experiment (MSRE) fuel drain tanks (FDTs) during salt removal operations. Since the salt was transferred to the FDTs in 1969, radiological and chemical reactions have been converting the uranium and fluorine in the salt to UF 6 and free fluorine. Significant amounts of uranium (at least 3 kg) and fluorine have migrated out of the FDTs and into the off-gas system (OGS) and the auxiliary charcoal bed (ACB). The loss of uranium and fluorine from the salt changes the chemical properties of the salt sufficiently to possibly allow the reduction of the UF 4 in the salt to uranium metal as the salt is remelted prior to removal. It has been postulated that up to 9 kg of the maximum 19.4 kg of uranium in one FDT could be reduced to metal and concentrated. This study shows that criticality becomes a concern when more than 5 kg of uranium concentrates to over 8 wt% of the salt in a favorable geometry

  15. Buckling and reaction rate measurements in graphite moderated lattices fuelled with plutonium-uranium oxide clusters at temperatures up to 400 deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Carter, D H; Gibson, M; King, D C; Marshall, J; Puckett, B J; Richards, A E; Wass, T; Wilson, D J [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1965-07-15

    The Report describes a series of experiments carried out in SCORPIO I and II on sub-critical graphite moderated lattices fuelled with 21-rod clusters of PuO{sub 2}/UO{sub 2} fuel. Three fuel batches with nominal plutonium: uranium ratios of 0.25%, 0.8% and 1.2% were investigated at temperatures between 20 deg. C and 400 deg. C. Because of the limited amounts of the three fuels, exponential measurements were made in 2-zone stacks, the outer regions of which were loaded with suitably matched 'reference fuel'. Fine structure distributions in the lattice cell were obtained with manganese and indium foils. Pu239/U235 fission ratios were determined both by fission chambers and by fission-product counting techniques. (author) 14 refs, 30 figs, 18 tabs

  16. Alecto - results obtained with homogeneous critical experiments on plutonium 239, uranium 235 and uranium 233

    International Nuclear Information System (INIS)

    Bruna, J.G.; Brunet, J.P.; Caizegues, R.; Clouet d'Orval, Ch.; Kremser, J.; Tellier, H.; Verriere, Ph.

    1965-01-01

    In this report are given the results of the homogeneous critical experiments ALECTO, made on plutonium 239, uranium 235 and uranium 233. After a brief description of the equipment, the critical masses for cylinders of diameters varying from 25 to 42 cm, are given and compared with other values (foreign results, criticality guide). With respect to the specific conditions of neutron reflection in the ALECTO experiments the minimal values of critical masses are: Pu239 M c = 910 ± 10 g, U235 M c = 1180 ± 12 g and U233 M c = 960 ± 10 g. Experiments relating to cross sections and constants to be used on these materials are presented. Lastly, kinetic experiments allow to compare pulsed neutron methods to fluctuation methods [fr

  17. Alternative repository criticality-control strategies for fissile uranium wastes

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1998-01-01

    Methods to prevent long term, disposal site nuclear criticality from fissile uranium isotopes in wastes were investigated. Long term refers to the time period after waste package (WP) failure and the subsequent loss of geometry and chemistry control within the WP. The preferred method of control was found to be the addition of sufficient depleted uranium to each WP so that the uranium enrichment is reduced to 235 U and 233 U in 238 U

  18. Criticality safety evaluation of the fuel cycle facility electrorefiner

    International Nuclear Information System (INIS)

    Lell, R.M.; Mariani, R.D.; Fujita, E.K.; Benedict, R.W.; Turski, R.B.

    1993-01-01

    The integral Fast Reactor (IFR) being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal cooled reactors and a closed-loop fuel cycle. Some of the primary advantages are passive safety for the reactor and resistance to diversion for the heavy metal in the fuel cycle. in addition, the IFR pyroprocess recycles all the long-lived actinide activation products for casting into new fuel pins so that they may be burned in the reactor. A key component in the Fuel Cycle Facility (FCF) recycling process is the electrorefiner (ER) in which the actinides are separated from the fission products. In the process, the metal fuel is electrochemically dissolved into a high-temperature molten salt, and electrorefined uranium or uranium/plutonium products are deposited at cathodes. This report addresses the new and innovative aspects of the criticality analysis ensuing from processing metallic fuel, rather than metal oxide fuel, and from processing the spent fuel in batch operations. in particular, the criticality analysis employed a mechanistic approach as opposed to a probabilistic one. A probabilistic approach was unsuitable because of a lack of operational experience with some of the processes, rendering the estimation of accident event risk factors difficult. The criticality analysis also incorporated the uncertainties in heavy metal content attending the process items by defining normal operations envelopes (NOES) for key process parameters. The goal was to show that reasonable process uncertainties would be demonstrably safe toward criticality for continuous batch operations provided the key process parameters stayed within their NOES. Consequently the NOEs became the point of departure for accident events in the criticality analysis

  19. Proton irradiated graphite grades for a long baseline neutrino facility experiment

    International Nuclear Information System (INIS)

    Simos, N.; Nocera, P.; Zwaska, R.; Mokhov, N.

    2017-01-01

    In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ~6.1×10"2"0 p/cm"2 and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ~10"2"0 cm"-"2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the similarity in

  20. Proton irradiated graphite grades for a long baseline neutrino facility experiment

    Directory of Open Access Journals (Sweden)

    N. Simos

    2017-07-01

    Full Text Available In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF of the Deep Underground Neutrino Experiment (DUNE four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ∼6.1×10^{20}  p/cm^{2} and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ∼10^{20}  cm^{−2} where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite

  1. Proton irradiated graphite grades for a long baseline neutrino facility experiment

    Science.gov (United States)

    Simos, N.; Nocera, P.; Zhong, Z.; Zwaska, R.; Mokhov, N.; Misek, J.; Ammigan, K.; Hurh, P.; Kotsina, Z.

    2017-07-01

    In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140-180 MeV, to peak fluence of ˜6.1 ×1020 p /cm2 and irradiation temperatures between 120 - 200 °C . The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young's modulus. The proton fluence level of ˜1020 cm-2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the similarity in

  2. Criticality assessment of initial operations at the Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Ha, B.C.; Williamson, T.G.

    1993-01-01

    At the Savannah River Site (SRS), high level radioactive wastes will be immobilized into borosilicate glass for long term storage and eventual disposal. Since the waste feed streams contain uranium and plutonium, the Defense Waste Processing Facility (DWPF) process has been evaluated to ensure that a subcritical condition is maintained. It was determined that the risk of nuclear criticality in the DWPF during initial, sludge-only operations is minimal due to the dilute concentration of fissile material in the sludge combined with excess neutron absorbers

  3. Hazard analysis in uranium hexafluoride production facility

    International Nuclear Information System (INIS)

    Marin, Maristhela Passoni de Araujo

    1999-01-01

    The present work provides a method for preliminary hazard analysis of nuclear fuel cycle facilities. The proposed method identify both chemical and radiological hazards, as well as the consequences associated with accident scenarios. To illustrate the application of the method, a uranium hexafluoride production facility was selected. The main hazards are identified and the potential consequences are quantified. It was found that, although the facility handles radioactive material, the main hazards as associated with releases of toxic chemical substances such as hydrogen fluoride, anhydrous ammonia and nitric acid. It was shown that a contention bung can effectively reduce the consequences of atmospheric release of toxic materials. (author)

  4. Development laser light facility for uranium isotope separation

    International Nuclear Information System (INIS)

    Dickinson, G.J.

    1992-01-01

    A laser light facility has been built and successfully commissioned as part of a programme to explore the economic potential of Laser Isotope Separation of Uranium. The laser systems are comprised of tunable dye lasers pumped by copper vapour lasers. The requirements for optical beam stability, alignment of lasers in chains, and protection of optical coatings have made challenging demands on the engineering design and operation of the facility. (Author)

  5. Critical experiments simulating accidental water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoi, N.N.; Glushkov, L.S.

    2003-01-01

    The paper focuses on experimental analysis of nuclear criticality safety at accidental water immersion of fuel elements of the Russian TOPAZ-2 space nuclear power system reactor. The structure of water-moderated heterogeneous critical assemblies at the NARCISS facility is described in detail, including sizes, compositions, densities of materials of the main assembly components for various core configurations. Critical parameters of the assemblies measured for varying number of fuel elements, height of fuel material in fuel elements and their arrangement in the water moderator with a uniform or variable spacing are presented. It has been found from the experiments that at accidental water immersion of fuel elements involved, the minimum critical mass equal to approximately 20 kg of uranium dioxide is achieved at 31-37 fuel elements. The paper gives an example of a physical model of the water-moderated heterogeneous critical assembly with a detailed characterization of its main components that can be used for calculations using different neutronic codes, including Monte Carlo ones. (author)

  6. The strategy on rehabilitation of the former uranium facilities at the 'Pridneprovsky chemical plant' in Ukraine

    International Nuclear Information System (INIS)

    Voitsekhovich, O.; Lavrova, T.; Skalskiy, A.S.; Ryazantsev, V.F.

    2007-01-01

    This paper describes current status of the former Uranium Facilities at the Pridneprovsky Chemical Plant in Ukraine, which are currently under development of action plan for its territory rehabilitation. The monitoring data carried out during recent several years show its impact to the Environment and gives a basis for justification of the number of measures aiming to reduce radiological and ecological risks of the Uranium tailings situated at the territory of PChP. The monitoring data and strategy for its remediation are considered in the presentation. Uranium mining has been intensively conducted in Ukraine since the end of the 40-s. Most of the uranium deposits have been explored in the Dnieper river basin, while some smaller deposits can be found within the basins of the Southern Bug and Severskiy Donets rivers. There also several large Uranium Milling facilities were in operation since the end of the 40-s till 1991, when due to disintegration of the former Soviet Union system the own uranium production has been significantly declined. The Milling Plant and Uranium extraction Facilities in ZhevtiVody is still in operation with UkrAtomprom Industrial Consortium. Therefore rehabilitation programme for all Uranium facilities in this site are in duty of the East Mining Combine and the Consortium. The most difficult case is to provide rehabilitation Action Plan for Uranium tailings and number of other facilities situated in Dnieprodzerzhinsk town and which were in operation by the former State Industrial Enterprise Pridneprovskiy Chemical Plant (PChP). In past PChP was one of the largest Uranium Milling facilities of the Former Soviet Union and has been in operation since 1948 till 1991. During Soviet time the Uranium extraction at this legacy site has been carried out using the ore raw products delivered also from Central Asia, Germany and Checz Republic. After extraction the uranium residue has been putting to the nearest landscape depressions at the vicinity of

  7. Analysis of Wigner energy release process in graphite stack of shut-down uranium-graphite reactor

    OpenAIRE

    Bespala, E. V.; Pavliuk, A. O.; Kotlyarevskiy, S. G.

    2015-01-01

    Data, which finding during thermal differential analysis of sampled irradiated graphite are presented. Results of computational modeling of Winger energy release process from irradiated graphite staking are demonstrated. It's shown, that spontaneous combustion of graphite possible only in adiabatic case.

  8. Single nanocrystals of uranium dicarbide encapsulated in carbon

    International Nuclear Information System (INIS)

    Pasqualini, Enrique; Adelfang, Pablo

    1996-01-01

    The simultaneous volatilization of carbon and uranium compounds in an inert atmosphere produces the encapsulation of uranium dicarbide in graphitic material. These composite particles are of nanoscopic dimensions (smaller than 100 nanometers) and have the appearance of a black soot. Nanocapsules are as chemically inert as graphite and providers a more secure handling and processing of nuclear toxic materials. They preserve uranium dicarbide from environmental decomposition. The kernel of the nanoparticles is a single crystal of uranium dicarbide. The covers of the crystal are of graphitic structures of two types: parallel graphite layers and randomly oriented graphite crystallites. Both type of covers surround the UC 2 nanocrystal kernel. Density separation and measurements of specific area (BET technique) of purified soot were done. Particular adsorption and desorption kinetics of nitrogen monolayers indicate the presence of nanoparticles agglomeration. This is confirmed by direct observation by transmission electron microscopy (TEM). From density measurement and high resolution transmission electron microscopy (HRTEM) observation it can be inferred that there are gaps between kernel and cover and probably closed porosity inside clusters. Raman spectroscopy indicates only the presence of graphitic type carbon. Rietveld refinement of the nanoencapsulated material helps in the interpretation of X-Ray diffraction spectra. (author)

  9. Single nanocrystals of uranium dicarbide encapsulated in carbon

    Energy Technology Data Exchange (ETDEWEB)

    Pasqualini, Enrique; Adelfang, Pablo [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Dept. de Combustibles Nucleares

    1996-07-01

    The simultaneous volatilization of carbon and uranium compounds in an inert atmosphere produces the encapsulation of uranium dicarbide in graphitic material. These composite particles are of nanoscopic dimensions (smaller than 100 nanometers) and have the appearance of a black soot. Nanocapsules are as chemically inert as graphite and providers a more secure handling and processing of nuclear toxic materials. They preserve uranium dicarbide from environmental decomposition. The kernel of the nanoparticles is a single crystal of uranium dicarbide. The covers of the crystal are of graphitic structures of two types: parallel graphite layers and randomly oriented graphite crystallites. Both type of covers surround the UC{sub 2} nanocrystal kernel. Density separation and measurements of specific area (BET technique) of purified soot were done. Particular adsorption and desorption kinetics of nitrogen monolayers indicate the presence of nanoparticles agglomeration. This is confirmed by direct observation by transmission electron microscopy (TEM). From density measurement and high resolution transmission electron microscopy (HRTEM) observation it can be inferred that there are gaps between kernel and cover and probably closed porosity inside clusters. Raman spectroscopy indicates only the presence of graphitic type carbon. Rietveld refinement of the nanoencapsulated material helps in the interpretation of X-Ray diffraction spectra. (author)

  10. New Prototype Safeguards Technology Offers Improved Confidence and Automation for Uranium Enrichment Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Brim, Cornelia P.

    2013-04-01

    An important requirement for the international safeguards community is the ability to determine the enrichment level of uranium in gas centrifuge enrichment plants and nuclear fuel fabrication facilities. This is essential to ensure that countries with nuclear nonproliferation commitments, such as States Party to the Nuclear Nonproliferation Treaty, are adhering to their obligations. However, current technologies to verify the uranium enrichment level in gas centrifuge enrichment plants or nuclear fuel fabrication facilities are technically challenging and resource-intensive. NNSA’s Office of Nonproliferation and International Security (NIS) supports the development, testing, and evaluation of future systems that will strengthen and sustain U.S. safeguards and security capabilities—in this case, by automating the monitoring of uranium enrichment in the entire inventory of a fuel fabrication facility. One such system is HEVA—hybrid enrichment verification array. This prototype was developed to provide an automated, nondestructive assay verification technology for uranium hexafluoride (UF6) cylinders at enrichment plants.

  11. Conversion and Blending Facility highly enriched uranium to low enriched uranium as metal. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-05

    The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal.

  12. Conversion and Blending Facility highly enriched uranium to low enriched uranium as metal. Revision 1

    International Nuclear Information System (INIS)

    1995-01-01

    The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal

  13. The preparation of reports of a significant event at a uranium processing or uranium handling facility

    International Nuclear Information System (INIS)

    1988-08-01

    Licenses to operate uranium processing or uranium handling facilities require that certain events be reported to the Atomic Energy Control Board (AECB) and to other regulatory authorities. Reports of a significant event describe unusual events which had or could have had a significant impact on the safety of facility operations, the worker, the public or on the environment. The purpose of this guide is to suggest an acceptable method of reporting a significant event to the AECB and to describe the information that should be included. The reports of a significant event are made available to the public in accordance with the provisions of the Access to Information Act and the AECB's policy on public access to licensing information

  14. Decommissioning plan depleted uranium manufacturing facility

    International Nuclear Information System (INIS)

    Bernhardt, D.E.; Pittman, J.D.; Prewett, S.V.

    1987-01-01

    Aerojet Ordnance Tennessee, Inc. (Aerojet) is decommissioning its California depleted uranium (DU) manufacturing facility. Aerojet has conducted manufacturing and research and development activities at the facility since 1977 under a State of California Source Materials License. The decontamination is being performed by a contractor selector for technical competence through competitive bid. Since the facility will be released for uncontrolled use it will be decontaminated to levels as low as reasonably achievable (ALARA). In order to fully apply the principles of ALARA, and ensure the decontamination is in full compliance with appropriate guides, Aerojet has retained Rogers and Associaties Engineering Corporation (RAE) to assist in the decommissioning. RAE has assisted in characterizing the facility and preparing contract bid documents and technical specifications to obtain a qualified decontamination contractor. RAE will monitor the decontamination work effort to assure the contractor's performance complies with the contract specifications and the decontamination plan. The specifications require a thorough cleaning and decontamination of the facility, not just sufficient cleaning to meet the numeric cleanup criteria

  15. Review of Sodium and Plutonium related Technical Standards in Trans-Uranium Fuel Fabrication Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Misuk; Jeon, Jong Seon; Kang, Hyun Sik; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

    2016-10-15

    In this paper, we would introduce and review technical standards related to sodium fire and plutonium criticality safety. This paper may be helpful to identify considerations in the development of equipment, standards, and etc., to meet the safety requirements in the design, construction and operating of TFFF, KAPF and SFR. The feasibility and conceptual designs are being examined on related facilities, for example, TRU Fuel Fabrication Facilities (TFFF), Korea Advanced Pyro-process Facility (KAPF), and Sodium Cooled Fast Reactor (SFR), in Korea. However, the safety concerns of these facilities have been controversial in part because of the Sodium fire accident and Plutonium related radiation safety caused by transport and handling accident. Thus, many researches have been performed to ensure safety and various documents including safety requirements have been developed. In separating and reducing the long-lived radioactive transuranic(TRU) in the spent nuclear fuel, reusing as the potential energy of uranium fuel resources and reducing the high level wastes, TFFF would be receiving the attention of many people. Thus, people would wonder whether compliance with technical standards that ensures safety. For new facility design, one of the important tasks is to review of technical standards, especially for sodium and Plutonium because of water related highly reactive characteristics and criticality hazard respectively. We have introduced and reviewed two important technical standards for TFFF, which are sodium fire and plutonium criticality safety, in this paper. This paper would provide a brief guidance, about how to start and what is important, to people who are responsible for the initial design to operation of TFFF.

  16. Review of Sodium and Plutonium related Technical Standards in Trans-Uranium Fuel Fabrication Facilities

    International Nuclear Information System (INIS)

    Jang, Misuk; Jeon, Jong Seon; Kang, Hyun Sik; Kim, Seoung Rae

    2016-01-01

    In this paper, we would introduce and review technical standards related to sodium fire and plutonium criticality safety. This paper may be helpful to identify considerations in the development of equipment, standards, and etc., to meet the safety requirements in the design, construction and operating of TFFF, KAPF and SFR. The feasibility and conceptual designs are being examined on related facilities, for example, TRU Fuel Fabrication Facilities (TFFF), Korea Advanced Pyro-process Facility (KAPF), and Sodium Cooled Fast Reactor (SFR), in Korea. However, the safety concerns of these facilities have been controversial in part because of the Sodium fire accident and Plutonium related radiation safety caused by transport and handling accident. Thus, many researches have been performed to ensure safety and various documents including safety requirements have been developed. In separating and reducing the long-lived radioactive transuranic(TRU) in the spent nuclear fuel, reusing as the potential energy of uranium fuel resources and reducing the high level wastes, TFFF would be receiving the attention of many people. Thus, people would wonder whether compliance with technical standards that ensures safety. For new facility design, one of the important tasks is to review of technical standards, especially for sodium and Plutonium because of water related highly reactive characteristics and criticality hazard respectively. We have introduced and reviewed two important technical standards for TFFF, which are sodium fire and plutonium criticality safety, in this paper. This paper would provide a brief guidance, about how to start and what is important, to people who are responsible for the initial design to operation of TFFF

  17. Structural characteristics of a graphite moderated critical assembly for a Zero Power reactor at IEA (Brazil)

    International Nuclear Information System (INIS)

    Almeida Ferreira, A.C. de; Hukai, R.Y.

    1975-01-01

    The structural characteristics of a graphite moderated core of a critical assembly to be installed in the Zero Power Reactor of IEA have been defined. These characteristics are the graphite block dimensions, the number and dimensions of the holes in the graphite, the pitch, the dimensions of the sticks of fuel and graphite to be inserted in the holes, and the mechanical reproducibility of the system. The composition of the fuel and moderator sticks were also defined. The main boundary conditions were the range of the relation C/U and C/TH used in commercial HTGR and the neutronics homogeneity

  18. Operation databook of the fuel treatment system of the Static Experiment Critical Facility (STACY) and the Transient Experiment Critical Facility (TRACY). JFY 2004 to JFY 2008

    International Nuclear Information System (INIS)

    Kokusen, Junya; Sumiya, Masato; Seki, Masakazu; Kobayashi, Fuyumi; Ishii, Junichi; Umeda, Miki

    2013-02-01

    Uranyl nitrate solution fuel used in the Static Experiment Critical Facility (STACY) and the Transient Experiment Critical Facility (TRACY) is adjusted in the Fuel Treatment System, in which such parameters are varied as concentration of uranium, free nitric acid, soluble neutron poison, and so on. Operations for concentration and denitration of the solution fuel were carried out with an evaporator from JFY 2004 to JFY 2008 in order to adjust the fuel to the experimental condition of the STACY and the TRACY. In parallel, the solution fuel in which some kinds of soluble neutron poison were doped was also adjusted in JFY 2005 and JFY 2006 for the purpose of the STACY experiments to determine neutron absorption effects brought by fission products, etc. After these experiments in the STACY, a part of the solution fuel including the soluble neutron poison was purified by the solvent extraction method with mixer-settlers in JFY 2006 and JFY 2007. This report summarizes operation data of the Fuel Treatment System from JFY 2004 to JFY 2008. (author)

  19. Nuclear criticality safety controls for uranium deposits during D and D at the Oak Ridge Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    Haire, M.J.; Jordan, W.C.; Jollay, L.J. III; Dahl, T.L.

    1997-01-01

    The US Department of Energy (DOE) Deputy Assistant Secretary of Energy for Environmental Management has issued a challenge to complete DOE environmental cleanup within a decade. The response for Oak Ridge facilities is in accordance with the DOE ten-year plan which calls for completion of > 95% of environmental management work by the year 2006. This will result in a 99% risk reduction and in a significant savings in base line costs in waste management (legacy waste); remedial action (groundwater, soil, etc.); and decontamination and decommissioning (D and D). It is assumed that there will be long-term institutional control of cascade equipment, i.e., there will be no walk away from sites, and that there will be firm radioactivity release limits by 1999 for recycle metals. An integral part of these plants is the removal of uranium deposits which pose nuclear criticality safety concerns in the shut down of the Oak Ridge Gaseous Diffusion Plant. DOE has initiated the Nuclear Criticality Stabilization Program to improve nuclear criticality safety by removing the larger uranium deposits from unfavorable geometry equipment. Nondestructive assay (NDA) measurements have identified the location of these deposits. The objective of the K-25 Site Nuclear Criticality Stabilization Program is to remove and place uranium deposits into safe geometry storage containers to meet the double contingency principle. Each step of the removal process results in safer conditions where multiple controls are present. Upon completion of the Program, nuclear criticality risks will be greatly reduced

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

    International Nuclear Information System (INIS)

    Perkins, B.L.

    1983-01-01

    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 UF 6 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

  1. Environmental monitoring data review of a uranium ore processing facility in Argentina

    International Nuclear Information System (INIS)

    Bonetto, J.

    2014-01-01

    An uranium ore processing facility in the province of Mendoza (Argentina) that has produced uranium concentrate from 1954 to 1986 is currently undergoing the last steps of environmental restoration. The operator has been performing post-closure environmental monitoring since 1986, while the Nuclear Regulatory Authority (ARN) has been carrying out its own independent radiological environmental monitoring for verification purposes since its creation, in 1995. A detailed revision of ARN´s monitoring plan for uranium mining and milling facilities has been undergoing since 2013, starting with this particular site. Results obtained from long-time sampling locations (some of them currently unused) have been analyzed and potentially new sampling points have been studied and proposed. In this paper, some statistical analysis and comparison of sampling-points’ datasets are presented (specifically uranium and radium concentration in groundwater, surface water and sediments) with conclusions pertaining to their keeping or discarding as sampling points in future monitoring plans. (author)

  2. Analysis of characteristics and radiation safety situation of uranium mining and metallurgy facilities in north area of China

    International Nuclear Information System (INIS)

    Liu Ruilan; Li Jianhui; Wang Xiaoqing; Huang Mingquan

    2014-01-01

    According to the radiation safety management of uranium mining and metallurgy facilities in north area of China, features and radiation safety conditions of uranium mining and metallurgy facilities in north area of China were analyzed based on summarizing the inspection data for 2011-2013. So the main problems of radiation environment security on uranium mine were studied. The relevant management measures and recommendations were put forward, and the basis for environmental radiation safety management decision making of uranium mining and metallurgy facilities in future was provided. (authors)

  3. Critical experiments facility and criticality safety programs at JAERI

    International Nuclear Information System (INIS)

    Kobayashi, Iwao; Tachimori, Shoichi; Takeshita, Isao; Suzaki, Takenori; Miyoshi, Yoshinori; Nomura, Yasushi

    1985-10-01

    The nuclear criticality safety is becoming a key point in Japan in the safety considerations for nuclear installations outside reactors such as spent fuel reprocessing facilities, plutonium fuel fabrication facilities, large scale hot alboratories, and so on. Especially a large scale spent fuel reprocessing facility is being designed and would be constructed in near future, therefore extensive experimental studies are needed for compilation of our own technical standards and also for verification of safety in a potential criticality accident to obtain public acceptance. Japan Atomic Energy Research Institute is proceeding a construction program of a new criticality safety experimental facility where criticality data can be obtained for such solution fuels as mainly handled in a reprocessing facility and also chemical process experiments can be performed to investigate abnormal phenomena, e.g. plutonium behavior in solvent extraction process by using pulsed colums. In FY 1985 detail design of the facility will be completed and licensing review by the government would start in FY 1986. Experiments would start in FY 1990. Research subjects and main specifications of the facility are described. (author)

  4. An assessment of the effectiveness of personal visual observation for a uranium enrichment facility

    International Nuclear Information System (INIS)

    Ohno, Fubito; Okamoto, Tsuyoshi; Yokochi, Akira; Nidaira, Kazuo

    2002-01-01

    In a centrifuge uranium enrichment facility, a cascade producing low enriched uranium is composed of a large number of UF 6 gas centrifuges interconnected with pipes. If new advanced centrifuges are developed and they are installed in the facility, the number of centrifuges in the unit cascade will decrease. This means that the number of pipes connecting centrifuges will decrease also. In addition, if integrated type centrifuges containing a few tens of centrifuges are adopted for economical reasons, the number of pipes will further decrease. The smaller the number of pipes, the less the labor required to reconstruct the cascade by changing the piping arrangement so that it can produce highly enriched uranium. Because personal visual observation by inspectors is considered as one of safeguards measures against changing the piping arrangement, its effectiveness is assessed in this study. An inspection in a cascade area is modeled as a two-person non-cooperative game between an inspector and a facility operator. As a result, it is suggested that personal visual observation of the piping arrangement is worth carrying out in an advanced centrifuge uranium enrichment facility. (author)

  5. Radiological characterization of two Spanish uranium mine facilities

    International Nuclear Information System (INIS)

    Quindos Poncela, L.S.; Fernandez, P.I.; Gomez Arozamena, J.; Bordonoba, M.L.

    2000-01-01

    During the last decade our Department of Applied and Medical Physics has been involved in the development of a Radiation Protection Programme to monitor and control the environmental radiation conditions existing in the only two Spanish uranium mill facilities located at La Haba (Badajoz) and Salacious (Salamanca). Both mines are located in the west of the country, geographical area with high natural radiation levels. In the framework of this Programme, measurements of radon, radon progeny and external gamma radiation indoors and outdoors, as well as of radon exhalation rate and natural radionuclide concentrations in tailings and soils, have been systematically carried out. In particular, two ore body areas in these uranium mill sites have been specially studied to determine the natural radiation background to be used as a reference value to design reliable criteria for the closure of both facilities, which is planned for the next year. This paper summarizes the main results obtained from the measurements of external gamma radiation, radon concentrations, radon exhalation rate and natural radionuclide activity concentrations made in the above mentioned facilities with special emphasis on the results achieved from the ore body areas. Correlations between experimental results and a short description of the devices and methods used in the measurements are also shown. (author)

  6. Installation places of criticality accident detectors in the plutonium conversion development facility

    International Nuclear Information System (INIS)

    Sanada, Yukihisa; Tsujimura, Norio; Shimizu, Yoshio; Izaki, Kenji; Furuta, Sadaaki

    2008-01-01

    At the Plutonium Conversion Development Facility (PCDF) in the Nuclear Fuel Cycle Engineering Laboratories, the co-conversion technologies to purify the mixed plutonium and uranium nitrate solution discharged from a reprocessing plant have been developed. The probability of a criticality accident in PCDF is extremely low. However, the criticality accident alarm system (CAAS) has been in place since 1982 to reduce the radiation dose to workers in case of such a rare criticality accident. The CAAS contains criticality accident detector units (CADs), one unit consisting of three plastic scintillation detectors, and using the 2 out of 3 voting system for the purpose of high reliability. Currently, eight CADs are installed in PCDF evaluating the dose using a simple equation allowing for a safety margin. The purpose of this study is to show the determination procedures for the adequate relocation of the CADs which adequately ensures safety in PCDF. (author)

  7. Stabilization of uranium hexafluoride by hydrolysis method for decommissioning of safeguard laboratory facility

    Energy Technology Data Exchange (ETDEWEB)

    Inagawa, Jun; Hotoku, Shinobu; Oda, Tetsuzo; Aoyagi, Noboru; Magara, Masaaki [Japan Atomic Energy Agency, Nuclear Science and Engineering Directorate, Tokai, Ibaraki (Japan)

    2014-03-15

    In safeguard laboratory (SGL) facility of Nuclear Science Research Institute of JAEA , uranium hexafluoride (UF{sub 6}) of enriched uranium of various enrichment was used for research and development of a spectrometric method for the determination of the enrichment of uranium in April 1983 through March 1993. After completion of this R and D, the UF{sub 6} has been stored in SGL facility. It was decided that the UF{sub 6} is carried to out of the facility, because the SGL facility will be decommissioning until March 2015. To transport and store in safety after transportation, it is necessary that the UF{sub 6} should be converted to stable chemical form. Hydrolysis of UF{sub 6} to uranyl fluoride (UO{sub 2}F{sub 2}) and evaporation to solid state were selected for the stabilization method. The equipment for hydrolysis and evaporation was installed in the SGL facility. Stabilization was operated in this equipment, and all of the UF{sub 6} in the SGL facility was converted to UO{sub 2}F{sub 2} solid state in October 2012 through August 2013. In this report, results of examination and operation for stabilization of UF{sub 6} were reported. (author)

  8. Development of ISA procedure for uranium fuel fabrication and enrichment facilities

    International Nuclear Information System (INIS)

    Yamate, Kazuki; Arakawa, Tomoyuki; Yamashita, Masahiro; Sasaki, Noriaki; Hirano, Mitsumasa

    2011-01-01

    The integrated safety analysis (ISA) procedure has been developed to apply risk-informed regulation to uranium fuel fabrication and enrichment facilities. The major development efforts are as follows: (a) preparing the risk level matrix as an index for items-relied-on-for-safety (IROFS) identification, (b) defining requirements of IROFS, and (c) determining methods of IROFS importance based on the results of risk- and scenario-based analyses. For the risk level matrix, the consequence and likelihood categories have been defined by taking into account the Japanese regulatory laws, rules, and safety standards. The trial analyses using the developed procedure have been performed for several representative processes of the reference uranium fuel fabrication and enrichment facilities. This paper presents the results of the ISA for the sintering process of the reference fabrication facility. The results of the trial analyses have demonstrated the applicability of the procedure to the risk-informed regulation of these facilities. (author)

  9. Criticality concerns in cleaning large uranium hexafluoride cylinders

    International Nuclear Information System (INIS)

    Sheaffer, M.K.; Keeton, S.C.; Lutz, H.F.

    1995-06-01

    Cleaning large cylinders used to transport low-enriched uranium hexafluoride (UF 6 ) presents several challenges to nuclear criticality safety. This paper presents a brief overview of the cleaning process, the criticality controls typically employed and their bases. Potential shortfalls in implementing these controls are highlighted, and a simple example to illustrate the difficulties in complying with the Double Contingency Principle is discussed. Finally, a summary of recommended criticality controls for large cylinder cleaning operations is presented

  10. Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

    2015-03-01

    In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.

  11. On the search for uranium

    International Nuclear Information System (INIS)

    Forland, A.

    1987-01-01

    The research reactor JEEP, which was completed in 1951 at Institutt for atomenergi (IFA), Kjeller, Norway, became the first reactor in the world to be built outside the big-power states. Due to Norwegian production of heavy water, the reactor was constructed as a heavy water reactor using natural uranium as fuel. A graphite reflector surrounded the reactor tank. Both uranium and graphite had to be purchased abroad. Because of the Anglo-American monopoly of all sizable uranium sources in the Western part of the world, no uranium for the reactor was available on the free market. The present study analyses Norway's and IFA's foreign relations at the time of the reactor project, and focuses in particular on the choice of the future partner that IFA had to make in order to solve its uranium problem. Political considerations were among the factors behind the decision in 1951 to establish a joint Dutch-Norwegian atomic energy research institute

  12. Obtention of uranium-molybdenum alloy ingots technique to avoid carbon contamination

    Energy Technology Data Exchange (ETDEWEB)

    Pedrosa, Tercio A.; Paula, Joao Bosco de; Reis, Sergio C.; Brina, Jose Giovanni M.; Faeda, Kelly Cristina M.; Ferraz, Wilmar B., E-mail: tap@cdtn.b, E-mail: jbp@cdtn.b, E-mail: jgmb@cdtn.b, E-mail: ferrazw@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The replacement of high enriched uranium (U{sup 235} > 85 wt%) by low enriched uranium (U{sup 235} < 20wt%) nuclear fuels in research and test reactors is being implemented as an initiative of the Reduced Enrichment for Research and Test Reactors (RERTR) program, conceived in the USA since mid-70s, in order to avoid nuclear weapons proliferation. Such replacement implies in the use of compounds or alloys with higher uranium densities. Among the several uranium alloys investigated since then, U-Mo presents great application potential due to its physical properties and good behavior during irradiation, which makes it an important option as a nuclear fuel material for the Brazilian Multipurpose Reactor - RMB. The development of the plate-type nuclear fuel based on U-Mo alloy is being performed at the Nuclear Technology Development Centre (CDTN) and also at IPEN. The carbon contamination of the alloy is one of the great concerns during the melting process. It was observed that U-Mo alloy is more critical considering carbon contamination when using graphite crucibles. Alternative melting technique was implemented at CDTN in order to avoid carbon contamination from graphite crucible using Yttria stabilized ZrO{sub 2} crucibles. Ingots with low carbon content and good internal quality were obtained. (author)

  13. Mineralogical study of uraniferous graphitic ore from Deogpyeong, Mogso and southern part of Daejeon area

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D J; Nam, S K [Korean Inst. of Energy and Resources, Seoul (Republic of Korea)

    1981-11-01

    Uranium minerals of torbernite, metatorbernite, metatyuyamunite and autunite have been identified from the uraniferous ores in graphitic beds of Ogcheon Group in Deogpyeong, Mogso and southern part of Daejeon area. Polarizing and ore microscopic studies, and chemical and X-ray powder diffraction analyses were carried out on the uraniferous graphite and associated materials. Main component minerals of uraniferous samples are graphite and quartz. Minor minerals are calcite, muscovite, sericite, andalusite, barite, kaolinite, hyaline opal, uranium minerals, sulfides such as pyrite, chalcopyrite, zincblende, and pyrrhotite, limonite, zeolite minerals such as laumontite and heulandite. Metatyuyamunite, torbernite, metatorbernite and autunite generally occur together with secondary minerals such as kaolinite, hyaline opal, calcite and limonite. They were found along the minor fissures or on the surface. Secondary uranium minerals described above were formed by supergenetic origin from primary uranium mineral. Uraniferous phosphatic nodule from Deogpyeong area are mainly composed of graphite and fluorapatite. And minor minerals are barite, quartz, muscovite and pyrite. Autoradiograph from uraniferous nodule shows that uranium enrichment in outer part of nodules is much higher than in inner part. This feature coincides with chemical analyses data of this uraniferous nodule.

  14. Critical experiments on low-enriched uranium oxide system with H/U=1.25

    International Nuclear Information System (INIS)

    Oh, I.; Rothe, R.E.; Tuck, G.

    1982-01-01

    Fifteen (15) critical experiments were performed on a horizontal split table machine using 4.48%-enriched sup(235)U uranium oxide(U 3 O 8 ). The oxide was compacted to a density of 4.68g/cm 3 and placed in 152 mm cubical aluminum cans. Water was added to achive an H/U of 1.25. Various arrays of oxide cans were distributed on each half of the split table, and the separation between halves reduced until criticality occurred. The critical table separation varied from 3.59 mm to 18.40 mm. Twelve (12) experiments required the addition of a high-enriched(-93 %sup(235)U) metal or solution driver to achieve criticality. These experiments were performed in a plastic, concrete, or thin steel reflector. Three additional experiments in the plastic reflector contained either 9.3-mm- or 24.3-mm-thick plastic moderator material between the oxide cans and did not require a driver to achieve criticality. Critical uranium driver masses ranged from 9.999 kg to 14.000 kg (solution driver), and from 25.378 kg to 29.278 kg (metal driver) for 5X5X5 arrays of uranium oxide cans. Always, one or four of these 125 cans had to be removed to make room for the drivers. Therefore, the uranium oxide masses used were 1823.8 kg and 1863.5 kg. For the moderated experiments, the uranium oxide mass ranged between 574.4 kg and 1210.0 kg. (Author)

  15. Uranium Processing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — An integral part of Y‑12's transformation efforts and a key component of the National Nuclear Security Administration's Uranium Center of Excellence, the Uranium...

  16. Proposal for Monitoring Within the Centrifuge Cascades of Uranium Enrichment Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Farrar, David R.

    2017-04-01

    Safeguards are technical measures implemented by the International Atomic Energy Agency (IAEA) to independently verify that nuclear material is not diverted from peaceful purposes to weapons (IAEA, 2017a). Safeguards implemented at uranium enrichment facilities (facilities hereafter) include enrichment monitors (IAEA, 2011). Figure 1 shows a diagram of how a facility could be monitored. The use of a system for monitoring within centrifuge cascades is proposed.

  17. The measurements of critical mass with uranium fuel elements and thorium rods

    International Nuclear Information System (INIS)

    Yao Zhiquan; Chen Zhicheng; Yao Zewu; Ji Huaxiang; Bao Borong; Zhang Jiahua

    1991-01-01

    The critical experiments with uranium elements and Thorium rods have been performed in zero power reactor at Shanghai Institute of Nuclear Research. The critical masses have been measured in various U/Th ratios. The fuels are 3% 235 U-enriched uranium. The Thorium rods are made from power of ThF 4 . Ratios of calculated values to experimental values are nearly constant at 0.995

  18. The use of graphite for the reduction of void reactivity in CANDU reactors

    International Nuclear Information System (INIS)

    Min, B.J.; Kim, B.G.; Sim, K-S.

    1995-01-01

    Coolant void reactivity can be reduced by using burnable poison in CANDU reactors. The use of graphite in the fuel bundle is introduced to reduce coolant void reactivity by adding an appropriate amount of burnable poison in the central rod. This study shows that sufficiently low void reactivity which in controllable by Reactor Regulating System (RRS) can be achieved by using graphite used fuel with slightly enriched uranium. Zero void reactivity can be also obtained by using graphite used fuel with a large central rod. A new fuel bundle with graphite rods can substantially reduce the void reactivity with less burnup penalty compared to previously proposed low void reactivity fuel with depleted uranium. (author)

  19. Analysis of RA-8 critical facility core in some configurations; Analisis del nucleo de la facilidad critica RA-8 en distintas configuraciones

    Energy Technology Data Exchange (ETDEWEB)

    Abbate, Maximo J [Instituto de Investigaciones Cientificas y Tecnicas de las Fuerzas Armadas (CITEFA), Villa Martelli (Argentina); Sbaffoni, Maria M [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Gerencia de Tecnologia

    2000-07-01

    The RA-8 critical facility was designated and built to be used in the experimental plan of the 'CAREM' Project but is, in itself, very versatile and adequate to perform many types of other experiments. The present paper includes calculated estimates of some critical configurations and comparisons with experimental results obtained during its start up. Results for Core 1 with homogeneous arrangement of rods containing 1.8 % enriched uranium, showed very good agreement. In fact, an experimentally critical configuration was reached with 1.300 rods and calculated values were: 1.310 using the WIMS code and 1.148 from the CONDOR code. Moreover, it was verified that the estimated number of 3.4% enriched uranium rods to be fabricated is enough to build a heterogeneous core or even a homogeneous core with this enrichment. The replacement of 3.4 % enriched uranium by 3.6 % will not present problems related with the original plan. (author)

  20. Nondestructive assay of special nuclear material for uranium fuel-fabrication facilities

    International Nuclear Information System (INIS)

    Smith, H.A. Jr.; Schillebeeckx, P.

    1997-01-01

    A high-quality materials accounting system and effective international inspections in uranium fuel-fabrication facilities depend heavily upon accurate nondestructive assay measurements of the facility's nuclear materials. While item accounting can monitor a large portion of the facility inventory (fuel rods, assemblies, storage items), the contents of all such items and mass values for all bulk materials must be based on quantitative measurements. Weight measurements, combined with destructive analysis of process samples, can provide highly accurate quantitative information on well-characterized and uniform product materials. However, to cover the full range of process materials and to provide timely accountancy data on hard-to-measure items and rapid verification of previous measurements, radiation-based nondestructive assay (NDA) techniques play an important role. NDA for uranium fuel fabrication facilities relies on passive gamma spectroscopy for enrichment and U isotope mass values of medium-to-low-density samples and holdup deposits; it relies on active neutron techniques for U-235 mass values of high-density and heterogeneous samples. This paper will describe the basic radiation-based nondestructive assay techniques used to perform these measurements. The authors will also discuss the NDA measurement applications for international inspections of European fuel-fabrication facilities

  1. Measurement of reactivity worths of burnable poison rods in enriched uranium graphite-moderated core simulated to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi; Takeuchi, Motoyoshi; Kitadate, Kenji; Yoshifuji, Hisashi; Kaneko, Yoshihiko

    1980-11-01

    As the core design for the Experimental Very High Temperature Gas Cooled Reactor progresses, evaluation of design precision has become increasingly important. For a high precision design, it is required to have adequate group constants based on accurate nuclear data, as well as calculation methods properly describing the physical behavior of neutrons. We, therefore, assembled a simulation core for VHTR, SHE-14, using a graphite-moderated 20%-enriched uranium Semi-Homogeneous Experimental Critical Facility (SHE), and obtained useful experimental data in evaluating the design precision. The VHTR is designed to accommodate burnable poison and control rods for reactivity compensation. Accordingly, the experimental burnable poison rods which are similar to those to be used in the experimental reactor were prepared, and their reactivity values were measured in the SHE-14 core. One to three rods of the above experimental burnable poison rods were inserted into the central column of the SHE-14 core, and the reactivity values were measured by the period and fuel rod substitution method. The results of the measurements have clearly shown that due to the self-shielding effect of B 4 C particles the reactivity value decreases with increasing particle diameter. For the particle diameter, the reactivity value is found to increase linearly with the logarithm of boron content. The measured values and those calculated are found to agree with each other within 5%. These results indicate that the reactivity of the burnable poison rod can be estimated fairly accurately by taking into account the self-shielding effect of B 4 C particles and the heterogeneity of the lattice cell. (author)

  2. Uranium Metal to Oxide Conversion by Air Oxidation –Process Development

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, A

    2001-12-31

    Published technical information for the process of metal-to-oxide conversion of uranium components has been reviewed and summarized for the purpose of supporting critical decisions for new processes and facilities for the Y-12 National Security Complex. The science of uranium oxidation under low, intermediate, and high temperature conditions is reviewed. A process and system concept is outlined and process parameters identified for uranium oxide production rates. Recommendations for additional investigations to support a conceptual design of a new facility are outlined.

  3. Organic matter and containment of uranium and fissiogenic isotopes at the Oklo natural reactors

    International Nuclear Information System (INIS)

    Nagy, B.; Rigali, M.J.; Davis, D.W.; Parnell, J.

    1991-01-01

    Some of the Precambrian natural fission reactors at Oklo in Gabon contain abundant organic matter, part of which was liquefied at the time of criticality and subsequently converted to a graphitic solid. The liquid organic matter helps to reduce U(VI) to U(IV) from aqueous solutions, resulting in the precipitation of uraninite. It is known that in the prevailing reactor environments, precipitated uraninite grains incorporated fission products. We report here observations which show that these uraninite crystals were held immobile within the re-solidified, graphitic bituminous organics at Oklo thus enhanced radionuclide containment. Uraninite encased in solid graphitic matter in the organic-rich reactor zones lost virtually no fissiogenic lanthanide isotopes. The first major episode of uranium and lead migration was caused by the intrusion of a swarm of adjacent dolerite dykes about 1,100 Myr after the reactors went critical. Our results from Oklo imply that the use of organic, hydrophobic solids such as graphitic bitumen as a means of immobilizing radionuclides in pre-treated nuclear waste warrants further investigation. (author)

  4. Study on graphite samples for nuclear usage

    International Nuclear Information System (INIS)

    Suarez, J.C.M.; Silva Roseira, M. da

    1994-01-01

    Available as short communication only. The graphite, due to its properties (mechanical strength, thermal conductivity, high-temperature stability, machinability etc.) have many industrial applications, and consequently, an important strategic value. In the nuclear area, it has been used as moderator and reflector of neutrons in the fission process of uranium. The graphite can be produced from many types of carbonaceous materials by a variety of process dominated by the manufactures. This is the reason why there are in the world market a lot of graphite types with different physical and mechanical properties. The present investigation studies some physical characteristics of the graphite samples destined to use in a nuclear reactor. (author). 8 refs, 1 fig, 1 tab

  5. ZPR-3 Assembly 11: A cylindrical sssembly of highly enriched uranium and depleted uranium with an average 235U enrichment of 12 atom % and a depleted uranium reflector

    International Nuclear Information System (INIS)

    Lell, R.M.; McKnight, R.D.; Tsiboulia, A.; Rozhikhin, Y.

    2010-01-01

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was 235 U or 239 Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core 235 U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group (CSEWG) Benchmark

  6. Recent developments concerning French fuel elements used in natural uranium - graphite - CO{sub 2} reactor systems; Developpements recents des elements combustibles francais de la filiere uranium naturel - graphite - CO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Salesse, M; Stohr, J A; Jeanpierre, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The policy followed in France for the development of fuel elements for reactors belonging to the Electricite de France has been to benefit as much as possible, for each new pile from the most recent technical progress by developing in each case a fuel element allowing the maximum power per channel. The two latest fuel elements thus studied by the French Atomic Energy Commission are of two different types: a tubular uranium element closed at both ends and cooled externally. (This type of element, chosen for the reactors EDF 2, EDF 3 and EDF 4 makes it possible to attain maximum specific powers of the order of 6 MW/metric ton.); an open tubular uranium element cooled both internally and externally, called an annular element which in being studied as a possibility for EDF 5. Such an element makes it possible to attain specific powers of over 12 MW/metric ton. The two types of element have the following common characteristics: - the can, for external cooling, has herning-bone type fins. This type of profile which has been vastly improved recently thereby increasing its thermal efficiency, has the important advantage of avoiding vibration of the element, but has posed problems of resistance to thermal cycling necessitating much research. - the fuel rods are placed inside graphite jackets, this limiting the vertical forces to which they are subjected and protecting them during charging and discharging. On the other hand, these elements present very different problems as for as the following points are concerned: - the characteristics required of the uranium tubes apart of course from a good dimensional stability during irradiation in the two cases are in the case of the closed tubes a very high resistance to external pressure, and in the case of the annular elements a low neutron absorption. Thus for each of these two cases it has been necessary to develop a suitable type of alloy. - a possible loosening of the can during thermal cycling, which is peculiar to the

  7. Procedure of uranium mine and mill facilities decommissioning work

    International Nuclear Information System (INIS)

    Li Renjie

    1995-01-01

    The procedure of decommissioning work of uranium mine and mill facilities includes three stages: preparation, on-the-spot construction and acceptance after being completed. The first stage, preparation, is discussed in detail, and it is presented to take the measures of strengthening leadership and improving leading body to conduct the decommissioning work best

  8. A study on the formation of uranium carbide in an induction furnace

    International Nuclear Information System (INIS)

    Song, In Young; Lee, Yoon Sang; Kim, Eung Soo; Lee, Don Bae; Kim, Chang Kyu

    2005-01-01

    Uranium is a typical carbide-forming element. Three carbides, UC, U 2 C 3 and UC 2 , are formed in the uranium-carbon system. The most important of these as fuel is uranium monocarbide UC. It is well known that Uranium carbides can be obtained by three basic methods: 1) by reaction of uranium metal with carbon; 2) by reaction of uranium metal powder with gaseous hydrocarbons; 3) by reaction of uranium oxides with carbon. The use of uranium monocarbide, or materials based on it, has great prospects as fuel for nuclear reactors. It is quite possible that uranium dicarbide UC 2 may also acquire great importance as a fuel, particularly in dispersion fuel elements with graphite matrix. In the present study, uranium carbides are obtained by direct reaction of uranium metal with graphite in a high frequency induction furnace

  9. ZPR-3 Assembly 11 : A cylindrical sssembly of highly enriched uranium and depleted uranium with an average {sup 235}U enrichment of 12 atom % and a depleted uranium reflector.

    Energy Technology Data Exchange (ETDEWEB)

    Lell, R. M.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; National Security; Inst. of Physics and Power Engineering

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation

  10. Criticality safety aspects of K-25 Building uranium deposit removal

    International Nuclear Information System (INIS)

    Haire, M.J.; Jordan, W.C.; Ingram, J.C. III; Stinnet, E.C. Jr.

    1995-01-01

    The K-25 Building of the Oak Ridge Gaseous Diffusion Plant (now the K-25 Site) went into operation during World War II as the first large scale production plant to separate 235 U from uranium by the gaseous diffusion process. It operated successfully until 1964, when it was placed in a stand-by mode. The Department of Energy has initiated a decontamination and decommissioning program. The primary objective of the Deposit Removal (DR) Project is to improve the nuclear criticality safety of the K-25 Building by removing enriched uranium deposits from unfavorable-geometry process equipment to below minimum critical mass. The method utilized to accomplish this are detailed in this report

  11. Criticality safety aspects of K-25 Building uranium deposit removal

    Energy Technology Data Exchange (ETDEWEB)

    Haire, M.J.; Jordan, W.C. [Oak Ridge National Lab., TN (United States); Ingram, J.C. III; Stinnet, E.C. Jr. [Oak Ridge K-25 Site, TN (United States)

    1995-12-31

    The K-25 Building of the Oak Ridge Gaseous Diffusion Plant (now the K-25 Site) went into operation during World War II as the first large scale production plant to separate {sup 235}U from uranium by the gaseous diffusion process. It operated successfully until 1964, when it was placed in a stand-by mode. The Department of Energy has initiated a decontamination and decommissioning program. The primary objective of the Deposit Removal (DR) Project is to improve the nuclear criticality safety of the K-25 Building by removing enriched uranium deposits from unfavorable-geometry process equipment to below minimum critical mass. The method utilized to accomplish this are detailed in this report.

  12. Monte Carlo criticality analysis of simple geometries containing tungsten-rhenium alloys engrained with uranium dioxide and uranium mononitride

    International Nuclear Information System (INIS)

    Webb, Jonathan A.; Charit, Indrajit

    2011-01-01

    Highlights: → The addition of rhenium to the tungsten matrix within W-UO 2 and W-UN CERMET materials can help reduce the risk of submersion criticality accidents while increasing the strength and ductility of tungsten based nuclear fuel elements. → The addition of rhenium up to 30 at.% to simple geometries containing W-UO 2 mixtures can increase the critical mass by 65 kg. → The addition of rhenium up to 30 at.% to simple geometries containing W-UN mixtures can increase the critical mass by 22 kg. → The addition of rhenium by up to 30 at.% to simple geometries containing W-UO 2 mixtures can reduce the change in reactivity change due to water submersion by $5.07. → The addition of rhenium by up to 30 at.% to simple geometries containing W-UN mixtures can reduce the change in reactivity due to water submersion by $3.24. - Abstract: The critical mass and dimensions of simple geometries containing highly enriched uranium dioxide (UO 2 ) and uranium mononitride (UN) encapsulated in tungsten-rhenium alloys are determined using MCNP5 criticality calculations. Spheres as well as cylinders with length to radius ratios of 1.82 are computationally built to consist of 60 vol.% fuel and 40 vol.% metal matrix. Within the geometries, the uranium is enriched to 93 wt.% uranium-235 and the rhenium content within the metal alloy was modeled over the range of 0-30 at.%. The spheres containing UO 2 were determined to have a critical radius of 18.29-19.11 cm and a critical mass ranging from 366 kg to 424 kg. The cylinders containing UO 2 were found to have a critical radius ranging from 17.07 cm to 17.84 cm with a corresponding critical mass of 406-471 kg. Spheres engrained with UN were determined to have a critical radius ranging from 14.82 cm to 15.19 cm and a critical mass between 222 kg and 242 kg. Cylinders which were engrained with UN were determined to have a critical radius ranging from 13.81 cm to 14.15 cm and a corresponding critical mass of 245-267 kg. The critical

  13. Uranium absorption study pile; Empilement pour le controle de l'absorption de l'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Raievski, V; Sautiez, B [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    The report describes a pile designed to measure the absorption of fuel slugs. The pile is of graphite and comprises a central section composed of uranium rods in a regular lattice. RaBe sources and BF{sub 3} counters are situated on either side of the center. A given uranium charge is compared with a specimen charge of about 560 kg, and the difference in absorption between the two noted. The sensitivity of the equipment will detect absorption variations of about a few ppm boron (10{sup -6} boron per gr. of uranium) or better. (author) [French] Nous decrivons un dispositif permettant de mesurer l'absorption des elements combustibles d'une pile. Ce dispositif est constitue par un empilement de graphite dont la region centrale est formee par un reseau regulier de barres d'uranium. Des sources de RaBe et des compteurs a BF{sub 3} sont places de part et d'autre de cette region. En comparant un chargement d'uranium a un chargement etalon d'environ 560 kg, on peut determiner la difference d'absorption entre ces deux chargements. La sensibilite permettrait de deceler une variation d'absorption de l'ordre du ppm de bore (10{sup -6} g de bore par gramme d'uranium) et peut-etre mieux. (auteur)

  14. Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.; Turner, J.C.

    1992-12-01

    A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO{sub 2}F{sub 2} and H{sub 2}O) and hydrofluoric-acid-moderated uranium hexaflouride (UF{sub 6} and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % {sup 235}U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

  15. Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.; Turner, J.C.

    1992-12-01

    A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO[sub 2]F[sub 2] and H[sub 2]O) and hydrofluoric-acid-moderated uranium hexaflouride (UF[sub 6] and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % [sup 235]U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

  16. Lessons learned from recent safety related incidents at A Canadian uranium conversion facility

    International Nuclear Information System (INIS)

    Jaferi, Jafir

    2013-01-01

    This paper presents the Canadian Nuclear Safety Commission's (CNSC) regulatory requirements for nuclear fuel facility licensees to report any situation or incident that results or is likely to result in a hazard to the health or safety of any person or the environment and to submit its incident investigation report with cause(s) of the incident and corrective actions taken or planned. In addition, the paper presents two recent safety-related incidents that occurred at a uranium conversion facility in Canada along with their consequences, causes, corrective actions and any lessons learned. The first incident resulted in a release of uranium hexafluoride (UF6) inside the UF6 cylinder filling station and the second one resulted in a spill of uranium tetrafluoride (UF 4 ) slurry inside the UF6 plant. Both incidents had no impact on the workers or the environment. (authors)

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

    International Nuclear Information System (INIS)

    LADURELLE, Laurent; LISBONNE, Pierre

    2003-01-01

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

  18. 77 FR 18272 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-03-27

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC, National Enrichment Facility, Eunice... Louisiana Energy Services (LES), LLC, National enrichment Facility in Eunice, New Mexico, and has verified...

  19. 77 FR 65729 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-10-30

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC, National Enrichment Facility, Eunice... Services (LES), LLC, National Enrichment Facility in Eunice, New Mexico, and has verified that cascades...

  20. Criticality accident of nuclear fuel facility. Think back on JCO criticality accident

    International Nuclear Information System (INIS)

    Naito, Keiji

    2003-09-01

    This book is written in order to understand the fundamental knowledge of criticality safety or criticality accident of nuclear fuel facility by the citizens. It consists of four chapters such as critical conditions and criticality accident of nuclear facility, risk of criticality accident, prevention of criticality accident and a measure at an occurrence of criticality accident. A definition of criticality, control of critical conditions, an aspect of accident, a rate of incident, damage, three sufferers, safety control method of criticality, engineering and administrative control, safety design of criticality, investigation of failure of safety control of JCO criticality accident, safety culture are explained. JCO criticality accident was caused with intention of disregarding regulation. It is important that we recognize the correct risk of criticality accident of nuclear fuel facility and prevent disasters. On the basis of them, we should establish safety culture. (S.Y.)

  1. Uranium Mining and Nuclear Facilities (Prohibitions) Act 1986 No. 194

    International Nuclear Information System (INIS)

    1986-01-01

    The purpose of this Act is to protect the health and safety of the people of New South Wales and its environment. Accordingly it prohibits prospecting or mining for uranium and the construction and operation of nuclear reactors and other facilities in the nuclear fuel cycle. (NEA) [fr

  2. The challenge of uranium exploration

    International Nuclear Information System (INIS)

    Fountain, D.K.

    1982-06-01

    The first uranium discoveries at Beaverlodge were made using simple radiometric methods: hand-held geiger counters. Since then techniques of uranium exploration have evolved through airborne radiometric surveys, tracking glacial boulder trains to their origins, and electromagnetic surveys to detect graphite associated with buried uranium deposits. Simple radiometric surveys can cost around $1 000. per day, while testing for deposits at depths of over 400 meters will cost more than $60 000. per drill hole

  3. A study of the material accountancy procedure at the uranium enrichment facility

    International Nuclear Information System (INIS)

    Shirahashi, J.; Akiba, M.; Omae, M.

    1984-01-01

    This paper describes an evaluation of material accountancy based on total uranium (U element MUF) to detect diversions of significant quantity in the uranium enrichment facility operating at a stated maximum enrichment level of 5%. Verification that material production is within the declared enrichment can be achieved by the inspection activities associated with limited - frequency unannounced access (LFUA) to cascade areas as treated extensively in HSP. According to the experience of the material accountancy at our facility, the reduction of the material accountancy capability by changing from U-235 isotope MUF to U element MUF is only about half. However, still the U element MUF approach can meet the current IAEA detection goals for the up to about 1000 tswu/a plant

  4. The National Criticality Experiments Research Center at the Device Assembly Facility, Nevada National Security Site: Status and Capabilities, Summary Report

    International Nuclear Information System (INIS)

    Bragg-Sitton, S.; Bess, J.; Werner, J.

    2011-01-01

    The National Criticality Experiments Research Center (NCERC) was officially opened on August 29, 2011. Located within the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS), the NCERC has become a consolidation facility within the United States for critical configuration testing, particularly those involving highly enriched uranium (HEU). The DAF is a Department of Energy (DOE) owned facility that is operated by the National Nuclear Security Agency/Nevada Site Office (NNSA/NSO). User laboratories include the Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL). Personnel bring their home lab qualifications and procedures with them to the DAF, such that non-site specific training need not be repeated to conduct work at DAF. The NNSS Management and Operating contractor is National Security Technologies, LLC (NSTec) and the NNSS Safeguards and Security contractor is Wackenhut Services. The complete report provides an overview and status of the available laboratories and test bays at NCERC, available test materials and test support configurations, and test requirements and limitations for performing sub-critical and critical tests. The current summary provides a brief summary of the facility status and the method by which experiments may be introduced to NCERC.

  5. The Amster concept: a configuration generating its own uranium with a mixed thorium and uranium support

    International Nuclear Information System (INIS)

    Vergnes, J.; Garzenne, C.; Lecarpentier, D.; Mouney, H.; Delpech, M.

    2001-01-01

    AMSTER is a continuously reloaded, graphite-moderated molten salt critical reactor, using a 238 U or 232 Th fuel support, slightly enriched with 235 U if necessary. Using this concept, one can define a large number of configurations according to the products loaded and recycled. The choice of thorium fuel support leads to two configurations requiring no additional 235 U as fissile material: a configuration with one moderating zone, incinerating Transuranium elements (TRU); a configuration with 2 moderating zones self-consuming TRU and regenerating the fissile uranium ( 233 U). In this configuration, it is even possible to burn 238 U (from depleted uranium) by adding it to the thorium support. These configurations use a minimum amount of fuel (100 kg of 232 Th or 100 kg of a 232 Th- 238 U mix per TWh) and produce very little TRU (a few tens of grams per TWh). (author)

  6. Criticality safety concerns of uranium deposits in cascade equipment

    International Nuclear Information System (INIS)

    Plaster, M.J.

    1996-01-01

    The Paducah and Portsmouth Gaseous Diffusion Plants enrich uranium in the 235 U isotope by diffusing gaseous uranium hexafluoride (UF 6 ) through a porous barrier. The UF 6 gaseous diffusion cascade utilized several thousand open-quotes stagesclose quotes of barrier to produce highly enriched uranium (HEU). Historically, Portsmouth has enriched the Paducah Gaseous Diffusion Plant's product (typically 1.8 wt% 235 U) as well as natural enrichment feed stock up to 97 wt%. Due to the chemical reactivity of UF 6 , particularly with water, the formation of solid uranium deposits occur at a gaseous diffusion plant. Much of the equipment operates below atmospheric pressure, and deposits are formed when atmospheric air enters the cascade. Deposits may also be formed from UF 6 reactions with oil, UF 6 reactions with the metallic surfaces of equipment, and desublimation of UF 6 . The major deposits form as a result of moist air in leakage due to failure of compressor casing flanges, blow-off plates, seals, expansion joint convolutions, and instrument lines. This report describes criticality concerns and deposit disposition

  7. Material control and accounting requirements for uranium enrichment facilities

    International Nuclear Information System (INIS)

    Ting, P.

    1991-01-01

    This paper reports that the U.S. Nuclear Regulatory Commission has defined material control and accounting (MC and A) requirement for low-enriched uranium enrichment plants licensed under 10 CFR parts 40 and 70. Following detailed assessment of potential safeguards issues relevant to these facilities, a new MC and A rule was developed. The primary safeguards considerations are detection of the loss of special nuclear material, detection of clandestine production of special nuclear material of low strategic significance for unauthorized use or distribution, and detection of unauthorized production of uranium enriched to ≥10 wt % U-235. The primary safeguards concerns identified were the large absolute limit of error associated with the material balance closing, the inability to shutdown some uranium enrichment technologies to perform a cleanout inventory of the process system, and the flexibility of some of these technologies to produce higher enrichments. Unauthorized production scenarios were identified for some technologies that could circumvent the detection of the production and removal of 5 kilograms of U-235 as high-enriched uranium through conventional material control and accounting programs. Safeguards techniques, including the use of production and process control information, measurements, and technical surveillance, were identified to compensate for these concerns

  8. Audit of the Uranium Solidification Facility at the Savannah River Site

    International Nuclear Information System (INIS)

    1994-01-01

    In the late 1980s, DOE decided to construct a Uranium Solidification Facility at the Savannah River Site to process liquid uranyl nitrate into powder. Since the need for weapons materials has been reduced, an audit was conducted to assess the need for this facility. The audit disclosed that DOE continued to construct the facility, because DOE's procedures did not ensure that projects of this type were periodically reassessed when significant program changes occurred. The audit identified more economical alternatives for processing existing quantities of liquid uranyl nitrate at the Savannah River Site

  9. Fast critical assembly safeguards: NDA methods for highly enriched uranium. Summary report, October 1978-September 1979

    International Nuclear Information System (INIS)

    Bellinger, F.O.; Winslow, G.H.

    1980-12-01

    Nondestructive assay (NDA) methods, principally passive gamma measurements and active neutron interrogation, have been studied for their safeguards effectiveness and programmatic impact as tools for making inventories of highly enriched uranium fast critical assembly fuel plates. It was concluded that no NDA method is the sole answer to the safeguards problem, that each of those emphasized here has its place in an integrated safeguards system, and that each has minimum facility impact. It was found that the 185-keV area, as determined with a NaI detector, was independent of highly-enriched uranium (HEU) plate irradiation history, though the random neutron driver methods used here did not permit accurate assay of irradiated plates. Containment procedures most effective for accurate assaying were considered, and a particular geometry is recommended for active interrogation by a random driver. A model, pertinent to that geometry, which relates the effects of multiplication and self-absorption, is described. Probabilities of failing to detect that plates are missing are examined

  10. Uranium-thorium fuel cycle in a very high temperature hybrid system

    International Nuclear Information System (INIS)

    Hernandez, C.R.G.; Oliva, A.M.; Fajardo, L.G.; Garcia, J.A.R.; Curbelo, J.P.; Abadanes, A.

    2011-01-01

    Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. Therefore, Thorium fuels can complement Uranium fuels and ensure long term sustainability of nuclear power. The main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a Uranium-Thorium (U + Th) fuel cycle are shown in this paper. Once-through and two step U + Th fuel cycle was evaluated. With this goal, a preliminary conceptual design of a hybrid system formed by a Graphite Moderated Gas-Cooled Very High Temperature Reactor and two ADSs is proposed. The main parameters related to the neutronic behavior of the system in a deep burn scheme are optimized. The parameters that describe the nuclear fuel breeding and Minor Actinide stockpile are compared with those of a simple Uranium fuel cycle. (author)

  11. About the elaboration of pure uranium dicarbide

    International Nuclear Information System (INIS)

    Besson, J.; Blum, P.; Guinet, Ph.; Spitz, J.

    1963-01-01

    In order to develop methods for the elaboration of as pure as possible uranium dicarbide, the authors report the study of different elaboration processes based on the reaction between uranium and carbon, or between uranium and hydrocarbon, or between uranium oxide and carbon. They finally choose a method which comprises an arc-induced fusion of a mixture of uranium dioxide and carbon. The fusion process is described. The influence of thermal treatments is discussed as well as the graphite electrode carburization

  12. Quality assurance in the enriched uranium operations NDA facility

    Energy Technology Data Exchange (ETDEWEB)

    May, P.K.; Ceo, R.N. [Oak Ridge Y-12 Plant, TN (United States)

    1997-11-01

    The Nondestructive Analysis (NDA) Facility at the Oak Ridge Y-12 Plant has characterized process wastes for Enriched Uranium Operations since 1978. Since that time, over 50,000 items have been analyzed. Analysis results are used to determine whether or not recovery of uranium from process wastes is economically feasible. Our instrument complement includes one large segmented gamma scanner (SGS), two smaller SGS, two solution assay systems (SAS), and Active Well Coincidence Counter (AWCC). The large SGS is used for analyzing High Efficiency Particulate Air (HEPA) filters ant 208-L drums filled with combustible contaminated waste. The smaller SGS are used to analyze 4-L containers of ash and leached residues. The SAS are used to analyze 125 ml bottles of aqueous or organic waste solutions that may contain uranium. The gamma-based NDA techniques are used to identify which process wastes can be discarded, and which must be recycled. The AWCC is used to analyze high-density materials which are not amenable to gamma-ray analysis. 1 ref., 4 figs.

  13. Criticality safety and facility design considerations

    International Nuclear Information System (INIS)

    Waltz, W.R.

    1991-06-01

    Operations with fissile material introduce the risk of a criticality accident that may be lethal to nearby personnel. In addition, concerns over criticality safety can result in substantial delays and shutdown of facility operations. For these reasons, it is clear that the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The emphasis of this report will be placed on engineering design considerations in the prevention of criticality. The discussion will not include other important aspects, such as the physics of calculating limits nor criticality alarm systems

  14. Development of ISA procedure for uranium fuel fabrication and enrichment facilities: overview of ISA procedure and its application

    International Nuclear Information System (INIS)

    Yamate, Kazuki; Yamada, Takashi; Takanashi, Mitsuhiro; Sasaki, Noriaki

    2013-01-01

    Integrated Safety Analysis (ISA) procedure for uranium fuel fabrication and enrichment facilities has been developed for aiming at applying risk-informed regulation to these uranium facilities. The development has carried out referring to the ISA (NUREG-1520) by the Nuclear Regulatory Commission (NRC). The paper presents purpose, principles and activities for the development of the ISA procedure, including Risk Level (RL) matrix and grading evaluation method of IROFS (Items Relied on for Safety), as well as general description and features of the procedure. Also described in the paper is current status in application of risk information from the ISA. Japanese four licensees of the uranium facilities have been conducting ISA for their representative processes using the developed procedure as their voluntary safety activities. They have been accumulating experiences and knowledge on the ISA procedure and risk information through the field activities. NISA (Nuclear and Industrial Safety Agency) and JNES (Japan Nuclear Energy Safety Organization) are studying how to use such risk information for the safety regulation of the uranium facilities, taking into account the licensees' experiences and knowledge. (authors)

  15. Uranium ore mining in Spain with a focus on the closure and remediation measures in former production facilities

    International Nuclear Information System (INIS)

    Koch, H.; Blunck, S.; Lopez Romero, A.R.

    2004-01-01

    In early 2000, the uranium ore mining activities in Spain ceased. Since the middle of the last century, Spain had pushed ahead its own production of uranium concentrate with the formation of several companies (ENUSA, J.E.N.). In that period, Spain produced around 6000 t of uranium. With the completion of the operations at Andujar, La Haba and Elephante as well as Quercus at Saelices el Chico, the corporate tasks have shifted from building-up of a strategic uranium reserve to remediation and subsequent use of the locations. The operations have reached different remediation phases. While at Saelices el Chico remediation is still proceeding, the Andujar and La Haba locations are undergoing a monitoring phase as agreed for all former operating facilities. The estimated closure and remediation costs for the three operating facilities described amount to approx. 85 mio. Euro. In all three cases dealt with, however, these limited financial resources have been sufficient to successfully implement a closure and remediation concept that minimizes the risks from the facilities of uranium ore mining and processing with regard to the environment. (orig.)

  16. Uranium conversion

    International Nuclear Information System (INIS)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina

    2006-03-01

    FOI, has performed a study on uranium conversion processes that are of importance in the production of different uranium compounds in the nuclear industry. The same conversion processes are of interest both when production of nuclear fuel and production of fissile material for nuclear weapons are considered. Countries that have nuclear weapons ambitions, with the intention to produce highly enriched uranium for weapons purposes, need some degree of uranium conversion capability depending on the uranium feed material available. This report describes the processes that are needed from uranium mining and milling to the different conversion processes for converting uranium ore concentrate to uranium hexafluoride. Uranium hexafluoride is the uranium compound used in most enrichment facilities. The processes needed to produce uranium dioxide for use in nuclear fuel and the processes needed to convert different uranium compounds to uranium metal - the form of uranium that is used in a nuclear weapon - are also presented. The production of uranium ore concentrate from uranium ore is included since uranium ore concentrate is the feed material required for a uranium conversion facility. Both the chemistry and principles or the different uranium conversion processes and the equipment needed in the processes are described. Since most of the equipment that is used in a uranium conversion facility is similar to that used in conventional chemical industry, it is difficult to determine if certain equipment is considered for uranium conversion or not. However, the chemical conversion processes where UF 6 and UF 4 are present require equipment that is made of corrosion resistant material

  17. Uranium management activities

    International Nuclear Information System (INIS)

    Jackson, D.; Marshall, E.; Sideris, T.; Vasa-Sideris, S.

    2001-01-01

    One of the missions of the Department of Energy's (DOE) Oak Ridge Office (ORO) has been the management of the Department's uranium materials. This mission has been accomplished through successful integration of ORO's uranium activities with the rest of the DOE complex. Beginning in the 1980's, several of the facilities in that complex have been shut down and are in the decommissioning process. With the end of the Cold War, the shutdown of many other facilities is planned. As a result, inventories of uranium need to be removed from the Department facilities. These inventories include highly enriched uranium (HEU), low enriched uranium (LEU), normal uranium (NU), and depleted uranium (DU). The uranium materials exist in different chemical forms, including metals, oxides, solutions, and gases. Much of the uranium in these inventories is not needed to support national priorities and programs. (author)

  18. Uranium absorption study pile

    International Nuclear Information System (INIS)

    Raievski, V.; Sautiez, B.

    1959-01-01

    The report describes a pile designed to measure the absorption of fuel slugs. The pile is of graphite and comprises a central section composed of uranium rods in a regular lattice. RaBe sources and BF 3 counters are situated on either side of the center. A given uranium charge is compared with a specimen charge of about 560 kg, and the difference in absorption between the two noted. The sensitivity of the equipment will detect absorption variations of about a few ppm boron (10 -6 boron per gr. of uranium) or better. (author) [fr

  19. Critical analysis of world uranium resources

    Science.gov (United States)

    Hall, Susan; Coleman, Margaret

    2013-01-01

    The U.S. Department of Energy, Energy Information Administration (EIA) joined with the U.S. Department of the Interior, U.S. Geological Survey (USGS) to analyze the world uranium supply and demand balance. To evaluate short-term primary supply (0–15 years), the analysis focused on Reasonably Assured Resources (RAR), which are resources projected with a high degree of geologic assurance and considered to be economically feasible to mine. Such resources include uranium resources from mines currently in production as well as resources that are in the stages of feasibility or of being permitted. Sources of secondary supply for uranium, such as stockpiles and reprocessed fuel, were also examined. To evaluate long-term primary supply, estimates of uranium from unconventional and from undiscovered resources were analyzed. At 2010 rates of consumption, uranium resources identified in operating or developing mines would fuel the world nuclear fleet for about 30 years. However, projections currently predict an increase in uranium requirements tied to expansion of nuclear energy worldwide. Under a low-demand scenario, requirements through the period ending in 2035 are about 2.1 million tU. In the low demand case, uranium identified in existing and developing mines is adequate to supply requirements. However, whether or not these identified resources will be developed rapidly enough to provide an uninterrupted fuel supply to expanded nuclear facilities could not be determined. On the basis of a scenario of high demand through 2035, 2.6 million tU is required and identified resources in operating or developing mines is inadequate. Beyond 2035, when requirements could exceed resources in these developing properties, other sources will need to be developed from less well-assured resources, deposits not yet at the prefeasibility stage, resources that are currently subeconomic, secondary sources, undiscovered conventional resources, and unconventional uranium supplies. This

  20. The report of the criticality accident in a uranium conversion test plant in Tokai-mura

    International Nuclear Information System (INIS)

    Murata, Hajime; Akashi, Makoto

    2002-03-01

    The criticality accident in the title occurred at around 10:35, on Sep. 30, 1999, cost the lives of two workers and caused many residents concern on their health. Moreover, rumors had both social and economic consequences. This report is a detailed account of the roles that many individuals and groups in the National Institute of Radiological Sciences (NIRS) performed in a range of the areas, and is published to discharge NIRS responsibilities in regards to the accident. The report involves chapters of detailed outline of the accident; acceptance of the victims and communications until the identification of the ''criticality'' accident; initial treatment; dose estimation (medical, hematological, physical and biological ones and that by dental metals activated by the neutron); decision making for therapeutic strategies; cooperation with the Network Council for Radiation Emergency Medicine and other medical facilities; emergency importation of medical supplies; treatment and progress (nursing system and radiation injuries); protection from radiation in medical facilities; response to nearby residents of the Plant; international response; press release; Uranium Processing Plant Criticality Accident Investigation Committee and the Health Management Committee organized by the Nuclear Safety Commission; handling of information; and radiation emergency medical preparedness at the NIRS (future issues and prospect). The report is hoped to be useful in preventing the occurrence of future accidents. (K.H.)

  1. NIRS report of the criticality accident in a uranium conversion test plant in Tokai-mura

    International Nuclear Information System (INIS)

    2001-01-01

    This report is a detailed account of the roles that National Institute of Radiological Sciences (NIRS) played at the criticality accident in the title, which occurred at around 10:35, on Sep. 30, 1999 and resulted in death of two workers after all, and is published to discharge NIRS responsibilities in regards to the accident. The accident caused many residents concern on their health and rumors had both social and economic consequences. The report involves chapters of detailed outline of the accident; demand for acceptance of the victims and communications until the identification of the criticality'' accident; the acceptance and initial treatment; the exposure dose estimation (based on acute symptoms, on physics, on chromosomal analyses and on neutron-activated dental metals, and detailed analyses for dose distribution); decision made for therapeutic strategies; cooperation with the Network Council for Radiation Emergency and with other medical facilities; the urgent import of medicine; treatment and processes (patients, nursing system and radiation injuries); radiation protection in medical facilities; response to nearby residents of the Plant; international response; press release; Uranium Processing Plant Criticality Accident Investigation Committee and the Health Management Committee organized by the Nuclear Safety Commission; handling of information; and radiation emergency medical preparedness at the NIRS (future issues and prospect). The report is hopefully useful in preventing the occurrence of future accidents. (N.I.)

  2. On the applicability of the critical safety function concept to a uranium hexafluoride conversion unit

    International Nuclear Information System (INIS)

    Santos, F.C.; Goncalves, J.S.; Melo, P.F. Frutuoso e; Medeiros, J.A.C.C.

    2013-01-01

    This paper presents a discussion on the applicability on the critical safety function (CSF) concept as a design criterion for the new UF 6 conversion plant of Industrias Nucleares do Brazil (INB). This discussion is in the context of accident management, under the safety function oriented management. Safety functions may be understood as those whose loss may lead to releases of radioactive material or highly toxic chemicals, having possible radiological and/or occupational consequences for workers, the public or the environment. They should be designed to prevent criticality and to ensure adequate process confinement, thus preventing radioactive material releases that might lead to internal or external exposure and highly toxic chemical releases and exposure. The main hazards is the potential release of chemicals, especially HF and UF 6 . A criticality hazard exists only if the conversion facility processes uranium with a 235 U concentration greater than 1% Industrial activities for UF 6 production include handling and processing explosive, toxic and lethal chemicals, such as HF, H 2 and elemental F 2 , besides intermediate compounds containing uranium. State trees and definition of logical arrangements to construct an annunciation system are the next development stages, resulting form the establishment of applicable CSFs as representative of the next development stages, resulting from the establishment of applicable CSFs as representative of the various systems that make up the conversion plant. Discussed also in the biggest challenge of the development of this innovation, that is, the uncertainties related to the impact of human factors (not subject to monitoring by sensors or process conventional instrumentation). (author)

  3. Facile synthesis and enhanced visible-light photocatalysis of graphitic carbon nitride composite semiconductors.

    Science.gov (United States)

    Li, Huiquan; Liu, Yuxing; Gao, Xing; Fu, Cong; Wang, Xinchen

    2015-04-13

    The semiconductor heterojunction has been an effective architecture to enhance photocatalytic activity by promoting photogenerated charge separation. Here, graphitic carbon nitride (CN) and B-modified graphitic carbon nitride (CNB) composite semiconductors were fabricated by a facile calcination process using cheap, sustainable, and easily available sodium tetraphenylboron and urea as precursors. The synthetic CN-CNB-25 semiconductor with a suitable CNB content showed the highest visible-light activity. Its degradation ratio for methyl orange and phenol was more than twice that of CN and CNB and its H2 evolution rate was ∼3.4 and ∼1.8 times higher than that of CN and CNB, respectively. It also displayed excellent stability and reusability. The enhanced activity of CN-CNB-25 was attributed predominantly to the efficient separation of photoinduced electrons and holes. This paper describes a visible-light-responsive CN composite semiconductor with great potential in environmental and energy applications. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  4. Heat Transfer During Evaporation of Cesium From Graphite Surface in an Argon Environment

    Directory of Open Access Journals (Sweden)

    Bespala Evgeny

    2016-01-01

    Full Text Available The article focuses on discussion of problem of graphite radioactive waste formation and accumulation. It is shown that irradiated nuclear graphite being inalienable part of uranium-graphite reactor may contain fission and activation products. Much attention is given to the process of formation of radioactive cesium on the graphite element surface. It is described a process of plasma decontamination of irradiated graphite in inert argon atmosphere. Quasi-one mathematical model is offered, it describes heat transfer process in graphite-cesium-argon system. Article shows results of calculation of temperature field inside the unit cell. Authors determined the factors which influence on temperature change.

  5. Domestic round robin exercise on analysis of uranium for nuclear material handling facilities in Japan

    International Nuclear Information System (INIS)

    Kato, Yoshiyasu; Nagai, Kohta; Handa, Takamitsu; Inoue, Shin-ichi; Sato, Yoshihiro

    2016-01-01

    Interlaboratory comparison programme as well as internal quality control system is an effective tool for an analytical laboratory responsible to nuclear material accountancy of a nuclear facility to maintain and enhance its capability for analysis. However, it is a burden on nuclear material handling facilities in Japan to attend interlaboratory comparison programme run by overseas institutions because of high costs and complicated procedure for importing nuclear materials, and therefore facilities which can participate in such international programme would be limited. Nuclear Material Control Center has hence started and organised an annual domestic round robin exercise on analysis of uranium standard materials, funded by the Japan Safeguards Office of the Nuclear Regulation Authority, since 2008 to enhance analytical capability of Japanese Facilities. The outline of the round robin exercise will be given and the results of uranium isotopic and concentration analysis reported by participant facilities from 2008 to 2015 will be summarised in the presentation. (author)

  6. Electrochemically induced chemical sensor properties in graphite screen-printed electrodes: The case of a chemical sensor for uranium

    International Nuclear Information System (INIS)

    Kostaki, Vasiliki T.; Florou, Ageliki B.; Prodromidis, Mamas I.

    2011-01-01

    Highlights: → Electrochemical treatment endows analytical characteristics to SPEs. → A sensitive chemical sensor for uranium is described. → Performance is due to a synergy between electrochemical treatment and ink's solvents. → The amount of the solvent controls the achievable sensitivity. - Abstract: We report for the first time on the possibility to develop chemical sensors based on electrochemically treated, non-modified, graphite screen-printed electrodes (SPEs). The applied galvanostatic treatment (5 μA for 6 min in 0.1 M H 2 SO 4 ) is demonstrated to be effective for the development of chemical sensors for the determination of uranium in aqueous solutions. A detailed study of the effect of various parameters related to the fabrication of SPEs on the performance of the resulting sensors along with some diagnostic experiments on conventional graphite electrodes showed that the inducible analytical characteristics are due to a synergy between electrochemical treatment and ink's solvents. Indeed, the amount of the latter onto the printed working layer controls the achievable sensitivity. The preconcentration of the analyte was performed in an electroless mode in an aqueous solutions of U(VI), pH 4.6, and then, the accumulated species was reduced by means of a differential pulse voltammetry scan in 0.1 M H 3 BO 3 , pH 3. Under selected experimental conditions, a linear calibration curve over the range 5 x 10 -9 to 10 -7 M U(VI) was constructed. The 3σ limit of detection at a preconcentration time of 30 min, and the relative standard deviation of the method were 4.5 x 10 -9 M U(VI) and >12% (n = 5, 5 x 10 -8 M U(VI)), respectively. The effect of potential interferences was also examined.

  7. Criticality safety strategy for the Fuel Cycle Facility electrorefiner at Argonne National Laboratory, West

    International Nuclear Information System (INIS)

    Mariani, R.D.; Benedict, R.W.; Lell, R.M.; Turski, R.B.; Fujita, E.K.

    1993-01-01

    The Integral Fast Reactor being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal-cooled reactors and a closed fuel cycle. Presently, the Fuel Cycle Facility (FCF) at ANL-West in Idaho Falls, Idaho is being modified to recycle spent metallic fuel from Experimental Breeder Reactor II as part of a demonstration project sponsored by the Department of Energy. A key component of the FCF is the electrorefiner (ER) in which the actinides are separated from the fission products. In the electrorefining process, the metal fuel is anodically dissolved into a high-temperature molten salt and refined uranium or uranium/plutonium products are deposited at cathodes. In this report, the criticality safety strategy for the FCF ER is summarized. FCF ER operations and processes formed the basis for evaluating criticality safety and control during actinide metal fuel refining. In order to show criticality safety for the FCF ER, the reference operating conditions for the ER had to be defined. Normal operating envelopes (NOES) were then defined to bracket the important operating conditions. To keep the operating conditions within their NOES, process controls were identified that can be used to regulate the actinide forms and content within the ER. A series of operational checks were developed for each operation that wig verify the extent or success of an operation. The criticality analysis considered the ER operating conditions at their NOE values as the point of departure for credible and incredible failure modes. As a result of the analysis, FCF ER operations were found to be safe with respect to criticality

  8. Los Alamos Critical Experiments Facility

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1991-01-01

    The Critical Experiments Facility of the Los Alamos National Laboratory has been in existence for 45 years. In that period of time, thousands of measurements have been made on assemblies containing every fissionable material in various configurations that included bare metal and compounds of the nitrate, sulfate, fluoride, carbide, and oxide. Techniques developed or applied include Rossi-α, source-jerk, rod oscillator, and replacement measurements. Many of the original measurements of delay neutrons were performed at the site, and a replica of the Hiroshima weapon was operated at steady state to assist in evaluating the relative biological effectiveness (RBE) of neutrons. Solid, liquid, and gas fissioning systems were run at critical. Operation of this original critical facility has demonstrated the margin of safety that can be obtained through remote operation. Eight accidental excursions have occurred on the site, ranging from 1.5 x 10 16 to 1.2 x 10 17 fissions, with no significant exposure to personnel or damage to the facility beyond the machines themselves -- and in only one case was the machine damaged beyond further use. The present status of the facility, operating procedures, and complement of machines will be described in the context of programmatic activity. New programs will focus on training, validation of criticality alarm systems, experimental safety assessment of process applications, and dosimetry. Special emphasis will be placed on the incorporation of experience from 45 years of operation into present procedures and programs. 3 refs

  9. Conversion and Blending Facility highly enriched uranium to low enriched uranium as oxide. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-05

    This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials into pure HEU oxide and (2) blend the pure HEU oxide with depleted and natural uranium oxide to produce an LWR grade LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU will be produced as a waste suitable for storage or disposal.

  10. Uranium doping and neutron irradiation of Bi-2223 superconduction tapes for improved critical current density

    International Nuclear Information System (INIS)

    Moss, S.D.; Wang, W.G.; Dou, S.X.; Weinstein, R.

    1998-01-01

    It is demonstrated that a combination of neutron irradiation with uranium doping introduce fission tracks through a Bi-2223 tape which act as effective pinning centres, leading to a substantial increase in critical current. Preliminary data suggests that the combination of uranium doping and neutron irradiation produces improved flux pinning in Bi-2223 tapes over neutron irradiation alone. Before irradiation, SEM, DTA and XRD analyses were performed on the tapes. Both before and after irradiation the trapped maximum magnetic flux was measured at 77K. Before neutron irradiation, uranium doping has no effect on critical current. Preliminary SEM data suggested that the uranium is homogeneously distributed throughout the oxide core of the tape. The presence of 2212 and other secondary phases in the doped tapes suggest further refinement of the sintering procedure is necessary. (authors)

  11. Experimental critical parameters of enriched uranium solution in annular tank geometries

    International Nuclear Information System (INIS)

    Rothe, R.E.

    1996-04-01

    A total of 61 critical configurations are reported for experiments involving various combinations of annular tanks into which enriched uranium solution was pumped. These experiments were performed at two widely separated times in the 1980s under two programs at the Rocky Flats Plant's Critical Mass Laboratory. The uranyl nitrate solution contained about 370 g of uranium per liter, but this concentration varied a little over the duration of the studies. The uranium was enriched to about 93% [sup 235]U. All tanks were typical of sizes commonly found in nuclear production plants. They were about 2 m tall and ranged in diameter from 0.6 m to 1.5 m. Annular thicknesses and conditions of neutron reflection, moderation, and absorption were such that criticality would be achieved with these dimensions. Only 13 of the entire set of 74 experiments proved to be subcritical when tanks were completely filled with solution. Single tanks of several radial thicknesses were studied as well as small line arrays (1 x 2 and 1 x 3) of annular tanks. Many systems were reflected on four sides and the bottom by concrete, but none were reflected from above. Many experiments also contained materials within and outside the annular regions that contained strong neutron absorbers. One program had such a thick external moderator/absorber combination that no reflector was used at all

  12. Experimental critical parameters of enriched uranium solution in annular tank geometries

    Energy Technology Data Exchange (ETDEWEB)

    Rothe, R.E.

    1996-04-01

    A total of 61 critical configurations are reported for experiments involving various combinations of annular tanks into which enriched uranium solution was pumped. These experiments were performed at two widely separated times in the 1980s under two programs at the Rocky Flats Plant`s Critical Mass Laboratory. The uranyl nitrate solution contained about 370 g of uranium per liter, but this concentration varied a little over the duration of the studies. The uranium was enriched to about 93% [sup 235]U. All tanks were typical of sizes commonly found in nuclear production plants. They were about 2 m tall and ranged in diameter from 0.6 m to 1.5 m. Annular thicknesses and conditions of neutron reflection, moderation, and absorption were such that criticality would be achieved with these dimensions. Only 13 of the entire set of 74 experiments proved to be subcritical when tanks were completely filled with solution. Single tanks of several radial thicknesses were studied as well as small line arrays (1 x 2 and 1 x 3) of annular tanks. Many systems were reflected on four sides and the bottom by concrete, but none were reflected from above. Many experiments also contained materials within and outside the annular regions that contained strong neutron absorbers. One program had such a thick external moderator/absorber combination that no reflector was used at all.

  13. Fire hazards analysis for the uranium oxide (UO3) facility

    International Nuclear Information System (INIS)

    Wyatt, D.M.

    1994-01-01

    The Fire Hazards Analysis (FHA) documents the deactivation end-point status of the UO 3 complex fire hazards, fire protection and life safety systems. This FHA has been prepared for the Uranium Oxide Facility by Westinghouse Hanford Company in accordance with the criteria established in DOE 5480.7A, Fire Protection and RLID 5480.7, Fire Protection. The purpose of the Fire Hazards Analysis is to comprehensively and quantitatively assess the risk from a fire within individual fire areas in a Department of Energy facility so as to ascertain whether the objectives stated in DOE Order 5480.7, paragraph 4 are met. Particular attention has been paid to RLID 5480.7, Section 8.3, which specifies the criteria for deactivating fire protection in decommission and demolition facilities

  14. Numerical simulation of radon migration from a uranium ore storage facility

    International Nuclear Information System (INIS)

    Vasil'ev, I.A.; Politov, V.Yu.; Chernov, V.V.; Shestakov, A.A.

    2007-01-01

    Data on geologic structure and radiation environment in the vicinity of the tailings storage facility (TSF) of Kara-Balta uranium hydrometallurgical factory in Kyrgyzstan were used to design a mathematical model of radon migration from the surface of TSF. Numerical calculations have been performed to describe prevalence of radon contamination [ru

  15. Hypertension and hematologic parameters in a community near a uranium processing facility

    International Nuclear Information System (INIS)

    Wagner, Sara E.; Burch, James B.; Bottai, Matteo; Pinney, Susan M.; Puett, Robin; Porter, Dwayne; Vena, John E.; Hebert, James R.

    2010-01-01

    Background: Environmental uranium exposure originating as a byproduct of uranium processing can impact human health. The Fernald Feed Materials Production Center functioned as a uranium processing facility from 1951 to 1989, and potential health effects among residents living near this plant were investigated via the Fernald Medical Monitoring Program (FMMP). Methods: Data from 8216 adult FMMP participants were used to test the hypothesis that elevated uranium exposure was associated with indicators of hypertension or changes in hematologic parameters at entry into the program. A cumulative uranium exposure estimate, developed by FMMP investigators, was used to classify exposure. Systolic and diastolic blood pressure and physician diagnoses were used to assess hypertension; and red blood cells, platelets, and white blood cell differential counts were used to characterize hematology. The relationship between uranium exposure and hypertension or hematologic parameters was evaluated using generalized linear models and quantile regression for continuous outcomes, and logistic regression or ordinal logistic regression for categorical outcomes, after adjustment for potential confounding factors. Results: Of 8216 adult FMMP participants 4187 (51%) had low cumulative uranium exposure, 1273 (15%) had moderate exposure, and 2756 (34%) were in the high (>0.50 Sievert) cumulative lifetime uranium exposure category. Participants with elevated uranium exposure had decreased white blood cell and lymphocyte counts and increased eosinophil counts. Female participants with higher uranium exposures had elevated systolic blood pressure compared to women with lower exposures. However, no exposure-related changes were observed in diastolic blood pressure or hypertension diagnoses among female or male participants. Conclusions: Results from this investigation suggest that residents in the vicinity of the Fernald plant with elevated exposure to uranium primarily via inhalation exhibited

  16. Hypertension and hematologic parameters in a community near a uranium processing facility

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, Sara E., E-mail: swagner@uga.edu [College of Public Health, Department of Epidemiology and Biostatistics, Paul D. Coverdell Center for Biomedical and Health Sciences, University of Georgia, 500 D.W. Brooks Drive, Athens, GA 30602-7396 (United States); Burch, James B. [Arnold School of Public Health, Department of Epidemiology and Biostatistics, University of South Carolina, Columbia, SC (United States); South Carolina Statewide Cancer Prevention and Control Program, Columbia, SC (United States); WJB Dorn Veteran' s Affairs Medical Center, Columbia, SC (United States); Bottai, Matteo [Arnold School of Public Health, Department of Epidemiology and Biostatistics, University of South Carolina, Columbia, SC (United States); Pinney, Susan M. [College of Medicine, Department of Environmental Health, University of Cincinnati, Cincinnati, OH (United States); Puett, Robin [Arnold School of Public Health, Department of Epidemiology and Biostatistics, University of South Carolina, Columbia, SC (United States); South Carolina Statewide Cancer Prevention and Control Program, Columbia, SC (United States); Arnold School of Public Health, Department of Environmental Health Sciences, University of South Carolina, Columbia, SC (United States); Porter, Dwayne [Arnold School of Public Health, Department of Environmental Health Sciences, University of South Carolina, Columbia, SC (United States); Vena, John E. [College of Public Health, Department of Epidemiology and Biostatistics, Paul D. Coverdell Center for Biomedical and Health Sciences, University of Georgia, 500 D.W. Brooks Drive, Athens, GA 30602-7396 (United States); Hebert, James R. [Arnold School of Public Health, Department of Epidemiology and Biostatistics, University of South Carolina, Columbia, SC (United States); South Carolina Statewide Cancer Prevention and Control Program, Columbia, SC (United States)

    2010-11-15

    Background: Environmental uranium exposure originating as a byproduct of uranium processing can impact human health. The Fernald Feed Materials Production Center functioned as a uranium processing facility from 1951 to 1989, and potential health effects among residents living near this plant were investigated via the Fernald Medical Monitoring Program (FMMP). Methods: Data from 8216 adult FMMP participants were used to test the hypothesis that elevated uranium exposure was associated with indicators of hypertension or changes in hematologic parameters at entry into the program. A cumulative uranium exposure estimate, developed by FMMP investigators, was used to classify exposure. Systolic and diastolic blood pressure and physician diagnoses were used to assess hypertension; and red blood cells, platelets, and white blood cell differential counts were used to characterize hematology. The relationship between uranium exposure and hypertension or hematologic parameters was evaluated using generalized linear models and quantile regression for continuous outcomes, and logistic regression or ordinal logistic regression for categorical outcomes, after adjustment for potential confounding factors. Results: Of 8216 adult FMMP participants 4187 (51%) had low cumulative uranium exposure, 1273 (15%) had moderate exposure, and 2756 (34%) were in the high (>0.50 Sievert) cumulative lifetime uranium exposure category. Participants with elevated uranium exposure had decreased white blood cell and lymphocyte counts and increased eosinophil counts. Female participants with higher uranium exposures had elevated systolic blood pressure compared to women with lower exposures. However, no exposure-related changes were observed in diastolic blood pressure or hypertension diagnoses among female or male participants. Conclusions: Results from this investigation suggest that residents in the vicinity of the Fernald plant with elevated exposure to uranium primarily via inhalation exhibited

  17. DRY TRANSFER FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Dry Transfer Facility Description Document'' (BSC 2005 [DIRS 173737], p. 3-8). A description of the changes is as follows: (1) Update the supporting calculations for the various Category 1 and 2 event sequences as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2005 [DIRS 171429], Section 7). (2) Update the criticality safety calculations for the DTF staging racks and the remediation pool to reflect the current design. This design calculation focuses on commercial spent nuclear fuel (SNF) assemblies, i.e., pressurized water reactor (PWR) and boiling water reactor (BWR) SNF. U.S. Department of Energy (DOE) Environmental Management (EM) owned SNF is evaluated in depth in the ''Canister Handling Facility Criticality Safety Calculations'' (BSC 2005 [DIRS 173284]) and is also applicable to DTF operations. Further, the design and safety analyses of the naval SNF canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. Also, note that the results for the Monitored Geologic Repository (MGR) Site specific Cask (MSC) calculations are limited to the

  18. The Fracture Toughness of Nuclear Graphites Grades

    Energy Technology Data Exchange (ETDEWEB)

    Burchell, Timothy D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Erdman, III, Donald L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lowden, Rick R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunter, James A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hannel, Cara C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-04-01

    New measurements of graphite mode I critical stress intensity factor, KIc (commonly referred to as the fracture toughness) and the mode II critical shear stress intensity, KIIc, are reported and compared with prior data for KIc and KIIc. The new data are for graphite grades PCEA, IG-110 and 2114. Variations of KIc and acoustic emission (AE) data with graphite texture are reported and discussed. The Codes and Standards applications of fracture toughness, KIc, data are also discussed. A specified minimum value for nuclear graphite KIc is recommended.

  19. Criticality Safety Evaluation of Hanford Tank Farms Facility

    Energy Technology Data Exchange (ETDEWEB)

    WEISS, E.V.

    2000-12-15

    Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste.

  20. Criticality Safety Evaluation of Hanford Tank Farms Facility

    International Nuclear Information System (INIS)

    WEISS, E.V.

    2000-01-01

    Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste

  1. Uranium conversion; Urankonvertering

    Energy Technology Data Exchange (ETDEWEB)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina [Swedish Defence Research Agency (FOI), Stockholm (Sweden)

    2006-03-15

    FOI, has performed a study on uranium conversion processes that are of importance in the production of different uranium compounds in the nuclear industry. The same conversion processes are of interest both when production of nuclear fuel and production of fissile material for nuclear weapons are considered. Countries that have nuclear weapons ambitions, with the intention to produce highly enriched uranium for weapons purposes, need some degree of uranium conversion capability depending on the uranium feed material available. This report describes the processes that are needed from uranium mining and milling to the different conversion processes for converting uranium ore concentrate to uranium hexafluoride. Uranium hexafluoride is the uranium compound used in most enrichment facilities. The processes needed to produce uranium dioxide for use in nuclear fuel and the processes needed to convert different uranium compounds to uranium metal - the form of uranium that is used in a nuclear weapon - are also presented. The production of uranium ore concentrate from uranium ore is included since uranium ore concentrate is the feed material required for a uranium conversion facility. Both the chemistry and principles or the different uranium conversion processes and the equipment needed in the processes are described. Since most of the equipment that is used in a uranium conversion facility is similar to that used in conventional chemical industry, it is difficult to determine if certain equipment is considered for uranium conversion or not. However, the chemical conversion processes where UF{sub 6} and UF{sub 4} are present require equipment that is made of corrosion resistant material.

  2. Preconcentration of trace elements from high-purity thorium and uranium on Chelex-100 and determination by graphite furnace atomic absorption spectrometry with Zeeman-effect background correction

    International Nuclear Information System (INIS)

    Raje, Naina; Kayasth, Satish; Asari, T.P.S.; Gangadharan, S.

    1994-01-01

    Preconcentration of trace impurities from large-sized samples of uranium metal and thorium oxide using a small column of Chelex-100 followed by their determination using graphite furnace atomic absorption spectrometry (GFAAS) is reported. A 0.5-10-g amount of the sample (uranium metal or thorium oxide) was dissolved, complexed with ammonium carbonate and subjected to the ion-exchange procedure. The retained analytes were eluted with 2-4 M nitric acid and brought to a small volume for a final dilution to 10-25 ml for their determination using GFAAS. The validity of the separation procedure and recoveries at μg kg -1 levels was checked by standard addition; the recoveries were >95%

  3. On the applicability of the critical safety function concept to a uranium hexafluoride conversion unit

    Energy Technology Data Exchange (ETDEWEB)

    Santos, F.C.; Goncalves, J.S.; Melo, P.F. Frutuoso e; Medeiros, J.A.C.C., E-mail: fcruz@nuclear.ufrj.br, E-mail: jsgoncalves@inb.gov.br, E-mail: frutuoso@nuclear.ufrj.br, E-mail: canedo@imp.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This paper presents a discussion on the applicability on the critical safety function (CSF) concept as a design criterion for the new UF{sub 6} conversion plant of Industrias Nucleares do Brazil (INB). This discussion is in the context of accident management, under the safety function oriented management. Safety functions may be understood as those whose loss may lead to releases of radioactive material or highly toxic chemicals, having possible radiological and/or occupational consequences for workers, the public or the environment. They should be designed to prevent criticality and to ensure adequate process confinement, thus preventing radioactive material releases that might lead to internal or external exposure and highly toxic chemical releases and exposure. The main hazards is the potential release of chemicals, especially HF and UF{sub 6}. A criticality hazard exists only if the conversion facility processes uranium with a {sup 235}U concentration greater than 1% Industrial activities for UF{sub 6} production include handling and processing explosive, toxic and lethal chemicals, such as HF, H{sub 2} and elemental F{sub 2}, besides intermediate compounds containing uranium. State trees and definition of logical arrangements to construct an annunciation system are the next development stages, resulting form the establishment of applicable CSFs as representative of the next development stages, resulting from the establishment of applicable CSFs as representative of the various systems that make up the conversion plant. Discussed also in the biggest challenge of the development of this innovation, that is, the uncertainties related to the impact of human factors (not subject to monitoring by sensors or process conventional instrumentation). (author)

  4. Alecto - results obtained with homogeneous critical experiments on plutonium 239, uranium 235 and uranium 233; Alecto - resultats des experiences critiques homogenes realisees sur le plutonium 239, l'uranium 235 et l'uranium 233

    Energy Technology Data Exchange (ETDEWEB)

    Bruna, J G; Brunet, J P; Caizegues, R; Clouet d' Orval, Ch; Kremser, J; Tellier, H; Verriere, Ph [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    In this report are given the results of the homogeneous critical experiments ALECTO, made on plutonium 239, uranium 235 and uranium 233. After a brief description of the equipment, the critical masses for cylinders of diameters varying from 25 to 42 cm, are given and compared with other values (foreign results, criticality guide). With respect to the specific conditions of neutron reflection in the ALECTO experiments the minimal values of critical masses are: Pu239 M{sub c} = 910 {+-} 10 g, U235 M{sub c} = 1180 {+-} 12 g and U233 M{sub c} = 960 {+-} 10 g. Experiments relating to cross sections and constants to be used on these materials are presented. Lastly, kinetic experiments allow to compare pulsed neutron methods to fluctuation methods. [French] On presente dans ce rapport les resultats des experiences critiques homogenes ALECTO, effectuees sur le plutonium 239, l'uranium 235 et l'uranium 233. Apres avoir rappele la description des installations, on donne les masses critiques pour des cylindres de diametres variant entre 25 et 42 cm, qui sont comparees avec d'autres chiffres (resultats etrangers, guide de criticite). Dans les gammes des diametres etudies pour des cuves a fond plat reflechies lateralement, la valeur minimale des masses critiques est la suivante: Pu239 M{sub c} = 910 {+-} 10 g, U235 M{sub c} = 1180 {+-} 12 g et U233 M{sub c} 960 {+-} 10 g. Des experiences portant sur les sections efficaces et les constantes a utiliser sur ces milieux sont ensuite presentees. Enfin des experiences de cinetique permettent une comparaison entre la methode des neutrons pulses et la methode des fluctuations. (auteur)

  5. Improved mechanical and functional properties of elastomer/graphite nanocomposites prepared by latex compounding

    Energy Technology Data Exchange (ETDEWEB)

    Yang Jian [Key Laboratory for Nano-materials, Beijing University of Chemical Technology, Ministry of Education of China, Beijing 100029 (China); Key Laboratory on Preparation and Processing of Novel Polymer Materials, Beijing University of Chemical Technology, Beijing 100029 (China); Tian Ming [Key Laboratory for Nano-materials, Beijing University of Chemical Technology, Ministry of Education of China, Beijing 100029 (China); Jia Qingxiu [Key Laboratory on Preparation and Processing of Novel Polymer Materials, Beijing University of Chemical Technology, Beijing 100029 (China); Shi Junhong [Key Laboratory for Nano-materials, Beijing University of Chemical Technology, Ministry of Education of China, Beijing 100029 (China); Zhang Liqun [Key Laboratory for Nano-materials, Beijing University of Chemical Technology, Ministry of Education of China, Beijing 100029 (China); Key Laboratory on Preparation and Processing of Novel Polymer Materials, Beijing University of Chemical Technology, Beijing 100029 (China)], E-mail: zhanglq@mail.buct.edu.cn; Lim Szuhui; Yu Zhongzhen [Centre for Advanced Materials Technology (CAMT), School of Aerospace, Mechanical and Mechatronic Engineering (J07), University of Sydney, Sydney, NSW 2006 (Australia); Mai Yiuwing [Centre for Advanced Materials Technology (CAMT), School of Aerospace, Mechanical and Mechatronic Engineering (J07), University of Sydney, Sydney, NSW 2006 (Australia)], E-mail: y.mai@usyd.edu.au

    2007-10-15

    The facile latex approach has been adopted to finely incorporate graphite nanosheets into elastomeric polymer matrix to obtain high-performance elastomeric nanocomposites with improved mechanical properties and functional properties. Scanning electron microscopy, transmission electron microscopy and X-ray diffraction experiments show that the nanostructures of the final nanocomposites exhibit a high degree of exfoliation and intercalation of graphite in the nitrile-butadiene rubber (NBR) matrix. Mechanical and dynamic-mechanical tests demonstrate that the NBR/graphite nanocomposites possess greatly increased elastic modulus and tensile strength, and desirably strong interfaces. The unexpected self-crosslinking of elastomer/graphite nanocomposites was discovered and then verified by oscillating disc rheometry and equilibrium swelling experiments. After critically examining various polymer types by X-ray photoelectron spectroscopy, electron spin resonance and Fourier transform infrared spectroscopy, a radical initiation mechanism was proposed to explain the self-crosslinking reaction. These NBR/graphite nanocomposites possess significantly improved wear resistance and gas barrier properties, and superior electrical/thermal conductivity. Such versatile functional properties make NBR nanocomposites a promising new class of advanced materials.

  6. Improved mechanical and functional properties of elastomer/graphite nanocomposites prepared by latex compounding

    International Nuclear Information System (INIS)

    Yang Jian; Tian Ming; Jia Qingxiu; Shi Junhong; Zhang Liqun; Lim Szuhui; Yu Zhongzhen; Mai Yiuwing

    2007-01-01

    The facile latex approach has been adopted to finely incorporate graphite nanosheets into elastomeric polymer matrix to obtain high-performance elastomeric nanocomposites with improved mechanical properties and functional properties. Scanning electron microscopy, transmission electron microscopy and X-ray diffraction experiments show that the nanostructures of the final nanocomposites exhibit a high degree of exfoliation and intercalation of graphite in the nitrile-butadiene rubber (NBR) matrix. Mechanical and dynamic-mechanical tests demonstrate that the NBR/graphite nanocomposites possess greatly increased elastic modulus and tensile strength, and desirably strong interfaces. The unexpected self-crosslinking of elastomer/graphite nanocomposites was discovered and then verified by oscillating disc rheometry and equilibrium swelling experiments. After critically examining various polymer types by X-ray photoelectron spectroscopy, electron spin resonance and Fourier transform infrared spectroscopy, a radical initiation mechanism was proposed to explain the self-crosslinking reaction. These NBR/graphite nanocomposites possess significantly improved wear resistance and gas barrier properties, and superior electrical/thermal conductivity. Such versatile functional properties make NBR nanocomposites a promising new class of advanced materials

  7. Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico

    International Nuclear Information System (INIS)

    Salazar, M.; Elder, J.

    1992-08-01

    The Ultra-High Temperature Reactor Experiment (UHTREX) facility was constructed in the late 1960s to advance high-temperature and gas-cooled reactor technology. The 3-MW reactor was graphite moderated and helium cooled and used 93% enriched uranium as its fuel. The reactor was run for approximately one year and was shut down in February 1970. The decommissioning of the facility involved removing the reactor and its associated components. This document details planning for the decommissioning operations which included characterizing the facility, estimating the costs of decommissioning, preparing environmental documentation, establishing a system to track costs and work progress, and preplanning to correct health and safety concerns in the facility. Work to decommission the facility began in 1988 and was completed in September 1990 at a cost of $2.9 million. The facility was released to Department of Energy for other uses in its Los Alamos program

  8. Ion irradiation to simulate neutron irradiation in model graphites: Consequences for nuclear graphite

    Science.gov (United States)

    Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

    2017-10-01

    Due to its excellent moderator and reflector qualities, graphite was used in CO2-cooled nuclear reactors such as UNGG (Uranium Naturel-Graphite-Gaz). Neutron irradiation of graphite resulted in the production of 14C which is a key issue radionuclide for the management of the irradiated graphite waste. In order to elucidate the impact of neutron irradiation on 14C behavior, we carried out a systematic investigation of irradiation and its synergistic effects with temperature in Highly Oriented Pyrolitic Graphite (HOPG) model graphite used to simulate the coke grains of nuclear graphite. We used 13C implantation in order to simulate 14C displaced from its original structural site through recoil. The collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic damage. However, a part of the recoil carbon atom energy is also transferred to the graphite lattice through electronic excitation. The effects of the different irradiation regimes in synergy with temperature were simulated using ion irradiation by varying Sn(nuclear)/Se(electronic) stopping power. Thus, the samples were irradiated with different ions of different energies. The structure modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry. The results show that temperature generally counteracts the disordering effects of irradiation but the achieved reordering level strongly depends on the initial structural state of the graphite matrix. Thus, extrapolating to reactor conditions, for an initially highly disordered structure, irradiation at reactor temperatures (200 - 500 °C) should induce almost no change of the initial structure. On the contrary, when the structure is initially less disordered, there should be a "zoning" of the reordering: In "cold" high flux irradiated zones where the ballistic damage is important, the structure should be poorly reordered; In "hot" low flux irradiated zones where the ballistic

  9. Isotopic dilution requirements for 233U criticality safety in processing and disposal facilities

    International Nuclear Information System (INIS)

    Elam, K.R.; Forsberg, C.W.; Hopper, C.M.; Wright, R.Q.

    1997-11-01

    The disposal of excess 233 U as waste is being considered. Because 233 U is a fissile material, one of the key requirements for processing 233 U to a final waste form and disposing of it is to avoid nuclear criticality. For many processing and disposal options, isotopic dilution is the most feasible and preferred option to avoid nuclear criticality. Isotopic dilution is dilution of fissile 233 U with nonfissile 238 U. The use of isotopic dilution removes any need to control nuclear criticality in process or disposal facilities through geometry or chemical composition. Isotopic dilution allows the use of existing waste management facilities, that are not designed for significant quantities of fissile materials, to be used for processing and disposing of 233 U. The amount of isotopic dilution required to reduce criticality concerns to reasonable levels was determined in this study to be ∼ 0.66 wt% 233 U. The numerical calculations used to define this limit consisted of a homogeneous system of silicon dioxide (SiO 2 ), water (H 2 O), 233 U, and depleted uranium (DU) in which the ratio of each component was varied to determine the conditions of maximum nuclear reactivity. About 188 parts of DU (0.2 wt% 235 U) are required to dilute 1 part of 233 U to this limit in a water-moderated system with no SiO 2 present. Thus, for the US inventory of 233 U, several hundred metric tons of DU would be required for isotopic dilution

  10. Validation of gamma-ray detection techniques for safeguards monitoring at natural uranium conversion facilities

    Energy Technology Data Exchange (ETDEWEB)

    Dewji, S.A., E-mail: dewjisa@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Road, MS-6335, Oak Ridge, TN 37831-6335 (United States); Lee, D.L.; Croft, S. [Oak Ridge National Laboratory, 1 Bethel Valley Road, MS-6335, Oak Ridge, TN 37831-6335 (United States); Hertel, N.E. [Oak Ridge National Laboratory, 1 Bethel Valley Road, MS-6335, Oak Ridge, TN 37831-6335 (United States); Nuclear and Radiological Engineering Program, Georgia Institute of Technology, 770 State Street, Atlanta, GA 30332-0745 (United States); Chapman, J.A.; McElroy, R.D.; Cleveland, S. [Oak Ridge National Laboratory, 1 Bethel Valley Road, MS-6335, Oak Ridge, TN 37831-6335 (United States)

    2016-07-01

    Recent IAEA circulars and policy papers have sought to implement safeguards when any purified aqueous uranium solution or uranium oxides suitable for isotopic enrichment or fuel fabrication exists. Under the revised policy, IAEA Policy Paper 18, the starting point for nuclear material under safeguards was reinterpreted, suggesting that purified uranium compounds should be subject to safeguards procedures no later than the first point in the conversion process. In response to this technical need, a combination of simulation models and experimental measurements were employed to develop and validate concepts of nondestructive assay monitoring systems in a natural uranium conversion plant (NUCP). In particular, uranyl nitrate (UO{sub 2}(NO{sub 3}){sub 2}) solution exiting solvent extraction was identified as a key measurement point (KMP), where gamma-ray spectroscopy was selected as the process monitoring tool. The Uranyl Nitrate Calibration Loop Equipment (UNCLE) facility at Oak Ridge National Laboratory was employed to simulate the full-scale operating conditions of a purified uranium-bearing aqueous stream exiting the solvent extraction process in an NUCP. Nondestructive assay techniques using gamma-ray spectroscopy were evaluated to determine their viability as a technical means for drawing safeguards conclusions at NUCPs, and if the IAEA detection requirements of 1 significant quantity (SQ) can be met in a timely way. This work investigated gamma-ray signatures of uranyl nitrate circulating in the UNCLE facility and evaluated various gamma-ray detector sensitivities to uranyl nitrate. These detector validation activities include assessing detector responses to the uranyl nitrate gamma-ray signatures for spectrometers based on sodium iodide, lanthanum bromide, and high-purity germanium detectors. The results of measurements under static and dynamic operating conditions at concentrations ranging from 10–90 g U/L of natural uranyl nitrate are presented. A range of

  11. Management of UKAEA graphite liabilities

    International Nuclear Information System (INIS)

    Wise, M.

    2001-01-01

    The UK Atomic Energy Authority (UKAEA) is responsible for managing its liabilities for redundant research reactors and other active facilities concerned with the development of the UK nuclear technology programme since 1947. These liabilities include irradiated graphite from a variety of different sources including low irradiation temperature reactor graphite (the Windscale Piles 1 and 2, British Energy Pile O and Graphite Low Energy Experimental Pile at Harwell and the Material Testing Reactors at Harwell and Dounreay), advanced gas-cooled reactor graphite (from the Windscale Advanced Gas-cooled Reactor) and graphite from fast reactor systems (neutron shield graphite from the Dounreay Prototype Fast Reactor and Dounreay Fast Reactor). The decommissioning and dismantling of these facilities will give rise to over 6,000 tonnes of graphite requiring disposal. The first graphite will be retrieved from the dismantling of Windscale Pile 1 and the Windscale Advanced Gas-cooled Reactor during the next five years. UKAEA has undertaken extensive studies to consider the best practicable options for disposing of these graphite liabilities in a manner that is safe whilst minimising the associated costs and technical risks. These options include (but are not limited to), disposal as Low Level Waste, incineration, or encapsulation and disposal as Intermediate Level Waste. There are a number of technical issues associated with each of these proposed disposal options; these include Wigner energy, radionuclide inventory determination, encapsulation of graphite dust, galvanic coupling interactions enhancing the corrosion of mild steel and public acceptability. UKAEA is currently developing packaging concepts and designing packaging plants for processing these graphite wastes in consultation with other holders of graphite wastes throughout Europe. 'Letters of Comfort' have been sought from both the Low Level Waste and the Intermediate Level Waste disposal organisations to support the

  12. Radiological considerations in the design of Reprocessing Uranium Plant (RUP) of Fast Reactor Fuel Cycle Facility (FRFCF), Kalpakkam

    International Nuclear Information System (INIS)

    Chandrasekaran, S.; Rajagopal, V.; Jose, M.T.; Venkatraman, B.

    2012-01-01

    A Fast Reactor Fuel Cycle Facility (FRFCF) being planned at Indira Gandhi Centre for Atomic Research, Kalpakkam is an integrated facility with head end and back end of fuel cycle plants co-located in a single place, to meet the refuelling needs of the prototype fast breeder reactor (PFBR). Reprocessed uranium oxide plant (RUP) is one such plant in FRFCF to built to meet annual requirements of UO 2 for fabrication of fuel sub-assemblies (FSAs) and radial blanket sub-assemblies (RSAs) for PFBR. RUP receives reprocessed uranium oxide powder (U 3 O 8 ) from fast reactor fuel reprocessing plant (FRP) of FRFCF. Unlike natural uranium oxide plant, RUP has to handle reprocessed uranium oxide which is likely to have residual fission products activity in addition to traces of plutonium. As the fuel used for PFBR is recycled within these plants, formation of higher actinides in the case of plutonium and formation of higher levels of 232 U in the uranium product would be a radiological problem to be reckoned with. The paper discussed the impact of handling of multi-recycled reprocessed uranium in RUP and the radiological considerations

  13. Surveillance and Maintenance Plan for the Uranium Trioxide(UO3) Facility

    International Nuclear Information System (INIS)

    McGuire, J.J.

    1999-01-01

    This document provides a plan for implementing surveillance and maintenance (S and M) activities to ensure the Uranium Oxide(UO3) Facility is maintained in a safe, environmentally secure, and cost effective manner until subsequent closure during the final disposition phase of decommissioning. This plan has been prepared in accordance with the guidelines provided in the U.S. Department of Energy (DOE) Office of Environmental Management (EM) Decommissioning Resource Manual (DOE 1995) and Section 8.6 of TPA change form P-08-97-01 to the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) (Ecology, et al. 1996)

  14. Studies on yttrium oxide coatings for corrosion protection against molten uranium

    International Nuclear Information System (INIS)

    Chakravarthy, Y.; Bhandari, Subhankar; Pragatheeswaran; Thiyagarajan, T.K.; Ananthapadmanabhan, P.V.; Das, A.K.; Kumar, Jay; Kutty, T.R.G.

    2012-01-01

    Yttrium oxide is resistant to corrosion by molten uranium and its alloys. Yttrium oxide is recommended as a protective oxide layer on graphite and metal components used for melting and processing uranium and its alloys. This paper presents studies on the efficacy of plasma sprayed yttrium oxide coatings for barrier applications against molten uranium

  15. An assessment of the effectiveness of personal visual observation for a uranium enrichment facility (2)

    International Nuclear Information System (INIS)

    Bando, Masatsugu; Okamoto, Tsuyoshi

    2003-01-01

    In a centrifuge uranium enrichment facility, a large number of unit cascades are operated to produce low enriched uranium for nuclear power reactors. Some thousands of UF 6 gas centrifuges are installed in unit cascade. If a new type of advanced centrifuge is developed in the near future, the number of stages and UF 6 gas centrifuges in the unit cascade would decrease dramatically. Furthermore, an integrated type of centrifuge, which is composed of a few tens of centrifuges, is adopted from the point of economic view, the piping arrangement among UF 6 gas centrifuges can be more simplified. It can be said that the simpler the piping arrangement, the less the operation time we are required to make any diverted cascade with the help of re-arrangement of the unit cascade piping. When two type of centrifuge, conventional and advanced centrifuge are used in the uranium enrichment facility, we predicted an inspection effort of personal visual observation for inspector by Game Theory. In our mathematical model, an activity of inspection in a cascade area is simplified into two-person non-cooperative game between inspector and facility operator. As a result of our calculation, it became clear that total inspection effort is likely to increase unless the integrated type of centrifuge is installed. (author)

  16. Measurement and instrumentation techniques for monitoring plutonium and uranium particulates released from nuclear facilities

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1976-08-01

    The purpose of this work has been an analysis and evaluation of the state-of-the-art of measurement and instrumentation techniques for monitoring plutonium and uranium particulates released from nuclear facilities. The occurrence of plutonium and uranium in the nuclear fuel cycle, the corresponding potential for releases, associated radiological protection standards and monitoring objectives are discussed. Techniques for monitoring via decay radiation from plutonium and uranium isotopes are presented in detail, emphasizing air monitoring, but also including soil sampling and survey methods. Additionally, activation and mass measurement techniques are discussed. The availability and prevalence of these various techniques are summarized. Finally, possible improvements in monitoring capabilities due to alterations in instrumentation, data analysis, or programs are presented

  17. Airborne uranium, its concentration and toxicity in uranium enrichment facilities

    International Nuclear Information System (INIS)

    Thomas, J.; Mauro, J.; Ryniker, J.; Fellman, R.

    1979-02-01

    The release of uranium hexafluoride and its hydrolysis products into the work environment of a plant for enriching uranium by means of gas centrifuges is discussed. The maximum permissible mass and curie concentration of airborne uranium (U) is identified as a function of the enrichment level (i.e., U-235/total U), and chemical and physical form. A discussion of the chemical and radiological toxicity of uranium as a function of enrichment and chemical form is included. The toxicity of products of UF 6 hydrolysis in the atmosphere, namely, UO 2 F 2 and HF, the particle size of toxic particulate material produced from this hydrolysis, and the toxic effects of HF and other potential fluoride compounds are also discussed. Results of an investigation of known effects of humidity and temperature on particle size of UO 2 F 2 produced by the reaction of UF 6 with water vapor in the air are reported. The relationship of the solubility of uranium compounds to their toxic effects was studied. Identification and discussion of the standards potentially applicable to airborne uranium compounds in the working environment are presented. The effectiveness of High Efficiency Particulate (HEPA) filters subjected to the corrosive environment imposed by the presence of hydrogen fluoride is discussed

  18. Remediation of uranium mill tailings wastes in Australia: a critical review

    International Nuclear Information System (INIS)

    Mudd, G.M.

    2000-01-01

    Australia has been an active participant in the global uranium mining industry since its inception in the 1940s. By the late 1950s five major mining and milling projects were operating, several small mines supplied custom ores. All of these projects were closed by the early 1960s, except for Rum Jungle which continued under government subsidy. Most sites have had lasting Environmental impacts. The advances in nuclear power in the 1960s saw increasing demand for uranium and Australia again explored with remarkable success in the Northern Territory, South Australia and Western Australia. After several government inquiries in the 1970s, Ranger, Nabarlek and Olympic Dam were operating by the mid 1980s. The principal risks from uranium mill tailings wastes arise from their radioactive nature and often their chemical toxicities. A critical review of the rehabilitation of abandoned uranium mines and mill tailings as a comparison for current projects is presented. It is concluded that the management of uranium mill tailings wastes is a complex task, requiring a sound multi-disciplinary approach. The problems include groundwater contamination, erosion, radon emanation and gamma radiation. evidence to data from the remediation of old and modern sites does not demonstrate effective long-term closure and safety

  19. Process for decontamination of surfaces in an facility of natural uranium hexafluoride production (UF6)

    International Nuclear Information System (INIS)

    Almeida, Claudio C. de; Silva, Teresinha M.; Rodrigues, Demerval L.; Carneiro, Janete C.G.G.

    2017-01-01

    The experience acquired in the actions taken during the decontamination process of an IPEN-CNEN / SP Nuclear and Energy Research Institute facility, for the purpose of making the site unrestricted, is reported. The steps of this operation involved: planning, training of facility operators, workplace analysis and radiometric measurements. The facility had several types of equipment from the natural uranium hexafluoride (UF 6 ) production tower and other facility materials. Rules for the transportation of radioactive materials were established, both inside and outside the facility and release of materials and installation

  20. Facile Synthesis of Nitrogen Doped Graphene Oxide from Graphite Flakes and Powders: A Comparison of Their Surface Chemistry.

    Science.gov (United States)

    Yokwana, Kholiswa; Ray, Sekhar C; Khenfouch, Mohammad; Kuvarega, Alex T; Mamba, Bhekie B; Mhlanga, Sabelo D; Nxumalo, Edward N

    2018-08-01

    Nitrogen-doped graphene oxide (NGO) nanosheets were prepared via a facile one-pot modified Hummer's approach at low temperatures using graphite powder and flakes as starting materials in the presence of a nitrogen precursor. It was found that the morphology, structure, composition and surface chemistry of the NGO nanosheets depended on the nature of the graphite precursor used. GO nanosheets doped with nitrogen atoms exhibited a unique structure with few thin layers and wrinkled sheets, high porosity and structural defects. NGO sheets made from graphite powder (NGOp) exhibited excellent thermal stability and remarkably high surface area (up to 240.53 m2 ·g-1) compared to NGO sheets made from graphite flakes (NGOf) which degraded at low temperatures and had an average surface area of 24.70 m2 ·g-1. NGOf sheets had a size range of 850 to 2200 nm while NGOp sheets demonstrated obviously small sizes (460-1600 nm) even when exposed to different pH conditions. The NGO nanosheets exhibited negatively charged surfaces in a wide pH range (1 to 12) and were found to be stable above pH 6. In addition, graphite flakes were found to be more suitable for the production of NGO as they produced high N-doping levels (0.65 to 1.29 at.%) compared to graphite powders (0.30 to 0.35 at.%). This study further demonstrates that by adjusting the amount of N source in the host GO, one can tailor its thermal stability, surface morphology, surface chemistry and surface area.

  1. Criticality safety considerations. Integral Monitored Retrievable Storage (MRS) Facility

    International Nuclear Information System (INIS)

    1986-09-01

    This report summarizes the criticality analysis performed to address criticality safety concerns and to support facility design during the conceptual design phase of the Monitored Retrievable Storage (MRS) Facility. The report addresses the criticality safety concerns, the design features of the facility relative to criticality, and the results of the analysis of both normal operating and hypothetical off-normal conditions. Key references are provided (Appendix C) if additional information is desired by the reader. The MRS Facility design was developed and the related analysis was performed in accordance with the MRS Facility Functional Design Criteria and the Basis for Design. The detailed description and calculations are documented in the Integral MRS Facility Conceptual Design Report. In addition to the summary portion of this report, explanatary notes for various terms, calculation methodology, and design parameters are presented in Appendix A. Appendix B provides a brief glossary of technical terms

  2. Irradiation-induced amorphization process in graphite

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Hiroaki [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

    1996-04-01

    Effects of the element process of irradiation damage on irradiation-induced amorphization processes of graphite was studied. High orientation thermal decomposed graphite was cut about 100 nm width and used as samples. The irradiation experiments are carried out under the conditions of electronic energy of 100-400 KeV, ion energy of 200-600 KeV, ionic species Xe, Ar, Ne, C and He and the irradiation temperature at from room temperature to 900 K. The critical dose ({phi}a) increases exponentially with increasing irradiation temperature. The displacement threshold energy of graphite on c-axis direction was 27 eV and {phi}a{sup e} = 0.5 dpa. dpa is the average number of displacement to atom. The critical dose of ion irradiation ({phi}a{sup i}) was 0.2 dpa at room temperature, and amorphous graphite was produced by less than half of dose of electronic irradiation. Amorphization of graphite depending upon temperature is discussed. (S.Y.)

  3. Compliance determination procedures for environmental radiation protection standards for uranium recovery facilities 40 CFR part 190

    International Nuclear Information System (INIS)

    1982-03-01

    Uranium Milling operations are licensed by the Nuclear Regulatory Commission and by some States in agreement with the Commission. The radiation dose to any individual from the operation of facilities within the uranium fuel cycle is limited to levels set by the Environmental Protection Agency. These levels are contained in the EPA Environmental Radiation Protection Standards for Nuclear Power Operations, in Part 190 of Title 40 of the Code of Federal Regulations (40 CFR Part 190). This report describes the procedures used within NRC's Uranium Recovery Licensing Branch for evaluating compliance with these regulations for uranium milling operations. The report contains descriptions of these procedures, dose factors for evaluating environmental measurement data, and guidance to the NRC staff reviewer

  4. Four critical facilities: their capabilities and programs

    International Nuclear Information System (INIS)

    Whitesides, G.E.

    1980-01-01

    Information is presented on the critical experiments facilities at Babcock and Wilcox, Lynchburg, Virginia; at Battelle Pacific Northwest Laboratory in Hanford, Washington; at Rockwell-International in Rocky Flats, Colorado; and at Los Alamos Scientific Laboratory in New Mexico. It is noted that the critical mass facilities which still exist in this country represent a bare minimum for maintaining a measurement program sufficient for meeting data requirements

  5. {sup 36}Cl and {sup 14}C behaviour in UNGG graphite during leaching experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pichon, C.; Guy, C.; Comte, J. [Commissariat a l' Energie Atomique - C.E.A., Laboratoire d' Analyses Radiochimiques et Chimiques (L.A.R.C.) 13108 Saint Paul lez Durance (France)

    2008-07-01

    Graphite has been used as a moderator in Natural Uranium Graphite Gas reactors. Among the radionuclides, the long-lived activation product {sup 36}Cl and {sup 14}C, which are abundant in graphite after irradiation can be the main contributors to the dose during disposal. This paper deals with the first results obtained on irradiated graphite from French G2 reactor. Both leaching and diffusion experiments have been performed in order to understand and quantify the radionuclides behaviour. Chlorine leaching seems to be controlled by diffusion transport through graphite matrix. On the contrary {sup 14}C leaching is very low, probably because after irradiation, the remaining {sup 14}C was produced from {sup 13}C activation in the crystalline structure of graphite. (authors)

  6. Rockwell International's Critical Mass Laboratory Program at the Rocky Flats Plant

    International Nuclear Information System (INIS)

    McCarthy, J.D.

    1984-01-01

    The primary mission of the laboratory is to provide data in support of plant operations. To fulfill this task, the facility has unique capabilities for perfoming general purpose critical mass experiment. The critical mass laboratory performed over 1000 critical measurements, primarily with plutonium metal and uranium metal, oxide and solution; it worked also on the NRC program (high-enriched uranium measurements). Presently the laboratory staff prepares for a series of critical measurements on a poisoned tube tank; the laboratory intends to continue to pursue basic plant support programs in the future

  7. Critical analysis of the management of waste system originated at the uranium mining and processing. A case study of the Concentrated Unit of Uranium - INB

    International Nuclear Information System (INIS)

    Araujo, Valeska Peres de

    2005-01-01

    The uranium world market faced a depreciation of this commodity during the last decades. Recently, decreases in the secondary supply (represented by highly enriched uranium - HEU - stocks detained by the former Soviet Union) turned out the market dependent upon primary supply again. In order to cope with this changing demands and market conditions, new uranium mining/milling projects must start operation, or at least, former uranium production plants must be improved. Environmental questions have been and certainly will continue to be a determinant factor concerning the operational feasibility of these facilities. Mining/milling activities have the potential to cause risks to the human health and to the environment. In case of uranium projects, radiological impacts shall also be taken into consideration. Amongst the most relevant environmental aspects associated with the operation of a uranium project, generated wastes are usually of major concern and deserve appropriate management strategies. As a result the objective of the present work was to examine the waste management system of the Brazilian uranium production unity located at the municipality of Caetite, northeast region of the country. An open pit mine and a milling facility compose this unit. The extraction method employed is acid heap leach (using H 2 SO 4 ). It could be assessed that the overall conceptual management strategy is in agreement with the practices adopted worldwide. Atmospheric impacts, caused by the emissions of radon and aerosols must be investigated in more details. Mathematical simulation revealed that no significant impact in groundwater is expected due to mobilization and transport of radionuclides from the milling wastes. However, the impacts of drainage water, accumulated in the open pit, into groundwater cannot be discarded yet. Screening techniques were applied to assess the potential contribution of the leached ore piles as a 226 Ra source of pollution. Our results did not allow

  8. Conversion and Blending Facility Highly enriched uranium to low enriched uranium as uranium hexafluoride. Revision 1

    International Nuclear Information System (INIS)

    1995-01-01

    This report describes the Conversion and Blending Facility (CBF) which will have two missions: (1) convert surplus HEU materials to pure HEU UF 6 and a (2) blend the pure HEU UF 6 with diluent UF 6 to produce LWR grade LEU-UF 6 . The primary emphasis of this blending be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The chemical and isotopic concentrations of the blended LEU product will be held within the specifications required for LWR fuel. The blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry

  9. Conversion and Blending Facility Highly enriched uranium to low enriched uranium as uranium hexafluoride. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-05

    This report describes the Conversion and Blending Facility (CBF) which will have two missions: (1) convert surplus HEU materials to pure HEU UF{sub 6} and a (2) blend the pure HEU UF{sub 6} with diluent UF{sub 6} to produce LWR grade LEU-UF{sub 6}. The primary emphasis of this blending be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The chemical and isotopic concentrations of the blended LEU product will be held within the specifications required for LWR fuel. The blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry.

  10. Earthquake research for the safer siting of critical facilities

    Energy Technology Data Exchange (ETDEWEB)

    Cluff, J.L. (ed.)

    1980-01-01

    The task of providing the necessities for living, such as adequate electrical power, water, and fuel, is becoming more complicated with time. Some of the facilities that provide these necessities would present potential hazards to the population if serious damage were to occur to them during earthquakes. Other facilities must remain operable immediately after an earthquake to provide life-support services to people who have been affected. The purpose of this report is to recommend research that will improve the information available to those who must decide where to site these critical facilities, and thereby mitigate the effects of the earthquake hazard. The term critical facility is used in this report to describe facilities that could seriously affect the public well-being through loss of life, large financial loss, or degradation of the environment if they were to fail. The term critical facility also is used to refer to facilities that, although they pose a limited hazard to the public, are considered critical because they must continue to function in the event of a disaster so that they can provide vital services.

  11. ZPR-3 Assembly 6F : A spherical assembly of highly enriched uranium, depleted uranium, aluminum and steel with an average {sup 235}U enrichment of 47 atom %.

    Energy Technology Data Exchange (ETDEWEB)

    Lell, R. M.; McKnight, R. D; Schaefer, R. W.; Nuclear Engineering Division

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average compositions of the core configurations were essentially identical in phases A - F. ZPR-3

  12. Carbon nanoencapsulation of uranium dicarbide

    International Nuclear Information System (INIS)

    Pasqualini, E.

    1996-01-01

    Nanoparticles of uranium dicarbide encapsulated in carbon smaller than 100 nm have been obtained by chemical reactions at high temperature. Two types of nanocapsules were identified and characterized. The majority of them had small diffuse kernel surfaces, with dimensions between 5 and 15 nm, surrounded by thick spherical carbon cover. Others, in minor quantity and ranging from 15 to 40 nm, were polyhedrical and surrounded with several perfect graphite layers oriented parallel to their external surface. The nanocapsules are as chemically inert as graphite. (orig.)

  13. Carbon nanoencapsulation of uranium dicarbide

    Energy Technology Data Exchange (ETDEWEB)

    Pasqualini, E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Dept. Combustibles Nucleares; Adelfang, P. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Dept. Combustibles Nucleares; Regueiro, M.N. [EPM-Matformag, CNRS, Grenoble (France)

    1996-07-01

    Nanoparticles of uranium dicarbide encapsulated in carbon smaller than 100 nm have been obtained by chemical reactions at high temperature. Two types of nanocapsules were identified and characterized. The majority of them had small diffuse kernel surfaces, with dimensions between 5 and 15 nm, surrounded by thick spherical carbon cover. Others, in minor quantity and ranging from 15 to 40 nm, were polyhedrical and surrounded with several perfect graphite layers oriented parallel to their external surface. The nanocapsules are as chemically inert as graphite. (orig.).

  14. Recent developments in graphite

    International Nuclear Information System (INIS)

    Cunningham, J.E.

    1983-01-01

    Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications

  15. Lead determination in uranium mineralization soils by atomic absorption spectrometry with graphite oven

    International Nuclear Information System (INIS)

    Teixeira, Gleber Tacio

    2001-01-01

    The contamination of soils by lead has a great environmental importance due to its toxicity to vegetables, animals and humans. In general, the mobility of the lead is small due to its low solubility and strong adsorption in the soil. However, its solubility can be altered by several conditions (pH, redox potential and ionic stronger). Consequently, lead can migrate through the soil and can contaminate superficial and underground waters. The objective of this work was to determine the concentration of total lead in soil samples with uranium mineralization, in an area at Ipora/GO, having been evaluated as economically insuitable the extraction of that mineral. The radiogenic lead appears as a product of natural radioactive elements decay. In the decay series of uranium-238 we found the isotope lead-214 (half-life of 26,8 min), lead-210 (half-life of 22,3 min), and lead-206 that is stable. The sampling was done in profiles around north, south, east and west directions, starting from a reference point (FT), chosen by presenting the largest radiation of that place (4800 cps). A mass of 1 Kg of superficial soil was collected to each 20 m, in each profile, until 150 m of FT. Approximately, 1 g of dry soil, fraction 2 mm, was digested with a mixture of acids HNO 3 /HClO 4 2:1 (v/v), and the resulting solution was analyzed by atomic absorption. An atomic absorption spectrometer was used with graphite furnace, with deuterium arc to background correction and pyrolytic coated tube. Phosphoric acid was used as chemical modifier. The obtained results, using the standard additions method, presented a decrease of the lead concentration, in all profiles, when the distance of FT was increased. It was also made a radiometric screening in each sampling point. The lead concentration variate from 115,1 μg.g -1 in FT, to less than 40 μg.g -1 at 150 m of distance of FT ( 3 ) 2 was used. The method was applied to a certified sample, showing a good agreement between certified and

  16. Criticality Analysis of the U-H2O Subcritical Assembly Modified for Rand D of the High Temperature Reactor

    International Nuclear Information System (INIS)

    Syarip; Tri-Wulan-Tjiptono; Tegas-Sutondo

    2000-01-01

    A criticality analysis of the natural uranium - light water sub-criticalassembly available at the P3TM-BATAN Yogyakarta, converted into a naturaluranium - graphite system has been performed. The purpose of this study is toprovide the research facility on the basic static and kinetics studies forthe high temperature reactor (HTR) in which the HTR fuel system is underdevelopment at the P3TM. For the purpose of this study, a neutroniccalculation was performed using WIMSD/4 code, to determine the neutronmultiplication factor for various fuel configurations of the sub-criticalassemblies. The results show that the effective neutron multiplication factor(k ef ) for U-Be-H 2 O and U-Be-He systems are 1.0474 and 1.4666 respectively,while for the graphite moderated systems with coolants of H 2 O or He(U-C-H 2 O and U-C-He) systems, the corresponding k ef are 0.787 and 0.4211respectively. The results conclude that the modification of U-H 2 O toU-C-H 2 O system, in accordance with neutronic is quite feasible, safe, cheapand practical, and in addition, the treatment of H 2 O is relatively easy.(author)

  17. Outline of criticality safety research project

    International Nuclear Information System (INIS)

    Kobayashi, Iwao; Tachimori, Shoichi; Suzaki, Takenori; Takeshita, Isao; Miyoshi, Yoshinori; Nakajima, Ken; Sakurai, Satoshi; Yanagisawa, Hiroshi

    1987-01-01

    As the power generation capacity of LWRs in Japan increased, the establishment and development of nuclear fuel cycle have become the important subject. Conforming to the safety research project of the nation, the Japan Atomic Energy Research Institute has advanced the project of constructing a new research facility, that is, Nuclear Fuel Cycle Engineering Research Facility (NUCEF). In this facility, it is planned to carry out the research on criticality safety, upgraded reprocessing techniques, and the treatment and disposal of transuranium element wastes. In this paper, the subjects of criticality safety research and the research carried out with a criticality safety experiment facility which is expected to be installed in the NUCEF are briefly reported. The experimental data obtained from the criticality safety handbooks and published literatures in foreign countries are short of the data on the mixture of low enriched uranium and plutonium which is treated in the reprocessing of spent fuel from LWRs. The acquisition of the criticality data for various forms of fuel, the elucidation of the scenario of criticality accidents, and the soundness of the confinement system for gaseous fission products and plutonium are the main subjects. The Static Criticality Safety Facility, Transient Criticality Safety Facility and pulse column system are the main facilities. (Kako, I.)

  18. Nuclear criticality safety program at the Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lell, R.M.; Fujita, E.K.; Tracy, D.B.; Klann, R.T.; Imel, G.R.; Benedict, R.W.; Rigg, R.H.

    1994-01-01

    The Fuel Cycle Facility (FCF) is designed to demonstrate the feasibility of a novel commercial-scale remote pyrometallurgical process for metallic fuels from liquid metal-cooled reactors and to show closure of the Integral Fast Reactor (IFR) fuel cycle. Requirements for nuclear criticality safety impose the most restrictive of the various constraints on the operation of FCF. The upper limits on batch sizes and other important process parameters are determined principally by criticality safety considerations. To maintain an efficient operation within appropriate safety limits, it is necessary to formulate a nuclear criticality safety program that integrates equipment design, process development, process modeling, conduct of operations, a measurement program, adequate material control procedures, and nuclear criticality analysis. The nuclear criticality safety program for FCF reflects this integration, ensuring that the facility can be operated efficiently without compromising safety. The experience gained from the conduct of this program in the Fuel cycle Facility will be used to design and safely operate IFR facilities on a commercial scale. The key features of the nuclear criticality safety program are described. The relationship of these features to normal facility operation is also described

  19. MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    2009-01-01

    The purpose of this analysis is to validate the Monte Carlo N-Particle 5 (MCNP5) code Version 1.40 (LA-UR-03-1987, 2005) and its cross-section database for k-code calculations of intermediate enriched uranium systems on INTEL(reg s ign) processor based PC's running any version of the WINDOWS operating system. Configurations with intermediate enriched uranium were modeled with the moderator range of 39 (le) H/Fissile (le) 1438. See Table 2-1 for brief descriptions of selected cases and Table 3-1 for the range of applicability for this validation. A total of 167 input cases were evaluated including bare and reflected systems in a single body or arrays. The 167 cases were taken directly from the previous (Version 4C [Lan 2005]) validation database. Section 2.0 list data used to calculate k-effective (k eff ) for the 167 experimental criticality benchmark cases using the MCNP5 code v1.40 and its cross section database. Appendix B lists the MCNP cross-section database entries validated for use in evaluating the intermediate enriched uranium systems for criticality safety. The dimensions and atom densities for the intermediate enriched uranium experiments were taken from NEA/NSC/DOC(95)03, September 2005, which will be referred to as the benchmark handbook throughout the report. For these input values, the experimental benchmark k eff is approximately 1.0. The MCNP validation computer runs ran to an accuracy of approximately ± 0.001. For the cases where the reported benchmark k eff was not equal to 1.0000 the MCNP calculational results were normalized. The difference between the MCNP validation computer runs and the experimentally measured k eff is the MCNP5 v1.40 bias. The USLSTATS code (ORNL 1998) was utilized to perform the statistical analysis and generate an acceptable maximum k eff limit for calculations of the intermediate enriched uranium type systems.

  20. Critical Facilities for Coastal Geographies

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The critical facilities data are derived from the USGS Structures Inventory Database (June, 2016). The structures in the derived dataset displays aggregated totals...

  1. Uranium oxide recovering method

    International Nuclear Information System (INIS)

    Ota, Kazuaki; Takazawa, Hiroshi; Teramae, Naoki; Onoue, Takeshi.

    1997-01-01

    Nitrates containing uranium nitrate are charged in a molten salt electrolytic vessel, and a heat treatment is applied to prepare molten salts. An anode and a cathode each made of a graphite rod are disposed in the molten salts. AC voltage is applied between the anode and the cathode to conduct electrolysis of the molten salts. Uranium oxides are deposited as a recovered product of uranium, on the surface of the anode. The nitrates containing uranium nitrate are preferably a mixture of one or more nitrates selected from sodium nitrate, potassium nitrate, calcium nitrate and magnesium nitrate with uranium nitrate. The nitrates may be liquid wastes of nitrates. The temperature for the electrolysis of the molten salts is preferably from 150 to 300degC. The voltage for the electrolysis of the molten salts is preferably an AC voltage of from 2 to 6V, more preferably from 4 to 6V. (I.N.)

  2. Highly Enriched Uranium Metal Cylinders Surrounded by Various Reflector Materials

    International Nuclear Information System (INIS)

    Bernard Jones; J. Blair Briggs; Leland Monteirth

    2007-01-01

    A series of experiments was performed at Los Alamos Scientific Laboratory in 1958 to determine critical masses of cylinders of Oralloy (Oy) reflected by a number of materials. The experiments were all performed on the Comet Universal Critical Assembly Machine, and consisted of discs of highly enriched uranium (93.3 wt.% 235U) reflected by half-inch and one-inch-thick cylindrical shells of various reflector materials. The experiments were performed by members of Group N-2, particularly K. W. Gallup, G. E. Hansen, H. C. Paxton, and R. H. White. This experiment was intended to ascertain critical masses for criticality safety purposes, as well as to compare neutron transport cross sections to those obtained from danger coefficient measurements with the Topsy Oralloy-Tuballoy reflected and Godiva unreflected critical assemblies. The reflector materials examined in this series of experiments are as follows: magnesium, titanium, aluminum, graphite, mild steel, nickel, copper, cobalt, molybdenum, natural uranium, tungsten, beryllium, aluminum oxide, molybdenum carbide, and polythene (polyethylene). Also included are two special configurations of composite beryllium and iron reflectors. Analyses were performed in which uncertainty associated with six different parameters was evaluated; namely, extrapolation to the uranium critical mass, uranium density, 235U enrichment, reflector density, reflector thickness, and reflector impurities. In addition to the idealizations made by the experimenters (removal of the platen and diaphragm), two simplifications were also made to the benchmark models that resulted in a small bias and additional uncertainty. First of all, since impurities in core and reflector materials are only estimated, they are not included in the benchmark models. Secondly, the room, support structure, and other possible surrounding equipment were not included in the model. Bias values that result from these two simplifications were determined and associated

  3. Uranium fixation by mineralization at the redox front

    International Nuclear Information System (INIS)

    Isobe, Hiroshi

    1998-01-01

    The behavior of actinide elements including uranium in geomedia is controlled by redox conditions. Under the oxidized conditions, uranium forms uranyl ion (UO 2 2+ ) and its complexes, and dissolves in ground water. Under the reduced conditions, U(IV) has much lower solubility than uranyl ion. In the Koongarra uranium deposit, Australia, lead-bearing uraninite, uranyl lead oxide and uranyl silicate minerals occur in the unweathered, primary ore zone, and uranyl phosphate minerals occur in the weathered, secondary ore zone. Between unweathered and weathered zones, the transition zone exists as a redox front. In the transition zone, graphite and sulfide minerals react as reducing agents for species dissolved in ground water. By SEM, spherical grains of uraninite were observed in veins with graphite. Pyrite had coffinite rim with crystals of uraninite. Calculation based on the ground water chemistry and hydrology at Koongarra shows that the uranium in the transition zone may be fixed from the ground water. In the Koongarra transition zone, recent mineralization of uranium by reduction takes place. Mineralization is much stronger fixation mechanism than adsorption on clay minerals. Pyrite in the buffer materials of possible radioactive waste repositories can fix radionuclides in oxidized ground water by mineralization with reducing reactions. (author)

  4. Characterization of aerosols in uranium handling facilities and its impact on the assessment of internal dose

    International Nuclear Information System (INIS)

    Roy, Ankush; Rao, D.D.; Sawant, Pramilla D.; Khan, Arshad; Srinivasan, P.; Chandrashekara, A.

    2016-01-01

    In nuclear facilities, compounds of uranium such as Magnesium DiUranate (MDU) U 3 O 8 , UO 2 etc. are handled in different stages of operation. There may be a possibility of intake of these compounds by radiation workers during the course of their work. The internal doses received by the workers depend not only on the quantity but also the physiochemical characteristics of the radioactive contaminant. The depositions in different regions of lung of these inhaled aerosols depend on their particle size; whereas the clearance is dependent upon the chemical nature. In this study, aerosol characterization is carried out in four different Uranium Handling Facilities (UF) for realistic assessment of internal dose to the radiation worker

  5. Seed irradiation facilities at TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Najzer, M.

    1972-01-01

    Fast neutrons and gamma-rays with their high and low LET respectively are excellent complementary tools for investigation of the effect of different types of mutations. TRIGA Irradiation Facility and Thermal Column Irradiation Facility were designed and installed for the first time in the TRIGA tank and thermal column respectively. The basic idea of design was the use of depleted uranium as gamma-ray and thermal neutron shield and simultaneously as thermal to fast neutron converter. Low LET radiation, due to direct and thermal neutron capture gamma-rays, is strongly attenuated while fast neutron flux is increased. GIF is made of a cadmium tube inserted in a graphite block. It is located in the central thermal column channel. The basic idea is to convert thermal neutrons to gamma-rays by capture in the cadmium

  6. Uranium Enrichment Determination of the InSTEC Sub Critical Ensemble Fuel by Gamma Spectrometry

    International Nuclear Information System (INIS)

    Borrell Munnoz, Jose L.; LopezPino, Neivy; Diaz Rizo, Oscar; D'Alessandro Rodriguez, Katia; Padilla Cabal, Fatima; Arbelo Penna, Yunieski; Garcia Rios, Aczel R.; Quintas Munn, Ernesto L.; Casanova Diaz, Amaya O.

    2009-01-01

    Low background gamma spectrometry was applied to analyze the uranium enrichment of the nuclear fuel used in the InSTEC Sub Critical ensemble. The enrichment was calculated by two variants: an absolute method using the Monte Carlo method to simulated detector volumetric efficiency, and an iterative procedure without using standard sources. The results confirm that the nuclear fuel of the ensemble is natural uranium without any additional degree of enrichment. (author)

  7. Environmental monitoring program design for uranium refining and conversion operations

    International Nuclear Information System (INIS)

    1984-08-01

    The objective of this study was to develop recommendations for the design of environmental monitoring programs at Canadian uranium refining and conversion operations. In order to develop monitoring priorities, chemical and radioactive releases to the air and water were developed for reference uranium refining and conversion facilities. The relative significance of the radioactive releases was evaluated through a pathways analysis which estimated dose to individual members of the critical receptor group. The effects of chemical releases to the environment were assessed by comparing predicted air and water contaminant levels to appropriate standards or guidelines. For the reference facilities studied, the analysis suggested that environmental effects are likely to be dominated by airborne release of both radioactive and nonradioactive contaminants. Uranium was found to be the most important radioactive species released to the air and can serve as an overall indicator of radiological impacts for any of the plants considered. The most important nonradioactive air emission was found to be fluoride (as hydrogen fluoride) from the uranium hexafluoride plant. For the uranium trioxide and uranium dioxide plants, air emissions of oxides of nitrogen were considered to be most important. The study recommendations for the design of an environmental monitoring program are based on consideration of those factors most likely to affect local air and water quality, and human radiation exposure. Site- and facility-specific factors will affect monitoring program design and the selection of components such as sampling media, locations and frequency, and analytical methods

  8. Special graphites; Graphites speciaux

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [French] Ameliorer les proprietes du graphite nucleaire pour empilements et ouvrir de nouveaux domaines d'application au graphite constituent une part importante de l'effort entrepris en commun par le Commissariat a l'Energie Atomique (CEA) et la compagnie PECHINEY. Des procedes nouveaux de fabrication de carbones et graphites speciaux ont ete mis au point: graphite forge, pyrocarbone, graphite de haute densite, agglomeration de poudres de graphite par craquage de gaz naturel, graphites impermeables. Les proprietes physiques de ces produits ainsi que leur reaction avec differents gaz oxydants sont decrites. Les premiers resultats d'irradiation sont aussi donnes. (auteurs)

  9. Operating procedures for the Pajarito Site Critical Assembly Facility

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1983-03-01

    Operating procedures consistent with DOE Order 5480.2, Chapter VI, and the American National Standard Safety Guide for the Performance of Critical Experiments are defined for the Pajarito Site Critical Assembly Facility of the Los Alamos National Laboratory. These operating procedures supersede and update those previously published in 1973 and apply to any criticality experiment performed at the facility

  10. Environmental radiological impact of some Portuguese uranium mines

    International Nuclear Information System (INIS)

    Galvao, J.P.; Bettencourt, A.O.; Teixeira, M.M.G.R.; Elias, M.D.T.

    1988-01-01

    An environmental radiological surveillance programme has been in progress around the most significant Portuguese uranium mines, from 1976 to 1983. A short description is given of the mines of Urgeirica (including uranium milling), Freixiosa, Pinhal de Souto and Bica. The results of the surveillance programme developed in the vicinity of these facilities are presented and the identified critical pathways are discussed. One of these pathways is the consumption of cabbage, which is an important component in the diet of the Portuguese population. The exposure of the critical groups, due to the intake of 226 Ra through the diet, calculated from the results of this monitoring programme, range from 0.02 to 1.5 mSv.y -1 for the different mines under study. (author)

  11. Basis for Interim Operation for Fuel Supply Shutdown Facility

    International Nuclear Information System (INIS)

    BENECKE, M.W.

    2003-01-01

    This document establishes the Basis for Interim Operation (BIO) for the Fuel Supply Shutdown Facility (FSS) as managed by the 300 Area Deactivation Project (300 ADP) organization in accordance with the requirements of the Project Hanford Management Contract procedure (PHMC) HNF-PRO-700, ''Safety Analysis and Technical Safety Requirements''. A hazard classification (Benecke 2003a) has been prepared for the facility in accordance with DOE-STD-1027-92 resulting in the assignment of Hazard Category 3 for FSS Facility buildings that store N Reactor fuel materials (303-B, 3712, and 3716). All others are designated Industrial buildings. It is concluded that the risks associated with the current and planned operational mode of the FSS Facility (uranium storage, uranium repackaging and shipment, cleanup, and transition activities, etc.) are acceptable. The potential radiological dose and toxicological consequences for a range of credible uranium storage building have been analyzed using Hanford accepted methods. Risk Class designations are summarized for representative events in Table 1.6-1. Mitigation was not considered for any event except the random fire event that exceeds predicted consequences based on existing source and combustible loading because of an inadvertent increase in combustible loading. For that event, a housekeeping program to manage transient combustibles is credited to reduce the probability. An additional administrative control is established to protect assumptions regarding source term by limiting inventories of fuel and combustible materials. Another is established to maintain the criticality safety program. Additional defense-in-depth controls are established to perform fire protection system testing, inspection, and maintenance to ensure predicted availability of those systems, and to maintain the radiological control program. It is also concluded that because an accidental nuclear criticality is not credible based on the low uranium enrichment

  12. Environmental monitoring program of the uranium enrichment facility Almirante Alvaro Alberto

    International Nuclear Information System (INIS)

    Hiromoto, G.; Jacomino, V.M.F.; Venturini, L.; Moreira, S.R.D.; Gordon, A.M.P.L.; Duarte, C.L.; Pocequilo, B.R.S.; Mazzilli, B.P.

    1988-11-01

    In this report, the Environmental Monitoring Program of the Uranium Enrichment Facility Almirante Alvaro Alberto is outlined and the results obtained during the preoperational period are presented. Information concerning the population distribution, the use of water and land, the local agricultural production and the local meteorology are also available. In order to evaluate the levels of the background radiation, sample of water, air and biological and terrestrial indicators were analysed. Measurements were performed of natural gamma emitters concentrations levels and of uranium in air, surface water, precipitation, groundwater, river sediment, soil, grass, vegetation and various foodstuffs. For direct measurement of background radiation levels a solid state dosimeter network was used. Results are also presented for the analysis of non radioactive pollutants in the water samples and for the particles and gaseous fluorides concentration in the atmosphere. (author) [pt

  13. Measurement of critical mass for an assembly of bare uranium shells

    International Nuclear Information System (INIS)

    Myers, W.L.; Goulding, C.A.; Hollas, C.L.

    1997-01-01

    As part of the research into nuclear measurement techniques, a series of measurements was performed that have applications to criticality safety and nuclear material handling. The critical mass of a set of bare, enriched-uranium metal hemispherical shells, known as the Rocky Flats shells, was measured for an assembly having an inside radius of 2.347 cm. The critical mass value was extrapolated from a series of subcritical measurements using three different kinds of sources (AmBe, AmF, and 252 Cf) placed at the center of the shells. Two kinds of neutron detection configurations (a 1% efficiency and a 25% efficiency configuration) were used to make the measurements

  14. Preliminary study for treatment methodology establishment of liquid waste containing uranium in refining facility lagoon

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Jik; Lee, Kune Woo; Won, Hui Jun; Ahn, Byung Gil; Shim, Joon Bo

    1999-12-01

    The preliminary study which establishes the treatment methodology of the sludge waste containing uranium in the conversion facility lagoon was performed. The property of lagoon liquid waste such as the initial water content, the density including radiochemical analysis results were obtained using the samples taken from the lagoon. The objective of this study is to provide some basically needed materials for selection of the most proper lagoon waste treatment methodology by reviewing the effective processes and methods for minimizing the secondary waste resulting from the treatment and disposition of large amount of radioactive liquid waste according to the facility closing. The lagoon waste can be classified into two sorts, such as supernatant and precipitate. The supernatants contain uranium less than 5 ppm and their water content are about 35 percent. Therefore, supernatants are solutions composed of mainly salt components. However, the precipitates have lots of uranium compound contained in the coagulation matrix, and are formed as two kinds of crystalline structures. The most proper method minimizing the secondary waste would be direct drying and solidification of the supernatants and precipitates after separation of them by filtering. (author)

  15. Benchmark critical experiments on low-enriched uranium oxide systems with H/U = 0.77

    International Nuclear Information System (INIS)

    Tuck, G.; Oh, I.

    1979-08-01

    Ten benchmark experiments were performed at the Critical Mass Laboratory at Rockwell International's Rocky Flats Plant, Golden, Colorado, for the US Nuclear Regulatory Commission. They provide accurate criticality data for low-enriched damp uranium oxide (U 3 O 8 ) systems. The core studied consisted of 152 mm cubical aluminum cans containing an average of 15,129 g of low-enriched (4.46% 235 U) uranium oxide compacted to a density of 4.68 g/cm 3 and with an H/U atomic ratio of 0.77. One hundred twenty five (125) of these cans were arranged in an approx. 770 mm cubical array. Since the oxide alone cannot be made critical in an array of this size, an enriched (approx. 93% 235 U) metal or solution driver was used to achieve criticality. Measurements are reported for systems having the least practical reflection and for systems reflected by approx. 254-mm-thick concrete or plastic. Under the three reflection conditions, the mass of the uranium metal driver ranged from 29.87 kg to 33.54 kg for an oxide core of 1864.6 kg. For an oxide core of 1824.9 kg, the weight of the high concentration (351.2 kg U/m 3 ) solution driver varied from 14.07 kg to 16.14 kg, and the weight of the low concentration (86.4 kg U/m 3 ) solution driver from 12.4 kg to 14.0 kg

  16. Parametric analyses of planned flowing uranium hexafluoride critical experiments

    Science.gov (United States)

    Rodgers, R. J.; Latham, T. S.

    1976-01-01

    Analytical investigations were conducted to determine preliminary design and operating characteristics of flowing uranium hexafluoride (UF6) gaseous nuclear reactor experiments in which a hybrid core configuration comprised of UF6 gas and a region of solid fuel will be employed. The investigations are part of a planned program to perform a series of experiments of increasing performance, culminating in an approximately 5 MW fissioning uranium plasma experiment. A preliminary design is described for an argon buffer gas confined, UF6 flow loop system for future use in flowing critical experiments. Initial calculations to estimate the operating characteristics of the gaseous fissioning UF6 in a confined flow test at a pressure of 4 atm, indicate temperature increases of approximately 100 and 1000 K in the UF6 may be obtained for total test power levels of 100 kW and 1 MW for test times of 320 and 32 sec, respectively.

  17. The role of public consultation in leading practice uranium mining

    International Nuclear Information System (INIS)

    Vance, R.

    2014-01-01

    facilities are performing as designed, the collection of baseline environmental data to objectively assess ecosystem impacts through the life of the mine by environmental monitoring programmes is essential to provide assurance of performance. Public consultation opportunities in presenting the results of the environmental monitoring programmes have proven critical to maintaining trust that the operations are performing as planned. Communication with neighbouring communities in decisions that affect them is critical to maintaining a social license to mine. An ongoing dialogue among the main stakeholders: the community, the mining company and the government has proven critical in this regard. With careful execution and participation of the main stakeholders, and adequate funding set aside for site remediation by mining companies, public funding for uranium mine site legacies and remediation should no longer be required and the goal of ensuring that no additional legacy uranium mining and milling issues are created will be assured. Included in this presentation are examples of public consultation in uranium mine operations that illustrate some of the challenges faced when undertaking public consultation and how such programmes can effectively increase public confidence and support of uranium mining, strengthening the social license to conduct the mining activity. (author)

  18. The structural innovative design of DN400 experimental facility with inversed U-shaped twin tower for uranium ion exchange

    International Nuclear Information System (INIS)

    Li Geng; Lei Ze'nan; Liu Qing

    2012-01-01

    A DN400 experimental facility with inversed U-shaped twin towers for uranium ion exchange is designed and developed to solve the existing problems in ion exchange equipment, such as easy agglomeration on the upper layer of resin bed, much resin abrasion, high sensitiveness for mud and sand, and difficulty in producing ideal plug flow in back-wash process. The facility combined with the valve control principle, cross flow type sieve plate and the unique inlet and outlet structures can solve the above problems, and the efficiency of uranium extraction can be improved significantly. (authors)

  19. Assessment of the effectiveness of personal visual observation as a safeguards measure in a uranium enrichment facility

    International Nuclear Information System (INIS)

    Ohno, Fubito; Okamoto, Tsuyoshi; Yokochi, Akira; Nidaira, Kazuo

    2003-01-01

    In a centrifuge enrichment facility, a cascade that produces low enriched uranium is composed of a large number of UF 6 gas centrifuges interconnected with pipes. It is possible to divert the cascade to the illegal production of highly enriched uranium (HEU) by changing the piping arrangement within the cascade. If integrated type centrifuges that contain a few tens of advanced centrifuges are introduced into the facility, the number of pipes will greatly decrease. The smaller the number of pipes, the less the labor required to change the piping arrangement. Because personal visual observation by an inspector is considered as one of measures against changing the piping arrangement, its effectiveness is assessed in this study. First, a model centrifuge enrichment facility that has a capacity of 2,400 ton-SWU/y is designed. In this model facility, integrated type centrifuges that contain advanced centrifuges are installed. Second, the diversion path analysis is carried out for the model facility under the assumption that a facility operator's goal is to produce 75 kg of HEU with 20% enrichment in a month. The analysis shows that, in our assumed diversion path, changes of the piping arrangement can be certainly detected by personal visual observation of a part of pipes connected with integrated type centrifuges that compose the cascade diverted to the HEU production. Finally, inspections in a cascade area are modeled as two-person noncooperative games between the inspector and the facility operator. As a result, it is found that all the cascades in the model facility will be investigated if the inspector can devote the inspection effort of 0.83 man-day per month to personal visual observation in the cascade area. Therefore, it is suggested that personal visual observation of the piping arrangement is worth carrying out in a uranium enrichment facility where integrated type centrifuges that contain advanced centrifuges are installed. (author)

  20. Validation of the Monte Carlo Criticality Program KENO V. a for highly-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Knight, J.R.

    1984-11-01

    A series of calculations based on critical experiments have been performed using the KENO V.a Monte Carlo Criticality Program for the purpose of validating KENO V.a for use in evaluating Y-12 Plant criticality problems. The experiments were reflected and unreflected systems of single units and arrays containing highly enriched uranium metal or uranium compounds. Various geometrical shapes were used in the experiments. The SCALE control module CSAS25 with the 27-group ENDF/B-4 cross-section library was used to perform the calculations. Some of the experiments were also calculated using the 16-group Hansen-Roach Library. Results are presented in a series of tables and discussed. Results show that the criteria established for the safe application of the KENO IV program may also be used for KENO V.a results.

  1. Validation of the Monte Carlo Criticality Program KENO V.a for highly-enriched uranium systems

    International Nuclear Information System (INIS)

    Knight, J.R.

    1984-11-01

    A series of calculations based on critical experiments have been performed using the KENO V.a Monte Carlo Criticality Program for the purpose of validating KENO V.a for use in evaluating Y-12 Plant criticality problems. The experiments were reflected and unreflected systems of single units and arrays containing highly enriched uranium metal or uranium compounds. Various geometrical shapes were used in the experiments. The SCALE control module CSAS25 with the 27-group ENDF/B-4 cross-section library was used to perform the calculations. Some of the experiments were also calculated using the 16-group Hansen-Roach Library. Results are presented in a series of tables and discussed. Results show that the criteria established for the safe application of the KENO IV program may also be used for KENO V.a results

  2. Criticality experiment for No.2 core of DF-VI fast neutron criticality facility

    International Nuclear Information System (INIS)

    Yang Lijun; Liu Zhenhua; Yan Fengwen; Luo Zhiwen; Chu Chun; Liang Shuhong

    2007-01-01

    At the completion of the DF-VI fast neutron criticality facility, its core changed, and it was restarted and a series of experiments and measurements were made. According to the data from 29 criticality experiments, the criticality element number and mass were calculated, the control rod reactivity worth were measured by period method and rod compensate method, reactivity worth of safety rod and safety block were measured using reactivity instrument; the reactivity worth of outer elements and radial distribution of elements were measured too. Based on all the measurements mentioned above, safety operation parameters for core 2 in DF-VI fast neutron criticality facility were conformed. (authors)

  3. A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

    2007-11-01

    The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.

  4. Analysis of nuclear grade uranium oxides by atomic absorption spectrometry with electrothermal atomization

    International Nuclear Information System (INIS)

    Batistoni, D.A.; Erlijman, L.H.; Pazos, A.L.

    1986-01-01

    The application of atomic absorption spectrometry for the determination of five trace impurities in nuclear grade uranium oxides is described. The elements were separated from the uranium matrix by extraction chromatography and determined in 5.5 M nitric acid by electrothermal atomization using pyrolytic graphite coated tubes. Two elements, cadmium and chromium, with different volatility characteristics were employed to investigate the operating conditions. Drying and ashing conditions were studied for both elements. Ramp and constant potential (step) heating modes have also been studied and compared. Good reproducibility and a longer life of graphite tubes were obtained with ramp atomization. Detection limits (in micrograms per gram of uranium) were: Cd 0.01; Cr 0.1; Cu 0.4; Mn 0.04 and Ni 0.2. (author) [es

  5. Criticality experiments with fast flux test facility fuel pins

    International Nuclear Information System (INIS)

    Bierman, S.R.

    1990-11-01

    A United States Department of Energy program was initiated during the early seventies at the Hanford Critical Mass Laboratory to obtain experimental criticality data in support of the Liquid Metal Fast Breeder Reactor Program. The criticality experiments program was to provide basic physics data for clean well defined conditions expected to be encountered in the handling of plutonium-uranium fuel mixtures outside reactors. One task of this criticality experiments program was concerned with obtaining data on PuO 2 -UO 2 fuel rods containing 20--30 wt % plutonium. To obtain this data a series of experiments were performed over a period of about twelve years. The experimental data obtained during this time are summarized and the associated experimental assemblies are described. 8 refs., 7 figs

  6. An automated, self-verifying system for monitoring uranium in effluent streams

    International Nuclear Information System (INIS)

    Reda, R.J.; Pickett, J.L.

    1992-01-01

    In nuclear facilities such as nuclear fuel fabrication plants, a constant vigil is required to ensure that the concentrations of uranium in process or waste streams do not exceed required specifications. The specifications may be dictated by the process owner, a regulatory agency such as the US Nuclear Regulatory Agency or Environmental Protection Agency, or by criticality safety engineering criteria. Traditionally, uranium monitoring in effluent streams has been accomplished by taking periodic samples of the liquid stream and determining the concentration by chemical analysis. Despite its accuracy, chemical sampling is not timely enough for practical use in continuously flowing systems because of the possibility that a significant quantity of uranium may be discharged between sampling intervals. To completely satisfy regulatory standards, the liquid waste stream must be monitored for uranium on a 100% basis. To this end, an automated, radioisotopic liquid-waste monitoring system was developed by GE Nuclear Energy as an integral part of the uranium conversion and waste recovery operations. The system utilizes passive gamma-ray spectroscopy and is thus a robust, on-line, and nondestructive assay for uranium. The system provides uranium concentration data for process monitoring and assures regulatory compliance for criticality safety. A summary of the principles of system operation, calibration, and verification is presented in this paper

  7. Surveillance and Maintenance Plan for the Plutonium Uranium Extraction (PUREX) Facility

    International Nuclear Information System (INIS)

    Woods, P.J.

    1998-05-01

    This document provides a plan for implementing surveillance and maintenance (S ampersand M) activities to ensure the Plutonium Uranium Extraction (PUREX) Facility is maintained in a safe, environmentally secure, and cost-effective manner until subsequent closure during the final disposition phase of decommissioning. This plan has been prepared in accordance with the guidelines provided in the U.S. Department of Energy (DOE), Office of Environmental Management (EM) Decommissioning Resource Manual (DOE/EM-0246) (DOE 1995), and Section 8.6 of TPA change form P-08-97-01 to the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) (Ecology, et al. 1996). Specific objectives of the S ampersand M program are: Ensure adequate containment of remaining radioactive and hazardous material. Provide security control for access into the facility and physical safety to surveillance personnel. Maintain the facility in a manner that will minimize potential hazards to the public, the environment, and surveillance personnel. Provide a plan for the identification and compliance with applicable environmental, safety, health, safeguards, and security requirements

  8. Determination of uranium and thorium contents inside different materials using track detectors and mean critical angles

    CERN Document Server

    Misdaq, M A; Ktata, A; Merzouki, A; Youbi, N

    1999-01-01

    The critical angles of the CR-39 (theta sub c) and LR-115 type II (theta sub c ') solid state nuclear track detectors (SSNTD) for detecting alpha-particles emitted by the uranium and thorium series have been evaluated by calculating the corresponding ranges of the emitted alpha-particles in different material samples and in the SSNTD studied. The influence of the emitted alpha-particles initial and residual energies on the critical angles of the SSNTD studied has been investigated. The uranium and thorium contents of different geological samples have been evaluated by exploiting data obtained for the critical angles of the CR-39 and LR-115 type II solid state nuclear track detectors and measuring the corresponding densities of tracks.

  9. CSER 94-012: Criticality safety evaluation report for 340 Facility

    International Nuclear Information System (INIS)

    Altschuler, S.J.

    1995-01-01

    This Criticality Safety Evaluation Report (CSER) covers the 340 Facility which acts as a collecting point for liquid and solid waste from various facilities in the 300 Area. Criticality safety is achieved by controlling the amount and concentration of the fissionable material sent to the 340 Facility from the originating facilities in the 300 Area, a method similar to that used elsewhere at Hanford for the waste tank farms. Unlike those, however, the waste received at the 340 Facility will be far less radioactive. It is concluded that present operations meet the two contingency criterion. The facility will still be safely subcritical even after two independent and concurrent failures (either of equipment or administrative controls). The solid waste storage and liquid waste will be managed separately. The solid waste storage area is classified as exempt because it contains less than 15 grams of fissionable materials. The Radioactive Liquid Waste System is classified as isolated because it contains less than one third of a minimum critical mass. The criticality safety of the 340 Facility devoted to the Radioactive Liquid Waste System (RLWS) is assured by the form and concentration of the fissile material and could also be classified as a limited control facility. However, the 340 Facility has been operated as an isolated facility which results in a more conservative limit

  10. Brazing graphite to graphite

    International Nuclear Information System (INIS)

    Peterson, G.R.

    1976-01-01

    Graphite is joined to graphite by employing both fine molybdenum powder as the brazing material and an annealing step that together produce a virtually metal-free joint exhibiting properties similar to those found in the parent graphite. Molybdenum powder is placed between the faying surfaces of two graphite parts and melted to form molybdenum carbide. The joint area is thereafter subjected to an annealing operation which diffuses the carbide away from the joint and into the graphite parts. Graphite dissolved by the dispersed molybdenum carbide precipitates into the joint area, replacing the molybdenum carbide to provide a joint of graphite

  11. Pursuit of improvement in uranium bulk analysis at the clear facility for safeguards environmental samples

    International Nuclear Information System (INIS)

    Sakurai, S.; Takahashi, M.; Sakakibara, T.; Magara, M.; Kurosawa, S.; Esaka, F.; Takai, K.; Watanabe, K.; Usuda, S.; Adachi, T.

    2002-01-01

    Full text: In order to contribute to the IAEA strengthened safeguards system, a project started in Japan Atomic Energy Research Institute (JAERI) in 1998. Consequently, a clean room facility called as CLEAR, the Clean Laboratory for Environmental Analysis and Research, was constructed in June 2001 at JAERI Tokai and the analytical techniques of ultra-trace nuclear materials in environmental samples are being developed. As for the bulk analysis, performance of inductively-coupled plasma mass spectrometry (ICP-MS) was mainly examined because sample preparation for ICP-MS is simpler than that for thermal ionization mass spectrometry (TIMS). Interference of polyatomic ion (such as PtAr + ) and coexisting element (such as Na) on the uranium ions, as well as mass bias caused by ICP-MS operating conditions, has been investigated for precise measurement on uranium isotope ratio. The authors have also studied on the uranium blanks during sample treatment process. The blank value below 10 pg uranium per sample treatment was obtained: dominant origins were elution from Teflon vessel surface in acid heating process of the sample to dry up. The work is in progress to minimize the blank. Compared with the process blank and the minimum uranium amount for isotope ratio measurement by ICP-MS (ca. 10 pg for natural uranium), the swipe cotton (Texwipe-304) which is currently used for IAEA Environmental Sampling includes much more amount of natural uranium in several nano-grams. If the amount of uranium collected on Texwipe-304 is small, sensitive and reliable measurement on isotope ratio will be impossible by bulk analysis. The authors are seeking alternative swipe materials with less amount of uranium. Recently, one of the authors devised an effective technique for recovery of uranium-containing particles from Texwipe-304. The technique, named as Vacuum Suction Method, uses a combination of polycarbonate membrane filters and a macro-pipette tip, which is connected to a vacuum pump

  12. Criticality safety for deactivation of the Rover dry headend process

    International Nuclear Information System (INIS)

    Henrikson, D.J.

    1995-01-01

    The Rover dry headend process combusted Rover graphite fuels in preparation for dissolution and solvent extraction for the recovery of 235 U. At the end of the Rover processing campaign, significant quantities of 235 U were left in the dry system. The Rover Dry Headend Process Deactivation Project goal is to remove the remaining uranium bearing material (UBM) from the dry system and then decontaminate the cells. Criticality safety issues associated with the Rover Deactivation Project have been influenced by project design refinement and schedule acceleration initiatives. The uranium ash composition used for calculations must envelope a wide range of material compositions, and yet result in cost effective final packaging and storage. Innovative thinking must be used to provide a timely safety authorization basis while the project design continues to be refined

  13. Utilization of AHWR critical facility for research and development work on large sample NAA

    International Nuclear Information System (INIS)

    Acharya, R.; Dasari, K.B.; Pujari, P.K.; Swain, K.K.; Reddy, A.V.R.; Verma, S.K.; De, S.K.

    2014-01-01

    The graphite reflector position of AHWR critical facility (CF) was utilized for analysis of large size (g-kg scale) samples using internal mono standard neutron activation analysis (IM-NAA). The reactor position was characterized by cadmium ratio method using In monitor for total flux and sub cadmium to epithermal flux ratio (f). Large sample neutron activation analysis (LSNAA) work was carried out for samples of stainless steel, ancient and new clay potteries and dross. Large as well as non-standard geometry samples (1 g - 0.5 kg) were irradiated. Radioactive assay was carried out using high resolution gamma ray spectrometry. Concentration ratios obtained by IM-NAA were used for provenance study of 30 clay potteries, obtained from excavated Buddhist sites of AP, India. Concentrations of Au and Ag were determined in not so homogeneous three large size samples of dross. An X-Z rotary scanning unit has been installed for counting large and not so homogeneous samples. (author)

  14. Control of uranium hazards - Portsmouth uranium enrichment plant

    International Nuclear Information System (INIS)

    Wagner, E.R.

    1985-01-01

    This report summarizes the Environmental, Safety and Health programs to control uranium hazards at the Portsmouth Gaseous Diffusion Plant. A description of the physical plant, the facility processes and the attendant uranium flows and effluents are provided. The hazards of uranium are discussed and the control systems are outlined. Finally, the monitoring programs are described and summaries of recent data are provided. 11 figs., 20 tabs

  15. Criticality accident in uranium fuel processing plant. The estimation of the total number of fissions with related reactor physics parameters

    International Nuclear Information System (INIS)

    Nishina, Kojiro; Oyamatsu, Kazuhiro; Kondo, Shunsuke; Sekimoto, Hiroshi; Ishitani, Kazuki; Yamane, Yoshihiro; Miyoshi, Yoshinori

    2000-01-01

    This accident occurred when workers were pouring a uranium solution into a precipitation tank with handy operation against the established procedure and both the cylindrical diameter and the total mass exceeded the limited values. As a result, nuclear fission chain reactor in the solution reached not only a 'criticality' state continuing it independently but also an instantly forming criticality state exceed the criticality and increasing further nuclear fission number. The place occurring the accident at this time was not reactor but a place having not to form 'criticality' called by a processing process of uranium fuel. In such place, as because of relating to mechanism of chain reaction, it is required naturally for knowledge on the reactor physics, it is also necessary to understand chemical reaction in chemical process, and functions of tanks, valves and pumps mounted at the processes. For this purpose, some information on uranium concentration ratio, atomic density of nuclides largely affecting to chain reaction such as uranium, hydrogen, and so forth in the solution, shape, inner structure and size of container for the solution, and its temperature and total volume, were necessary for determining criticality volume of the accident uranium solution by using nuclear physics procedures. Here were described on estimation of energy emission in the JCO accident, estimation from analytical results on neutron and solution, calculation of various nuclear physics property estimation on the JCO precipitation tank at JAERI. (G.K.)

  16. New insights into canted spiro carbon interstitial in graphite

    Science.gov (United States)

    EL-Barbary, A. A.

    2017-12-01

    The self-interstitial carbon is the key to radiation damage in graphite moderator nuclear reactor, so an understanding of its behavior is essential for plant safety and maximized reactor lifetime. The density functional theory is applied on four different graphite unit cells, starting from of 64 carbon atoms up to 256 carbon atoms, using AIMPRO code to obtain the energetic, athermal and mechanical properties of carbon interstitial in graphite. This study presents first principles calculations of the energy of formation that prove its high barrier to athermal diffusion (1.1 eV) and the consequent large critical shear stress (39 eV-50 eV) necessary to shear graphite planes in its presence. Also, for the first time, the gamma surface of graphite in two dimensions is calculated and found to yield the critical shear stress for perfect graphite. Finally, in contrast to the extensive literature describing the interstitial of carbon in graphite as spiro interstitial, in this work the ground state of interstitial carbon is found to be canted spiro interstitial.

  17. Graphite oxidation and structural strength of graphite support column in VHTR

    International Nuclear Information System (INIS)

    Park, Byung Ha; No, Hee Cheno; Kim, Eung Soo; Oh, Chang H.

    2009-01-01

    The air-ingress event by a large pipe break is an important accident considered in design of very high-temperature gas-cooled reactors (VHTR). Core-collapse prediction is a main safety issue. Structural failure model are technically required. The objective of this study is to develop structural failure model for the supporting graphite material in the lower plenum of the GT-MHR (gas-turbine-modular high temperature reactor). Graphite support column is important for VHTR structural integrity. Graphite support columns are under the axial load. Critical strength of graphite column is related to slenderness ratio and bulk density. Through compression tests for fresh and oxidized graphite columns we show that compressive strength of IG-110 was 79.46 MPa. And, the buckling strength of IG-110 column was expressed by the empirical formula: σ 0 =σ straight-line - C L/r, σ straight-line =91.31 MPa, C=1.01. The results of uniform and non-uniform oxidation tests show that the strength degradation of oxidized graphite column is expressed in the following non-dimensional form: σ/σ 0 =exp(-kd), k=0.111. Also, from the results of the uniform oxidation test with a complicated-shape column, we found out that the above non-dimensional equation obtained from the uniform oxidation test is applicable to a uniform oxidation case with a complicated-shape column. (author)

  18. Analysis of beryllium and depleted uranium: An overview of detection methods in aerosols and soils

    International Nuclear Information System (INIS)

    Camins, I.; Shinn, J.H.

    1988-06-01

    We conducted a survey of commercially available methods for analysis of beryllium and depleted uranium in aerosols and soils to find a reliable, cost-effective, and sufficiently precise method for researchers involved in environmental testing at the Yuma Proving Ground, Yuma, Arizona. Criteria used for evaluation include cost, method of analysis, specificity, sensitivity, reproducibility, applicability, and commercial availability. We found that atomic absorption spectrometry with graphite furnace meets these criteria for testing samples for beryllium. We found that this method can also be used to test samples for depleted uranium. However, atomic absorption with graphite furnace is not as sensitive a measurement method for depleted uranium as it is for beryllium, so we recommend that quality control of depleted uranium analysis be maintained by testing 10 of every 1000 samples by neutron activation analysis. We also evaluated 45 companies and institutions that provide analyses of beryllium and depleted uranium. 5 refs., 1 tab

  19. ORALLOY (93.15 235U) METAL ANNULI WITH BERYLLIUM CORE

    International Nuclear Information System (INIS)

    Bess, John D.; Montierth, Leland M.; Reed, Raymond L.; Mihalczo, John T.

    2010-01-01

    A variety of critical experiments were constructed of enriched uranium metal during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, two were performed that consisted of uranium metal annuli with a solid beryllium metal core. The outer diameter of the annuli was approximately 13 or 15 inches with an inner diameter of 7 inches. The diameter of the core was approximately 7 inches. The critical height of the configurations was approximately 5 and 4 inches, respectively. The uranium annuli consisted of multiple stacked rings with diametral thicknesses of approximately 2 inches apiece and varying heights. The 15-inch experiment was performed on June 4, 1963, and the 13-inch experiment on July 12, 1963 by J. T. Mihalczo and R. G. Taylor (Ref. 1) with accompanying logbook. Both detailed and simplified model specifications are provided in this evaluation. Both of these fast-spectra experiments were determined to represent acceptable benchmarks. The calculated eigenvalues for both the detailed and simple models are within approximately 0.6% of the benchmark values, but significantly greater than 3s from the benchmark value because the uncertainty in the benchmark is very small: eff of ∼0.67%. Unreflected and unmoderated experiments with the same highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in HEU MET

  20. Enriched uranium recovery at Los Alamos

    International Nuclear Information System (INIS)

    Herrick, C.C.

    1984-01-01

    Graphite casting scrap, fuel elements and nongraphite combustibles are calcined to impure oxides. These materials along with zircaloy fuel elements and refractory solids are leach-dissolved separately in HF-HNO 3 acid to solubilize the contained enriched uranium. The resulting slurry is filtered and the clear filtrate (to which mineral acid solutions bearing enriched uranium may be added) are passed through solvent extraction. The solvent extraction product is filtered, precipitated with H 2 O 2 and the precipitate calcined to U 3 O 8 . Metal is made from U 3 O 8 by conversion to UO 2 , hydrofluorination and reduction to metal. Throughput is 150 to 900 kg uranium per year depending on the type of scrap

  1. Accelerator-driven sub-critical research facility with low-enriched fuel in lead matrix: Neutron flux calculation

    Directory of Open Access Journals (Sweden)

    Avramović Ivana

    2007-01-01

    Full Text Available The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF being developed over the last couple of years at the Vinča Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutron flux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutron flux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step are used to specify a neutron source for further numerical simulations of the neutron flux density in the sub-critical core (second step. The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield for the selected sub-critical core cells for both beams have also been presented in this paper.

  2. Japan-IAEA sefeguards demonstration programme in the gas centrifuge uranium enrichment facility

    International Nuclear Information System (INIS)

    Akiba, Mitsunori; Iwamoto, Tomonori; Omae, Masayoshi

    1985-01-01

    The Hexa-partite Safequard Project was started for the purpose of examining the effective techniques of safeguards for gas centrifuge uranium enrichment facilities. By the proposal of respective participating countries, it was decided to carry out the verifying test of various safeguard techniques at the actual plants. Japan carried out the verifying test of safeguard techniques at the Ningyotoge uranium enrichment pilot plant in June, 1982, and from November, 1983, to August, 1984. The contents of this test is reported. In Japan, this verifying test was positioned as a part of JASPAS (Japanese project of supporting IAEA safeguards). The verifying test of realtime and in-operation inventories, the verifying test of IAEA load cell type weighing machines for UF 6 cylinders, the verifying test of the measurement of the degree of enrichment in UF 6 cylinders by nondestructive test, the verifying test of confinement/watch system, and the verifying test of IAEA gas phase uranium enrichment monitors were carried out. The results were presented as the data for examination in the HSP, and evaluated as useful, informative and well compiled. It is necessary to pursue more cost-effective approaches. (Kako, I.)

  3. Characterization of uranium in surface-waters collected at the Rocky Flats Facility

    International Nuclear Information System (INIS)

    Efurd, D.W.; Rokop, D.J.; Aguilar, R.D.; Roensch, F.R.; Perrin, R.E.; Banar, J.C.

    1994-01-01

    The Rocky Flats Plant (RFP) is a Department of Energy (DOE) facility where plutonium and uranium components were manufactured for nuclear weapons. During plant operations radioactivity was inadvertently released into the environment. This study was initiated to characterize the uranium present in surface-waters at RFP. Three drainage basins and natural ephemeral streams transverse RFP. The Woman Creek drainage basin traverses and drains the southern portion of the site. The Rock Creek drainage basin drains the northwestern portion of the plant complex. The Walnut Creek drainage basin traverses the western, northern, and northeastern portions of the RFP site. Dams, detention ponds, diversion structures, and ditches have been constructed at RFP to control the release of plant discharges and surface (storm water) runoff. The ponds located downstream of the plant complex on North Walnut Creek are designated A-1 through A-4. Ponds on South Walnut Creek are designated B-1 through B-5. The ponds in the Woman Creek drainage basin are designated C-1 and C-2. Water samples were collected from each pond and the uranium was characterized by TIMS measurement techniques

  4. Critical experiments of JMTRC MEU cores

    International Nuclear Information System (INIS)

    Nagaoka, Y.; Takeda, K.; Shimakawa, S.; Koike, S.; Oyamada, R.

    1984-01-01

    The JMTRC, the critical facility of the Japan Materials Testing Reactor (JMTR), went critical on August 29, 1983, with 14 medium enriched uranium (MEU, 45%) fuel elements. Experiments are now being carried out to measure the change in various reactor characteristics between the previous HEU core and the new MEU fueled core. This paper describes the results obtained thus far on critical mass, excess reactivity, control rod worths and flux distribution, including preliminary neutronics calculations for the experiments using the SRAC code. (author)

  5. Reducing emissions from uranium dissolving

    International Nuclear Information System (INIS)

    Griffith, W.L.; Compere, A.L.; Huxtable, W.P.; Googin, J.M.

    1992-10-01

    This study was designed to assess the feasibility of decreasing NO x emissions from the current uranium alloy scrap tray dissolving facility. In the current process, uranium scrap is dissolved in boiling nitric acid in shallow stainless-steel trays. As scrap dissolves, more metal and more nitric acid are added to the tray by operating personnel. Safe geometry is assured by keeping liquid level at or below 5 cm, the depth of a safe infinite slab. The accountability batch control system provides additional protection against criticality. Both uranium and uranium alloys are dissolved. Nitric acid is recovered from the vapors for reuse. Metal nitrates are sent to uranium recovery. Brown NO x fumes evolved during dissolving have occasionally resulted in a visible plume from the trays. The fuming is most noticeable during startup and after addition of fresh acid to a tray. Present environmental regulations are expected to require control of brown NO x emissions. A detailed review of the literature, indicated the feasibility of slightly altering process chemistry to favor the production of NO 2 which can be scrubbed and recycled as nitric acid. Methods for controlling the process to manage offgas product distribution and to minimize chemical reaction hazards were also considered

  6. Uranium hexafluoride purification

    International Nuclear Information System (INIS)

    Araujo, Eneas F. de

    1986-01-01

    Uranium hexafluoride might contain a large amount of impurities after manufacturing or handling. Three usual methods of purification of uranium hexafluoride were presented: selective sorption, sublimation, and distillation. Since uranium hexafluoride usually is contaminated with hydrogen fluoride, a theoretical study of the phase equilibrium properties was performed for the binary system UF 6 -HF. A large deviation from the ideal solution behaviour was observed. A purification unity based on a constant reflux batch distillation process was developed. A procedure was established in order to design the re boiler, condenser and packed columns for the UF 6 -HF mixture separation. A bench scale facility for fractional distillation of uranium hexafluoride was described. Basic operations for that facility and results extracted from several batches were discussed. (author)

  7. Methods for nondestructive assay holdup measurements in shutdown uranium enrichment facilities

    International Nuclear Information System (INIS)

    Hagenauer, R.C.; Mayer, R.L. II.

    1991-09-01

    Measurement surveys of uranium holdup using nondestructive assay (NDA) techniques are being conducted for shutdown gaseous diffusion facilities at the Oak Ridge K-25 Site (formerly the Oak Ridge Gaseous Diffusion Plant). When in operation, these facilities processed UF 6 with enrichments ranging from 0.2 to 93 wt % 235 U. Following final shutdown of all process facilities, NDA surveys were initiated to provide process holdup data for the planning and implementation of decontamination and decommissioning activities. A three-step process is used to locate and quantify deposits: (1) high-resolution gamma-ray measurements are performed to generally define the relative abundances of radioisotopes present, (2) sizable deposits are identified using gamma-ray scanning methods, and (3) the deposits are quantified using neutron measurement methods. Following initial quantitative measurements, deposit sizes are calculated; high-resolution gamma-ray measurements are then performed on the items containing large deposits. The quantitative estimates for the large deposits are refined on the basis of these measurements. Facility management is using the results of the survey to support a variety of activities including isolation and removal of large deposits; performing health, safety, and environmental analyses; and improving facility nuclear material control and accountability records. 3 refs., 1 tab

  8. 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.

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

    International Nuclear Information System (INIS)

    Montierth, Leland M.

    2016-01-01

    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.

  10. Calculation of reactivity of control rods in graphite moderated reactors

    International Nuclear Information System (INIS)

    Nakata, H.

    1978-01-01

    A study about the method of calculation for the reactivity of control rods in graphite-moderated critical assemblies, is presented. The result of theoretical calculation, developed by super celles and Nordheim-Scalettar methods are compared with experimental results for the critical Assembly of General Atomic. The two methods are then applicable to reactivity calculation of the control rods of graphite moderated critical assemblies [pt

  11. Contaminant distributions at typical U.S. uranium milling facilities and their effect on remedial action decisions

    International Nuclear Information System (INIS)

    Hamp, S.; Dotson, P.W.

    1995-01-01

    Past operations at uranium processing sites throughout the US have resulted in local contamination of soils and ground water by radionuclides, toxic metals, or both. Understanding the origin of contamination and how the constituents are distributed is a basic element for planning remedial action decisions. This report describes the radiological and nonradiological species found in ground water at a typical US uranium milling facility. The report will provide the audience with an understanding of the vast spectrum of contaminants that must be controlled in planning solutions to the long-term management of these waste materials

  12. Development of a stable uranium recovery regulatory framework for uranium recovery activities in the United States

    International Nuclear Information System (INIS)

    Layton, M.C.; Abrams, C.E.

    2000-01-01

    The U.S. Nuclear Regulatory Commission (NRC) has historically regulated operations at all uranium and thorium recovery facilities under the authority of the Atomic Energy Act of 1954, as amended. Uranium recovery facilities are those plants, or portions of facilities that process uranium- or thorium-bearing material primarily for its source material content. The uranium recovery industry expressed some concerns over several aspects of the NRC's practices, as described in the NRC's guidance documents. In April 1998, the National Mining Association submitted a report to the Commission, that identified specific concerns with NRC's current position and guidance regarding concurrent jurisdiction at uranium mills; dual regulatory authority at in situ leach facilities; the use of mill tailings impoundments for disposal of radioactive material other than 11e.(2) byproduct material; and the ability to process alternate feed material at uranium mills. The NRC staff addressed most of these concerns in two SECY (staff recommendations) papers that were concurrently provided to the Commission, along with a SECY paper on a draft rulemaking plan relating to these and other issues. The issues addressed in these papers included a new rulemaking, disposal of materials other than 11 e.(2) byproduct material, processing of materials other than natural ores, and improved efficiency for regulating in situ leach uranium facilities. The Commission issued final policy decisions on these issues and directions for NRC staff to implement those decisions in July 2000. (author)

  13. From Core to Capture: Graphite Management by Gasification and Carbon Capture & Storage (CCS)

    International Nuclear Information System (INIS)

    Goodwin, J.; Bradbury, D.; Black, S.; Tomlinson, T.; Livesey, B.; Robinson, J.; Lindberg, M.; Newton, C.; Jones, A.; Wickham, A.

    2016-01-01

    Radioactive graphite waste arises principally from the moderators of graphite/gas-cooled reactors at the end of life of the reactors. Commercial power producing reactors (for example, Magnox, AGR and RBMK) have graphite moderators, each containing several thousand tonnes of graphite, with the UK having the largest inventory of over 90,000 tonnes. Additionally, there are smaller quantities of graphite arising from other sources such as fuel element components. The current long term strategy for management of reactor graphite in the UK is for these wastes to be conditioned for disposal followed by transfer to a geological disposal facility (GDF). With this baseline position, these wastes will account for about 30% of the ILW inventory in a GDF. As the volume of the graphite waste is so large, it is not currently economic to retrieve and process the graphite in advance of the availability of a geological disposal facility. Recent work by the NDA has ascribed a much smaller “incremental” volume of 2% due to graphite, calculated on the basis that the GDF has to be a certain size anyway in order to dissipate the decay heat from high level waste

  14. Extrapolated experimental critical parameters of unreflected and steel-reflected massive enriched uranium metal spherical and hemispherical assemblies

    International Nuclear Information System (INIS)

    Rothe, R.E.

    1997-12-01

    Sixty-nine critical configurations of up to 186 kg of uranium are reported from very early experiments (1960s) performed at the Rocky Flats Critical Mass Laboratory near Denver, Colorado. Enriched (93%) uranium metal spherical and hemispherical configurations were studied. All were thick-walled shells except for two solid hemispheres. Experiments were essentially unreflected; or they included central and/or external regions of mild steel. No liquids were involved. Critical parameters are derived from extrapolations beyond subcritical data. Extrapolations, rather than more precise interpolations between slightly supercritical and slightly subcritical configurations, were necessary because experiments involved manually assembled configurations. Many extrapolations were quite long; but the general lack of curvature in the subcritical region lends credibility to their validity. In addition to delayed critical parameters, a procedure is offered which might permit the determination of prompt critical parameters as well for the same cases. This conjectured procedure is not based on any strong physical arguments

  15. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

  16. Progress in radioactive graphite waste management

    International Nuclear Information System (INIS)

    2010-07-01

    Radioactive graphite constitutes a major waste stream which arises during the decommissioning of certain types of nuclear installations. Worldwide, a total of around 250 000 tonnes of radioactive graphite, comprising graphite moderators and reflectors, will require management solutions in the coming years. 14 C is the radionuclide of greatest concern in nuclear graphite; it arises principally through the interaction of reactor neutrons with nitrogen, which is present in graphite as an impurity or in the reactor coolant or cover gas. 3 H is created by the reactions of neutrons with 6 Li impurities in graphite as well as in fission of the fuel. 36 Cl is generated in the neutron activation of chlorine impurities in graphite. Problems in the radioactive waste management of graphite arise mainly because of the large volumes requiring disposal, the long half-lives of the main radionuclides involved and the specific properties of graphite - such as stored Wigner energy, graphite dust explosibility and the potential for radioactive gases to be released. Various options for the management of radioactive graphite have been studied but a generally accepted approach for its conditioning and disposal does not yet exist. Different solutions may be appropriate in different cases. In most of the countries with radioactive graphite to manage, little progress has been made to date in respect of the disposal of this material. Only in France has there been specific thinking about a dedicated graphite waste-disposal facility (within ANDRA): other major producers of graphite waste (UK and the countries of the former Soviet Union) are either thinking in terms of repository disposal or have no developed plans. A conference entitled 'Solutions for Graphite Waste: a Contribution to the Accelerated Decommissioning of Graphite Moderated Nuclear Reactors' was held at the University of Manchester 21-23 March 2007 in order to stimulate progress in radioactive graphite waste management

  17. Impact of uranium-233/thorium cycle on advanced accountability concepts and fabrication facilities. Addendum 2 to application of advanced accountability concepts in mixed oxide fabrication

    International Nuclear Information System (INIS)

    Bastin, J.J.; Jump, M.J.; Lange, R.A.; Crandall, C.C.

    1977-11-01

    The Phase I study of the application of advanced accountability methods (DYMAC) in a uranium/plutonium mixed oxide facility was extended to cover the possible fabrication of uranium-233/thorium fuels. Revisions to Phase II of the DYMAC plan which would be necessitated by such a process are specified. These revisions include shielding requirements, measurement systems, licensing conditions, and safeguards considerations. The impact of the uranium/thorium cycle on a large-scale fuel fabrication facility was also reviewed; it was concluded that the essentially higher radioactivity of uranium/thorium feeds would lead to increased difficulties which tend to preclude early commercial application of the process. An amended schedule for Phase II is included

  18. Natural uranium metallic fuel elements: fabrication and operating experience

    International Nuclear Information System (INIS)

    Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.

    1980-01-01

    The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)

  19. Facile synthesis of nanorod-type graphitic carbon nitride/Fe2O3 composite with enhanced photocatalytic performance

    International Nuclear Information System (INIS)

    Wang, Jiangpeng; Li, Changqing; Cong, Jingkun; Liu, Ziwei; Zhang, Hanzhuo; Liang, Mei; Gao, Junkuo; Wang, Shunli; Yao, Juming

    2016-01-01

    Here we report a facile synthesis of nanorod-type graphitic carbon nitride/Fe 2 O 3 composite (Fe 2 O 3 -g-C 3 N 4 ) by using Fe-melamine supramolecular framework as precursor. The chemical and optical properties of the nanocomposites are well-characterized. The Fe 2 O 3 -g-C 3 N 4 nanocomposite demonstrated excellent photocatalytic activities under visible light due to the efficient utilization of sunlight and the construction of Z-scheme electron transfer pathway. The results indicated that it could be a promising approach for the preparation of efficient g-C 3 N 4 nanocomposites photocatalysts by using metal-melamine supramolecular framework as precursors. - Graphical abstract: Nanorod-type graphitic carbon nitride/Fe 2 O 3 composite (Fe 2 O 3 -g-C 3 N 4 ) was synthesized by using Fe-melamine supramolecular framework as precursor. The Fe 2 O 3 -g-C 3 N 4 nanocomposite demonstrated excellent photocatalytic activities under visible light. Display Omitted - Highlights: • Nanorod-type graphitic carbon nitride/Fe 2 O 3 composite (Fe 2 O 3 -g-C 3 N 4 ) was synthesized. • Fe 2 O 3 -g-C 3 N 4 showed strong optical absorption in the visible-light region. • The Fe 2 O 3 -g-C 3 N 4 nanocomposite demonstrated excellent photocatalytic activities.

  20. Uranium retrieval support, storage, and marketing

    International Nuclear Information System (INIS)

    Jackson, J.D.; Marshall, E.M.

    2001-01-01

    The United States Department of Energy is implementing a stewardship approach to management of uranium assets. This life-cycle approach to managing uranium addresses current needs in the context of a long-term strategy. In June 1998, the United States Department of Energy established the Uranium Management Group. The mission of the UMG is to safely collect and store commercially viable uranium from various DOE facilities at a central location. The Oak Ridge Operations Office, in Oak Ridge, Tennessee, was given the task to establish a facility for the storage of these uranium materials. Materials collected are non-waste uranium and packaged to allow transport and long-term storage. Coordination of uranium management under the Uranium Management Group offers significant opportunities for sayings through improved planning and efficiency and creates an environmentally sound approach for the storage and reuse of excess uranium. (author)

  1. Uranium retrieval support, storage, and marketing

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, J.D.; Marshall, E.M. [U.S. Department of Energy, Oak Ridge, Tennessee (United States)

    2001-07-01

    The United States Department of Energy is implementing a stewardship approach to management of uranium assets. This life-cycle approach to managing uranium addresses current needs in the context of a long-term strategy. In June 1998, the United States Department of Energy established the Uranium Management Group. The mission of the UMG is to safely collect and store commercially viable uranium from various DOE facilities at a central location. The Oak Ridge Operations Office, in Oak Ridge, Tennessee, was given the task to establish a facility for the storage of these uranium materials. Materials collected are non-waste uranium and packaged to allow transport and long-term storage. Coordination of uranium management under the Uranium Management Group offers significant opportunities for sayings through improved planning and efficiency and creates an environmentally sound approach for the storage and reuse of excess uranium. (author)

  2. Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project

    International Nuclear Information System (INIS)

    2016-05-01

    Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants

  3. Criticality evaluations with moderators other than water for uranium metal fuels

    International Nuclear Information System (INIS)

    Toffer, H.; Tollefson, D.A.; Finfrock, S.H.

    1986-01-01

    Occasionally, nuclear criticality safety analyses of fissile material handling operations or transport situations require consideration of moderation other than water. Such moderators could be oils, plastics, wood, concrete, carbon, or even wet sand. All of these materials contain either hydrogen, carbon, or mixtures of the two elements as the principal moderators. Other elements as part of the compounds or mixtures contribute less to the neutron slowing down process and can possibly be significant parasitic neutron absorbers. Results of a series of calculations are presented illustrating the impact of various moderators on critical masses or critical parameters as a function of lattice pitch for different uranium metal fuel elements at low 235 U enrichments. Several nuclear criticality safety analyses performed at the Hanford N Reactor, operated by UNC Nuclear Industries for the US Department of Energy, have considered alternative moderators to assure that water moderation represented the most limiting case

  4. Reactivity of hydrogen with uranium in the presence of Pt

    International Nuclear Information System (INIS)

    Balooch, M.; Siekhaus, W.J.

    1997-07-01

    The surface-reaction of di-hydrogen with uranium in the presence of Pt clusters has been studied using scanning tunneling microscopy (STM). Uranium was deposited on highly oriented pyrolytic graphite (HOPG) and annealed at temperatures up to 1200 degrees C to obtain atomically pyrolytic flat surfaces. Pt clusters were then formed using evaporation from a Pt source onto the surface and subsequent annealing. Hydrogen mainly attacked uranium in the vicinity of Pt clusters and formed hydride. The hydride formation probability is almost constant at 2.3x10 -4 over the range of exposures studied

  5. High temperature tests for graphite materials

    OpenAIRE

    Zhmurikov, Evgenij

    2015-01-01

    This study was performed within the framework of the EURISOL for facilities SPIRAL-II (GANIL, France) and SPES (LNL, Italy), and aims to investigate the anticipated strength properties of fine-grained graphite at elevated temperatures. It appears that the major parameters that affect to the lifetime of a graphite target of this IP are the temperature and heating time. High temperature tests were conducted to simulate the heating under the influence of a beam of heavy particles by passing thro...

  6. Concepts for inventory verification in critical facilities

    International Nuclear Information System (INIS)

    Cobb, D.D.; Sapir, J.L.; Kern, E.A.; Dietz, R.J.

    1978-12-01

    Materials measurement and inventory verification concepts for safeguarding large critical facilities are presented. Inspection strategies and methods for applying international safeguards to such facilities are proposed. The conceptual approach to routine inventory verification includes frequent visits to the facility by one inspector, and the use of seals and nondestructive assay (NDA) measurements to verify the portion of the inventory maintained in vault storage. Periodic verification of the reactor inventory is accomplished by sampling and NDA measurement of in-core fuel elements combined with measurements of integral reactivity and related reactor parameters that are sensitive to the total fissile inventory. A combination of statistical sampling and NDA verification with measurements of reactor parameters is more effective than either technique used by itself. Special procedures for assessment and verification for abnormal safeguards conditions are also considered. When the inspection strategies and inventory verification methods are combined with strict containment and surveillance methods, they provide a high degree of assurance that any clandestine attempt to divert a significant quantity of fissile material from a critical facility inventory will be detected. Field testing of specific hardware systems and procedures to determine their sensitivity, reliability, and operational acceptability is recommended. 50 figures, 21 tables

  7. Safety analysis of the Los Alamos critical experiments facility

    International Nuclear Information System (INIS)

    Paxton, H.C.

    1975-10-01

    The safety of Pajarito Site critical assembly operations depends upon protection built into the facility, upon knowledgeable personnel, and upon good practice as defined by operating procedures and experimental plans. Distance, supplemented by shielding in some cases, would protect personnel against an extreme accident generating 10 19 fissions. During the facility's 28-year history, the direct cost of criticality accidents has translated to a risk of less than $200 per year

  8. Highly Enriched Uranium Metal Annuli and Cylinders with Polyethylene Reflectors and/or Internal Polyethylene Moderator

    International Nuclear Information System (INIS)

    Tyler Sumner; J. Blair Briggs; Leland Montierth

    2007-01-01

    A variety of critical experiments were constructed of enriched uranium metal during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, experiments of uranium metal annuli with and without polyethylene reflectors and with the central void region either empty or filled with polyethylene were evaluated under ICSBEP Identifier HEU-MET-FAST-076. The outer diameter of the uranium annuli varied from 9 to 15 inches in two-inch increments. In addition, there were uranium metal cylinders with diameters varying from 7 to 15 inches with complete reflection and reflection on one flat surface to simulate floor reflection. Most of the experiments were performed between February 1964 and April 1964. Five partially reflected (reflected on the top only) experiments were assembled in November 1967, but are judged by the evaluators not to be of benchmark quality. Twenty-four of the twenty-five experiments have been determined to have fast spectra. The only exception has a mixed spectrum. Analyses were performed in which uncertainty associated with five different parameters associated with the uranium parts and three associated with the polyethylene parts was evaluated. Included were uranium mass, height, diameter, isotopic content, and impurity content and polyethylene mass, diameter, and impurity content. There were additional uncertainties associated with assembly alignment, support structure, and the value

  9. Irradiation Creep in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  10. Dismantlement and Radioactive Waste Management of DPRK Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jooho, W.; Baldwin, G. T.

    2005-04-01

    One critical aspect of any denuclearization of the Democratic People’s Republic of Korea (DPRK) involves dismantlement of its nuclear facilities and management of their associated radioactive wastes. The decommissioning problem for its two principal operational plutonium facilities at Yongbyun, the 5MWe nuclear reactor and the Radiochemical Laboratory reprocessing facility, alone present a formidable challenge. Dismantling those facilities will create radioactive waste in addition to existing inventories of spent fuel and reprocessing wastes. Negotiations with the DPRK, such as the Six Party Talks, need to appreciate the enormous scale of the radioactive waste management problem resulting from dismantlement. The two operating plutonium facilities, along with their legacy wastes, will result in anywhere from 50 to 100 metric tons of uranium spent fuel, as much as 500,000 liters of liquid high-level waste, as well as miscellaneous high-level waste sources from the Radiochemical Laboratory. A substantial quantity of intermediate-level waste will result from disposing 600 metric tons of graphite from the reactor, an undetermined quantity of chemical decladding liquid waste from reprocessing, and hundreds of tons of contaminated concrete and metal from facility dismantlement. Various facilities for dismantlement, decontamination, waste treatment and packaging, and storage will be needed. The shipment of spent fuel and liquid high level waste out of the DPRK is also likely to be required. Nuclear facility dismantlement and radioactive waste management in the DPRK are all the more difficult because of nuclear nonproliferation constraints, including the call by the United States for “complete, verifiable and irreversible dismantlement,” or “CVID.” It is desirable to accomplish dismantlement quickly, but many aspects of the radioactive waste management cannot be achieved without careful assessment, planning and preparation, sustained commitment, and long

  11. Dismantlement and Radioactive Waste Management of DPRK Nuclear Facilities

    International Nuclear Information System (INIS)

    Jooho, W.; Baldwin, G.T.

    2005-01-01

    One critical aspect of any denuclearization of the Democratic People's Republic of Korea (DPRK) involves dismantlement of its nuclear facilities and management of their associated radioactive wastes. The decommissioning problem for its two principal operational plutonium facilities at Yongbyun, the 5MWe nuclear reactor and the Radiochemical Laboratory reprocessing facility, alone present a formidable challenge. Dismantling those facilities will create radioactive waste in addition to existing inventories of spent fuel and reprocessing wastes. Negotiations with the DPRK, such as the Six Party Talks, need to appreciate the enormous scale of the radioactive waste management problem resulting from dismantlement. The two operating plutonium facilities, along with their legacy wastes, will result in anywhere from 50 to 100 metric tons of uranium spent fuel, as much as 500,000 liters of liquid high-level waste, as well as miscellaneous high-level waste sources from the Radiochemical Laboratory. A substantial quantity of intermediate-level waste will result from disposing 600 metric tons of graphite from the reactor, an undetermined quantity of chemical decladding liquid waste from reprocessing, and hundreds of tons of contaminated concrete and metal from facility dismantlement. Various facilities for dismantlement, decontamination, waste treatment and packaging, and storage will be needed. The shipment of spent fuel and liquid high level waste out of the DPRK is also likely to be required. Nuclear facility dismantlement and radioactive waste management in the DPRK are all the more difficult because of nuclear nonproliferation constraints, including the call by the United States for 'complete, verifiable and irreversible dismantlement,' or 'CVID.' It is desirable to accomplish dismantlement quickly, but many aspects of the radioactive waste management cannot be achieved without careful assessment, planning and preparation, sustained commitment, and long completion times

  12. Validation of the Continuous-Energy Monte Carlo Criticality-Safety Analysis System MVP and JENDL-3.2 Using the Internationally Evaluated Criticality Benchmarks

    International Nuclear Information System (INIS)

    Mitake, Susumu

    2003-01-01

    Validation of the continuous-energy Monte Carlo criticality-safety analysis system, comprising the MVP code and neutron cross sections based on JENDL-3.2, was examined using benchmarks evaluated in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments'. Eight experiments (116 configurations) for the plutonium solution and plutonium-uranium mixture systems performed at Valduc, Battelle Pacific Northwest Laboratories, and other facilities were selected and used in the studies. The averaged multiplication factors calculated with MVP and MCNP-4B using the same neutron cross-section libraries based on JENDL-3.2 were in good agreement. Based on methods provided in the Japanese nuclear criticality-safety handbook, the estimated criticality lower-limit multiplication factors to be used as a subcriticality criterion for the criticality-safety evaluation of nuclear facilities were obtained. The analysis proved the applicability of the MVP code to the criticality-safety analysis of nuclear fuel facilities, particularly to the analysis of systems fueled with plutonium and in homogeneous and thermal-energy conditions

  13. Study of immobilization of waste from treatment of acid waters of a uranium mining facility

    International Nuclear Information System (INIS)

    Goda, R.T.; Oliveira, A.P. de; Silva, N.C. da; Villegas, R.A.S.; Ferreira, A.M.

    2017-01-01

    This study aimed to produce scientific and technical knowledge aiming at the development of techniques to immobilize the waste generated in the treatment of acid waters in the UTM-INB Caldas uranium mining and processing facility using Portland cement. This residue (calcium diuranate - DUCA) contains uranium compounds and metal hydroxides in a matrix of calcium sulfate. It is observed that this material, in contact with the lake of acid waters of the mine's own pit, undergoes resolubilization and, therefore, changes the quality of the acidic water contained therein, changing the treatment parameters. For the study of immobilization of this residue, the mass of water contained in both the residue deposited in the pit of the mine and in the pulp resulting from the treatment of the acid waters was determined. In addition, different DUCA / CEMENT / WATER ratios were used for immobilization and subsequent mechanical strength and leaching tests. The results showed that in the immobilized samples with 50% cement mass condition, no uranium was detected in the leaching tests, and the mechanical strength at compression was 9.4 MPa, which indicates that more studies are needed, but indicate a good capacity to immobilize uranium in cement

  14. Removal of uranium from uranium-contaminated soils -- Phase 1: Bench-scale testing. Uranium in Soils Integrated Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Francis, C. W.

    1993-09-01

    To address the management of uranium-contaminated soils at Fernald and other DOE sites, the DOE Office of Technology Development formed the Uranium in Soils Integrated Demonstration (USID) program. The USID has five major tasks. These include the development and demonstration of technologies that are able to (1) characterize the uranium in soil, (2) decontaminate or remove uranium from the soil, (3) treat the soil and dispose of any waste, (4) establish performance assessments, and (5) meet necessary state and federal regulations. This report deals with soil decontamination or removal of uranium from contaminated soils. The report was compiled by the USID task group that addresses soil decontamination; includes data from projects under the management of four DOE facilities [Argonne National Laboratory (ANL), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), and the Savannah River Plant (SRP)]; and consists of four separate reports written by staff at these facilities. The fundamental goal of the soil decontamination task group has been the selective extraction/leaching or removal of uranium from soil faster, cheaper, and safer than current conventional technologies. The objective is to selectively remove uranium from soil without seriously degrading the soil`s physicochemical characteristics or generating waste forms that are difficult to manage and/or dispose of. Emphasis in research was placed more strongly on chemical extraction techniques than physical extraction techniques.

  15. Criticality safety training at the Hot Fuel Examination Facility

    International Nuclear Information System (INIS)

    Garcia, A.S.; Courtney, J.C.; Thelen, V.N.

    1983-01-01

    HFEF comprises four hot cells and out-of-cell support facilities for the US breeder program. The HFEF criticality safety program includes training in the basic theory of criticality and in specific criticality hazard control rules that apply to HFEF. A professional staff-member oversees the implementation of the criticality prevention program

  16. Exposure of critical group of population to water radionuclides in area affected by uranium ore mining

    Energy Technology Data Exchange (ETDEWEB)

    Hladka, E; Zavadsky, M; Solnicka, H; Heroldova, J

    1985-08-01

    Waste waters from the uranium industry are decontaminated and then discharged into water courses. Inhabitants of the nearest village on the river form the critical group with regard to radiation burden. The critical radionuclides are Usub(nat), Ra 226, Pb 210 and Po 210 whose concentrations were determined in drinking water, in the water course and in plants watered with water from the river. From obtained data on the consumption of foods of own production and of water for drinking and cooking, a weighted sum was made of the intake of critical radionuclides per year on the conservative assumption that ingestion is the sole form of intake (permissible ingestion under Notice 59/72, Coll. of Laws). Under the said criteria the intake of radionuclides from water and foods of own production is for the critical population group 27 times less than the permissible intake for the population. Decontaminated waste waters from the operation of uranium industries contribute to the radiation burden of the population only negligibly. Radionuclides from the investigated sources represent a minute fraction of permissible intake.

  17. Some equipment for graphite research in swimming pool reactors; Quelques dispositifs d'etude du graphite dans les piles piscines

    Energy Technology Data Exchange (ETDEWEB)

    Seguin, M; Arragon, Ph; Dupont, G; Gentil, J; Tanis, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-07-01

    The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [French] Les dispositifs d'irradiation decrits servent aux etudes relatives a la filiere des reacteurs a uranium naturel, moderes au graphite et refroidis par le gaz carbonique. Ils sont generalement concus pour etre utilises dans des piles piscines. L'accent a ete mis sur: - l'utilisation au maximum du volume d'irradiation, - le recours aux solutions technologiques les plus simples, - la standardisation de certaines parties constitutives. Cette standardisation impose un usinage precis et un montage soigne, lesquels sont egalement necessaires lorsqu'on doit obtenir une temperature d'irradiation relativement basse alors que l'echauffement nucleaire est important. Enfin, la conception de ces dispositifs est valable pour irradier d'autres materiaux non fissiles ou fissiles. (auteurs)

  18. Advice about the safety of graphite storage silos of Saint Laurent des Eaux facility

    International Nuclear Information System (INIS)

    2005-01-01

    This document is the safety analysis made by the national association of the local commissions of information about nuclear activities (ANCLI), about the safety of graphite storage silos of Saint Laurent des Eaux nuclear facility. The analysis covers: the operation safety and the accident hypothesis, the monitoring of indoor and outdoor contamination in routine situation, the geotechnical characteristics of the site environment, the isotopic inventory and the estimation of radioactivity in routine and accidental situation, the estimation of doses received by the population in accidental situation and the internal emergency plan. After examination of these different points, the scientific committee of the ANCLI considers that a new global evaluation of risks, which integrates more recent exposure data, has to be carried out. (J.S.)

  19. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for

  20. Treatment of wastewater for removal of soluble uranium species at Cameco's Port Hope Conversion Facility

    International Nuclear Information System (INIS)

    Dumont, H.; Tairova, G.; Kwong, A.K.; Smith, B.D.

    2000-01-01

    Ion exchange (IX) resin processes have been used for many years in the uranium mining industry for the recovery of uranium from both acid and alkaline leach solutions. More recently, IX processes have been shown to be an effective approach to control the uranium levels in non-process waters, such as mine water, public drinking water supply and well water. Bench scale and mini-pilot plant tests were conducted at the Cameco's Port Hope Conversion Facility to demonstrate the economic and technical viability of an IX process as an uranium remediation treatment for trace amounts of uranium in non-process laundry water. In the mini-pilot plant study, waste laundry water containing between 10 mg U/L and 200 mg U/L was treated at a rate ranging from 120 L/h to 240 L/h, using a typical 'merry-go-round' fixed-bed ion exchange system with three ion exchange columns. Each column contained 14 L of strongly basic Purolite A300 resin type II. The results indicated that the breakthrough limit, set at 0.1 mg U/L was obtained after a minimum of 1,200 equivalent bed volumes, while saturation was obtained at 3,300 equivalent bed volumes. Recovery parameters are discussed along with feed and effluent stream quality and modifications to the upstream operation. (author)

  1. Leaching study of heavy and radioactive elements present in wastes discarded by a uranium extraction and processing facility

    International Nuclear Information System (INIS)

    Pihlak, A.; Lippmaa, E.; Maremaee, E.; Sirk, A; Uustalu, E.

    1995-08-01

    The present report provides a systematic leaching study of the waste depository at the Sillamaee metallurgical plant 'Silmet' (former uranium extraction and processing facility), its construction and environmental impact. The following data are presented: γ-activity data of the depository and two drill cores, chemical composition and physical properties of depository material and leaching waters, results of γ- and α-spectrometric studies, leaching (with demineralized and sea water) intensities of loparite and uranium ore processing waste components. Environmental danger presented by the Sillamaee waste dump to the Gulf of Finland and the surrounding environment in Estonia is mainly due to uranium leaching and the presence of a large array of chemically poisonous substances

  2. Hydrology of Quirke and Panel uranium tailings basins

    International Nuclear Information System (INIS)

    King, R.M.; MacPhie, L.

    1991-11-01

    The research project described by this report provides the AECB with an independent assessment of the 'saturated tailings concept' for the decommissioning of tailing areas at the Rio Algom Quirke and Panel uranium mines near Elliot Lake in Northern Ontario. Hydrologic and hydraulic modelling for each facility showed the interrelation between the design of the water-level control facilities and the water levels in each cell for design flood and extreme low-flow conditions, taking into account all water-balance components. The estimate of seepage rates through the tailings mass is identified as a critical issue

  3. Experimental facilities of Valduc critical station

    International Nuclear Information System (INIS)

    Mangin, D.; Maubert, L.

    1975-01-01

    The critical facility of Valduc and its experimentation possibilities are described. The different experimental programs carried out since 1962 are briefly reviewed. The last program involving a plutonium nitrate solution (18.9wt% 240 Pu) in a large parallelepipedic tank is presented and main results given [fr

  4. Collective modes in superconducting rhombohedral graphite

    Energy Technology Data Exchange (ETDEWEB)

    Kauppila, Ville [O.V. Lounasmaa Laboratory, Aalto University (Finland); Hyart, Timo; Heikkilae, Tero [University of Jyvaeskylae (Finland)

    2015-07-01

    Recently it was realized that coupling particles with a Dirac dispersion (such as electrons in graphene) can lead to a topologically protected state with flat band dispersion. Such a state could support superconductivity with unusually high critical temperatures. Perhaps the most promising way to realize such coupling in real materials is in the surface of rhombohedrally stacked graphite. We consider collective excitations (i.e. the Higgs modes) in surface superconducting rhombohedral graphite. We find two amplitude and two phase modes corresponding to the two surfaces of the graphite where the superconductivity lives. We calculate the dispersion of these modes. We also derive the Ginzburg-Landau theory for this material. We show that in superconducting rhombohedral graphite, the collective modes, unlike in conventional BCS superconductors, give a large contribution to thermodynamic properties of the material.

  5. Use of critical pathway models and log-normal frequency distributions for siting nuclear facilities

    International Nuclear Information System (INIS)

    Waite, D.A.; Denham, D.H.

    1975-01-01

    The advantages and disadvantages of potential sites for nuclear facilities are evaluated through the use of environmental pathway and log-normal distribution analysis. Environmental considerations of nuclear facility siting are necessarily geared to the identification of media believed to be sifnificant in terms of dose to man or to be potential centres for long-term accumulation of contaminants. To aid in meeting the scope and purpose of this identification, an exposure pathway diagram must be developed. This type of diagram helps to locate pertinent environmental media, points of expected long-term contaminant accumulation, and points of population/contaminant interface for both radioactive and non-radioactive contaminants. Confirmation of facility siting conclusions drawn from pathway considerations must usually be derived from an investigatory environmental surveillance programme. Battelle's experience with environmental surveillance data interpretation using log-normal techniques indicates that this distribution has much to offer in the planning, execution and analysis phases of such a programme. How these basic principles apply to the actual siting of a nuclear facility is demonstrated for a centrifuge-type uranium enrichment facility as an example. A model facility is examined to the extent of available data in terms of potential contaminants and facility general environmental needs. A critical exposure pathway diagram is developed to the point of prescribing the characteristics of an optimum site for such a facility. Possible necessary deviations from climatic constraints are reviewed and reconciled with conclusions drawn from the exposure pathway analysis. Details of log-normal distribution analysis techniques are presented, with examples of environmental surveillance data to illustrate data manipulation techniques and interpretation procedures as they affect the investigatory environmental surveillance programme. Appropriate consideration is given these

  6. Critical management issues for the Uranium Mill Tailings Remedial Action (UMTRA) Project

    International Nuclear Information System (INIS)

    Themelis, J.G.; Krishnan, K.R.

    1985-01-01

    The Uranium Mill Tailings Radiation Control Act of 1978 (PL95-604) authorized the Secretary of Energy to enter into cooperative agreements with certain states and Indian Tribes to clean up 24 inactive uranium mill tailing sites and associated vicinity properties. The Uranium Mill Tailings Remedial Action (UMTRA) Project includes the three Federal agencies (EPA, DOE, and NRC), eleven state, Indian Tribes, and at least four major contractors. The UMTRA Project extends over a period of ten years. The standards for the Project require a design life of 1000 years with a minimum performance period of 200 years. This paper discusses the critical management issues in dealing with the UMTRA Project and identifies the development of solutions for many of those issues. The highlights to date are promulgation of EPA standards, continued support from Congress and participating states and Indian Tribes, significant leadership shown at all levels, establishment of credibility with the public, and continued motivation of the team. The challenge for tomorrow is making certain NRC will license the sites and maintaining the high level of coordination exhibited to date to assure Project completion on schedule

  7. Present situation of unused uranium fuel in Tokyo Institute of Technology

    International Nuclear Information System (INIS)

    Obara, T.; Ogawa, M.

    2008-01-01

    Present situation of unused enriched uranium fuel in Tokyo Institute of Technology is described. The fuels were for sub-critical experiments. There is no special facility for transportation in the site. But there is no technical problem for it. One of the important issues to be done is a duty by national regulation against nuclear disaster. (author)

  8. Large sample NAA of a pottery replica utilizing thermal neutron flux at AHWR critical facility and X-Z rotary scanning unit

    International Nuclear Information System (INIS)

    Acharya, R.; Dasari, K.B.; Pujari, P.K.; Swain, K.K.; Shinde, A.D.; Reddy, A.V.R.

    2013-01-01

    Large sample neutron activation analysis (LSNAA) of a clay pottery replica from Peru was carried out using low neutron flux graphite reflector position of Advanced Heavy Water Reactor (AHWR) critical facility. This work was taken up as a part of inter-comparison exercise under IAEA CRP on LSNAA of archaeological objects. Irradiated large size sample, placed on an X-Z rotary scanning unit, was assayed using a 40% relative efficiency HPGe detector. The k 0 -based internal monostandard NAA (IM-NAA) in conjunction with insitu relative detection efficiency was used to calculate concentration ratios of 12 elements with respect to Na. Analyses of both small and large size samples were carried out to check homogeneity and to arrive at absolute concentrations. (author)

  9. CSER-98-002: Criticality analysis for the storage of special nuclear material sources and standards in the WRAP Facility

    International Nuclear Information System (INIS)

    Goldberg, H.J.

    1998-01-01

    The Waste Receiving and Processing (WRAP) Facility will store uranium and transuranic (TRU) sources and standards for certification that WRAP meets the requirements of the Quality Assurance Program Plan (QAPP) for the Waste Isolation Pilot Plant (WIPP). In addition, WRAP must meet internal requirements for testing and validation of measuring instruments for nondestructive assay (NDA). In order to be certified for WIPP, WRAP will participate in the NDA Performance Demonstration Program (PDP). This program is a blind test of the NDA capabilities for TRU waste. It is intended to ensure that the NDA capabilities of this facility satisfy the requirements of the quality assurance program plan for the WIPP. The PDP standards have been provided by the Los Alamos National Laboratory (LANL) for this program. These standards will be used in the WRAP facility. To internally check the accuracy and sensitivity of the NDA instruments, a further set of sources and standards will also be used by the facility. Each sealed source or standard will be referred to herein as a unit. Various combinations of these units will be placed in test drums and/or boxes which will be subject to their own limits until unloaded. There will be two sealed test drums with five grams of weapons grade plutonium loaded in them. These drums will be appropriately marked and will be subject to the unit limits rather than the drum limits. This analysis shows that the storage and use of special nuclear material sources and standards within the limited control facility of WRAP (Rooms 101 and 104) is safe from a criticality standpoint. With the form, geometry, and masses involved with this evaluation, a criticality is not possible. The limits given in Section 2 should be imposed on facility operations

  10. Criticality safety research on nuclear fuel cycle facility

    Energy Technology Data Exchange (ETDEWEB)

    Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2004-07-01

    This paper present d s current status and future program of the criticality safety research on nuclear fuel cycle made by Japan Atomic Energy Research Institute. Experimental research on solution fuel treated in reprocessing plant has been performed using two critical facilities, STACY and TRACY. Fundamental data of static and transient characteristics are accumulated for validation of criticality safety codes. Subcritical measurements are also made for developing a monitoring system for criticality safety. Criticality safety codes system for solution and power system, and evaluation method related to burnup credit are developed. (author)

  11. Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Matsuo, Hideto [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokyo (Japan)

    2002-12-01

    The graphite moderated gas cooled reactor operated by the Japan Atomic Power Company was stopped its commercial operation on March 1998, and the decommissioning process has been started. Graphite material is often used as the moderator and the reflector materials in the core of the gas cooled reactor. During the operation, a long life nuclide of {sup 14}C is generated in the graphite by several transmutation reactions. Separation of {sup 14}C isotope and the development of the separation method have been recognized to be critical issues for the decommissioning of the reactor core. To understand the current methodologies for the carbon isotope separation, literature on the subject was surveyed. Also, those on the physical and chemical behavior of {sup 14}C were surveyed. This is because the larger part of the nuclides in the graphite is produced from {sup 14}N by (n,p) reaction, and the location of them in the material tends to be different from those of the other carbon atoms. This report summarizes the result of survey on the open literature about the behavior of {sup 14}C and the separation methods, including the list of the literature on these subjects. (author)

  12. A graphite foam reinforced by graphite particles

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, J.J.; Wang, X.Y.; Guo, L.F.; Wang, Y.M.; Wang, Y.P.; Yu, M.F.; Lau, K.T.T. [DongHua University, Shanghai (China). College of Material Science and Engineering

    2007-11-15

    Graphite foam was obtained after carbonization and graphitization of a pitch foam formed by the pyrolysis of coal tar based mesophase pitch mixed with graphite particles in a high pressure and temperature chamber. The graphite foam possessed high mechanical strength and exceptional thermal conductivity after adding the graphite particles. Experimental results showed that the thermal conductivity of modified graphite foam reached 110W/m K, and its compressive strength increased from 3.7 MPa to 12.5 MPa with the addition of 5 wt% graphite particles. Through the microscopic observation, it was also found that fewer micro-cracks were formed in the cell wall of the modified foam as compared with pure graphite foam. The graphitization degree of modified foam reached 84.9% and the ligament of graphite foam exhibited high alignment after carbonization at 1200{sup o}C for 3 h and graphitization at 3000{sup o}C for 10 min.

  13. National Uranium Resource Evaluation: Spartanburg Quadrangle, South Carolina and North Carolina

    International Nuclear Information System (INIS)

    Schot, E.H.; Galipeau, J.M.

    1980-11-01

    The Spartanburg Quadrangle, South Carolina and North Carolina, was evaluated for uranium favorability using National Uranium Resource Evaluation criteria. The evaluation included the study and analysis of published and collected geologic, geophysical, and geochemical data from subsurface, surface, and aerial studies. Five environments are favorable for uranium deposits. The Triassic Wadesboro Basin has ground waters with anomalously high uranium concentrations and uranium-to-conductivity ratios. The Upper Cretaceous Tuscaloosa-Middendorf Formation is near a uranium source and has sediments favorable for uranium deposition. The contact-metamorphic aureoles associated with the Liberty Hill-Kershaw and Winnsboro-Rion plutonic complexes are close to uranium sources and contain the reductants (sulfides, graphite) necessary for precipitation. The East Fork area in the Charlotte Belt has ground waters with uranium concentrations 4 to 132 times the mean concentration reported for the surrounding Piedmont area. Unfavorable environments include the Catawba Granite, the area west of the Winnsboro-Rion complex, gold-quartz veins, the vermiculite district, and the Western Monazite Belt

  14. Method of preventing criticality of fresh fuel assembly in storage facility

    International Nuclear Information System (INIS)

    Kawamura, Makoto.

    1990-01-01

    With an aim of improving the operation efficiency of a reactor, extention of the operation cycle by increasing U 235 enrichment degree of fuel uranium is planned. However, along with the increase of the enrichment degree of the fuel uranium, there occurs a problem of criticality upon fuel handling. Then, in the present invention, boric acid incorporating B-10 of great neutron absorption effect are packed with water soluble polymeric materials which are further packed with a fuel packing sheet, or the water soluble polymeric materials incorporating boric acids are packed with fuel packing sheets which are disposed to a fresh fuel assembly and stored in a store house as they are. The fuel packing sheet is a perforated sheet having a plurality of water intruding pores. Then, if water should intrude to the store house accidentally, the water soluble polymeric materials are dissolved, so that the intruded water is converted into aqueous boric acid easily and absorbs neutrons effectively to thereby attain the prevention of criticality. (T.M.)

  15. Graphite

    Science.gov (United States)

    Robinson, Gilpin R.; Hammarstrom, Jane M.; Olson, Donald W.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Graphite is a form of pure carbon that normally occurs as black crystal flakes and masses. It has important properties, such as chemical inertness, thermal stability, high electrical conductivity, and lubricity (slipperiness) that make it suitable for many industrial applications, including electronics, lubricants, metallurgy, and steelmaking. For some of these uses, no suitable substitutes are available. Steelmaking and refractory applications in metallurgy use the largest amount of produced graphite; however, emerging technology uses in large-scale fuel cell, battery, and lightweight high-strength composite applications could substantially increase world demand for graphite.Graphite ores are classified as “amorphous” (microcrystalline), and “crystalline” (“flake” or “lump or chip”) based on the ore’s crystallinity, grain-size, and morphology. All graphite deposits mined today formed from metamorphism of carbonaceous sedimentary rocks, and the ore type is determined by the geologic setting. Thermally metamorphosed coal is the usual source of amorphous graphite. Disseminated crystalline flake graphite is mined from carbonaceous metamorphic rocks, and lump or chip graphite is mined from veins in high-grade metamorphic regions. Because graphite is chemically inert and nontoxic, the main environmental concerns associated with graphite mining are inhalation of fine-grained dusts, including silicate and sulfide mineral particles, and hydrocarbon vapors produced during the mining and processing of ore. Synthetic graphite is manufactured from hydrocarbon sources using high-temperature heat treatment, and it is more expensive to produce than natural graphite.Production of natural graphite is dominated by China, India, and Brazil, which export graphite worldwide. China provides approximately 67 percent of worldwide output of natural graphite, and, as the dominant exporter, has the ability to set world prices. China has significant graphite reserves, and

  16. Nuclear Criticality Safety Data Book

    Energy Technology Data Exchange (ETDEWEB)

    Hollenbach, D. F. [Y-12 National Security Complex, Oak Ridge, TN (United States)

    2016-11-14

    The objective of this document is to support the revision of criticality safety process studies (CSPSs) for the Uranium Processing Facility (UPF) at the Y-12 National Security Complex (Y-12). This design analysis and calculation (DAC) document contains development and justification for generic inputs typically used in Nuclear Criticality Safety (NCS) DACs to model both normal and abnormal conditions of processes at UPF to support CSPSs. This will provide consistency between NCS DACs and efficiency in preparation and review of DACs, as frequently used data are provided in one reference source.

  17. Nuclear Criticality Safety Data Book

    International Nuclear Information System (INIS)

    Hollenbach, D. F.

    2016-01-01

    The objective of this document is to support the revision of criticality safety process studies (CSPSs) for the Uranium Processing Facility (UPF) at the Y-12 National Security Complex (Y-12). This design analysis and calculation (DAC) document contains development and justification for generic inputs typically used in Nuclear Criticality Safety (NCS) DACs to model both normal and abnormal conditions of processes at UPF to support CSPSs. This will provide consistency between NCS DACs and efficiency in preparation and review of DACs, as frequently used data are provided in one reference source.

  18. Uranium Task Force final report

    International Nuclear Information System (INIS)

    1991-03-01

    Site-specific data on the management of uranium of 17 facilities have been assembled and analyzed to develop a comprehensive report on uranium processes, treatment, storage, and disposal on a Department of Energy-wide basis. By integrating a variety of waste generation sources, treatment processes, storage facilities, and disposal options, this waste management system study aims to effectively characterize and evaluate the performance and effectiveness of the total Department of Energy system for the management of uranium, as well as the individual sites. 7 refs., 7 figs., 2 tabs

  19. Reactor Sharing at Rensselaer Critical Facility

    International Nuclear Information System (INIS)

    D. Steiner, D. Harris, T. Trumbull

    2006-01-01

    This final report summarizes the reactor sharing activities at the Rensselaer Critical Facility. An example of a typical tour is also included. Reactor sharing at the RCF brings outside groups into the facility for a tour, an explanation of reactor matters, and a reactor measurement. It has involved groups ranging from high school classes to advanced college groups and in size from a few to about 50 visitors. The RCF differs from other university reactors in that its fuel is like that of large power reactors, and its research and curriculum are dedicated to power reactor matters

  20. Deuterium migration in nuclear graphite: consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste

    International Nuclear Information System (INIS)

    Le-Guillou, Mael

    2014-01-01

    In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2 H( 3 He,p) 4 He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300 C, and would be more efficient in dry inert gas than in humid gas. (author)

  1. Long-term testing of HTR fuel elements in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Nickel, H.

    1986-12-01

    The extensive results from irradiation experiments carried out on coated particles, on graphitic matrices of different composition and on integral fuel elements have shown that the spherical fuel elements with high-enriched uranium/thorium mixed-oxide particles and optimized graphitic matrix are available for use in the planned HTR facilities. A concentrated qualification programme is on the way in order to bring the fuel elements with particles from low-enriched uranium dioxide (LEU) and TRISO coating to a comparable level of experience and knowledge, i.e. to make them licensable for the planned HTR facilities. (orig.) [de

  2. Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

    International Nuclear Information System (INIS)

    Watanabe, Norio; Tamaki, Hitoshi

    2000-04-01

    On September 30, 1999, a criticality accident occurred at the Tokai-mura uranium processing plant operated by JCO Co., Ltd., 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. There have now been 21 criticality accidents reported in nuclear fuel processing facilities in foreign countries: seven in the United States, one in the United Kingdom and thirteen in Russia. Most of them occurred during the period from mid-1950's to mid-1960's, but one criticality accident tool place in Russian in 1997. This report reviews and compiles the published information on these accidents, including the latest information, focusing on the event sequence, the consequence of accident, and the cause of accident. The observations from the reviews are summarized as follows: Twenty of the 21 accidents occurred with the fissile material in a liquid. Twenty of the 21 accidents occurred in vessels/tanks with unfavorable geometry but one occurred in the vessel with favorable geometry. There were seven fatalities that were involved in five accidents. Three accidents involved a re-criticality condition caused by inadequate operator actions and two of them led to the death of the operators. One accident reached a re-criticality condition several hours after the first excursion was terminated by injecting borated water into the affected vessel. This accident implies the possibility that the borated water injection might not be effective to the criticality termination due to solubility of boric acid. Mechanisms of the criticality termination vary as follows: ejection or splashing of the solution at the time of power excursion, boiling or evaporation, addition of neutron poisons, or manual draining of solutions. (author)

  3. Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Norio [Planning and Analysis Division, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Tamaki, Hitoshi [Department of Safety Research Technical Support, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2000-04-01

    On September 30, 1999, a criticality accident occurred at the Tokai-mura uranium processing plant operated by JCO Co., Ltd., 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. There have now been 21 criticality accidents reported in nuclear fuel processing facilities in foreign countries: seven in the United States, one in the United Kingdom and thirteen in Russia. Most of them occurred during the period from mid-1950's to mid-1960's, but one criticality accident tool place in Russian in 1997. This report reviews and compiles the published information on these accidents, including the latest information, focusing on the event sequence, the consequence of accident, and the cause of accident. The observations from the reviews are summarized as follows: Twenty of the 21 accidents occurred with the fissile material in a liquid. Twenty of the 21 accidents occurred in vessels/tanks with unfavorable geometry but one occurred in the vessel with favorable geometry. There were seven fatalities that were involved in five accidents. Three accidents involved a re-criticality condition caused by inadequate operator actions and two of them led to the death of the operators. One accident reached a re-criticality condition several hours after the first excursion was terminated by injecting borated water into the affected vessel. This accident implies the possibility that the borated water injection might not be effective to the criticality termination due to solubility of boric acid. Mechanisms of the criticality termination vary as follows: ejection or splashing of the solution at the time of power excursion, boiling or evaporation, addition of neutron poisons, or manual draining of solutions. (author)

  4. Civilian inventories of plutonium and highly enriched uranium

    International Nuclear Information System (INIS)

    Albright, D.

    1987-01-01

    In the future, commercial laser isotope enrichment technologies, currently under development, could make it easier for national to produce highly enriched uranium secretly. The head of a US firm that is developing a laser enrichment process predicts that in twenty years, major utilities and small countries will have relatively small, on-site, laser-based uranium enrichment facilities. Although these plants will be designed for the production of low enriched uranium, they could be modified to produce highly enriched uranium, an option that raises the possibility of countries producing highly enriched uranium in small, easily hidden facilities. Against this background, most of this report describes the current and future quantities of plutonium and highly enriched uranium in the world, their forms, the facilities in which they are produced, stored, and used, and the extent to which they are transported. 5 figures, 10 tables

  5. Detailed description of an SSAC at the facility level for a low-enriched uranium conversion and fuel fabrication facility

    International Nuclear Information System (INIS)

    Jones, R.J.

    1984-09-01

    Some States have expressed a need for more detailed guidance with regard to the technical elements in the design and operation of SSACs for both the national and the international objectives. To meet this need the present document has been prepared, describing the technical elements of an SSAC in considerable detail. The purpose of this document is therefore, to provide a detailed description of a system for the accounting for and control of nuclear material in a model low enriched uranium conversion and fuel fabrication facility which can be used by a facility operator to establish his own system in a way which will provide the necessary information for compliance with a national system for nuclear material accounting and control and for the IAEA to carry out its safeguards responsibilities

  6. High temperature soldering of graphite

    International Nuclear Information System (INIS)

    Anikin, L.T.; Kravetskij, G.A.; Dergunova, V.S.

    1977-01-01

    The effect is studied of the brazing temperature on the strength of the brazed joint of graphite materials. In one case, iron and nickel are used as solder, and in another, molybdenum. The contact heating of the iron and nickel with the graphite has been studied in the temperature range of 1400-2400 ged C, and molybdenum, 2200-2600 deg C. The quality of the joints has been judged by the tensile strength at temperatures of 2500-2800 deg C and by the microstructure. An investigation into the kinetics of carbon dissolution in molten iron has shown that the failure of the graphite in contact with the iron melt is due to the incorporation of iron atoms in the interbase planes. The strength of a joint formed with the participation of the vapour-gas phase is 2.5 times higher than that of a joint obtained by graphite recrystallization through the carbon-containing metal melt. The critical temperatures are determined of graphite brazing with nickel, iron, and molybdenum interlayers, which sharply increase the strength of the brazed joint as a result of the formation of a vapour-gas phase and deposition of fine-crystal carbon

  7. Design and construction of a fast critical facility

    International Nuclear Information System (INIS)

    Kato, W.Y.; Dates, L.R.

    1962-01-01

    Design and construction of a fast critical facility. In a fast-power-reactor development programme, a critical facility is found to be a highly useful tool to ascertain calculational techniques, to verify neutron cross-section sets, and to obtain integral reactor-physics parameters necessary for the nuclear design of a power system. Since it is primarily a physics instrument, the design of a fast critical facility itself poses a number of different problems not found in the design of a power reactor. In addition to usual questions of site, containment, core design and instrumentation , there arise such problems as: how to obtain a large degree of flexibility consistent with safety, the determination of the size and type of facility to meet the experimental physics requirements, the determination of the number and location of control and safety rods minimizing perturbation effects and the specification of the reproducibility of control rods and other movable components to obtain the accuracy required in reactivity measurements. These are some of the problems which are discussed in this paper based on recent experience at the Argonne National Laboratory which has under construction a fast critical facility, ZPR-VI at its Lemont, Illinois site for fast-reactor-physics studies. The ZPR-VI is a movable half- or split-table-type machine similar to ZPR-III. It has a matrix about two and a half times the volume of the earlier machine and will be used to investigate the physics of large, highly dilute, metal and cermet, unmoderated and partially moderated systems having core volumes up to about 1500 l. A detailed description of the ZPR-VI with a discussion on the criteria used in the design of its various components from the point of view of reactor physics is presented. In addition, such topics as management and operating procedures, potential hazards during operation, experimental techniques to be used and construction costs are also included. (author) [fr

  8. Sharing of Rensselaer Polytechnic Institute Reactor Critical Facility (RCF)

    International Nuclear Information System (INIS)

    1995-01-01

    The RPI Reactor Critical Facility (RCF) operated successfully over the period fall 1994 - fall 1995. During this period, the RCF was used for Critical Reactor Laboratory spring 1995 (12 students); Reactor Operations Training fall 1994 (3 students); Reactor Operations Training spring 1995 (3 students); and Reactor Operations Training fall 1995 (3 students). Thirty-two Instrumentation and Measurement students used the RCF for one class for hands-on experiments with nuclear instruments. In addition, a total of nine credits of PhD thesis work were carried out at the RCF. This document constitutes the 1995 Report of the Rensselaer Polytechnic Institute's Reactor Critical Facility (RCF) to the USNRC, to the USDOE, and to RPI management

  9. Technical specifications for the Oak Ridge Critical Experiments Facility

    International Nuclear Information System (INIS)

    Stinnett, R.M.

    1986-01-01

    These Technical Specifications for the Oak Ridge Critical Experiments Facility (CEF) delineate limiting conditions of operation for the facility. The CEF is used primarily for testing the High Flux Isotope Reactor (HFIR) fuel assemblies. Specifically, the Criticality Testing Unit, Liquid (CTUL), located in the CEF, is used for the HFIR fuel assembly test. The test is performed to satisfy the surveillance requirements of the HFIR Technical Specifications. The test is used to determine the water-submerged shutdown margin for each fuel assembly. 11 refs

  10. Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors; Etude de la tenue de la gaine interne pour-element combustible a refroidissement interne et externe d'un reacteur graphite-gaz

    Energy Technology Data Exchange (ETDEWEB)

    Boudouresque, B; Courcon, P; Lestiboubois, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The cartridge of an internally and externally cooled annular fuel element used in gas-graphite reactors is made up of an uranium fuel tube, an external can and an internal can made of magnesium alloy. For the thermal exchange between the internal can and the fuel to be satisfactory, it is necessary for the can to stay in contact with the uranium under all temperature conditions. This report, based on a theoretical study, shows how the internal can fuel gap varies during the processes of canning, charging into the reactor and thermal cycling. The following parameters are considered: tube diameter, pressure of the heat carrying gas, gas entry temperature, plasticity of the can alloy. It is shown that for all operating conditions the internal can of a 77 x 95 element, planned for a gas-graphite reactor with a 40 kg/cm{sup 2} gas pressure, should remain in contact with the fuel. (authors) [French] La cartouche d'un element combustible annulaire, a refroidissement interne et externe pour reacteur graphite-gaz, est composee d'un tube combustible en uranium, d'une gaine externe et d'une gaine interne en alliage de magnesium. Pour que l'echange thermique entre la gaine interne et le combustible soit bon, il faut que la gaine reste appliquee sur l'uranium quel que soit le regime de temperature. Cette note a pour but de montrer comment, d'apres une etude theorique, le jeu combustible-gaine interne varie au cours des operations de gainage, de chargement dans le reacteur, et des cyclages thermiques. Les parametres suivants sont etudies: diametres de tube, pression du gaz caloporteur, temperature d'entree du gaz, plasticite de l'alliage de gaine. Il est montre que, quel que soit le regime de fonctionnement, la gaine interne d'un element 77 x 95, en projet pour un reacteur graphite-gaz sous pression de 40 kg/cm{sup 2}, doit rester appliquee sur le combustible. (auteurs)

  11. Source-driven noise analysis measurements with neptunium metal reflected by high enriched uranium

    International Nuclear Information System (INIS)

    Valentine, Timothy E.; Mattingly, John K.

    2003-01-01

    Subcritical noise analysis measurements have been performed with neptunium ( 237 Np) sphere reflected by highly enriched uranium. These measurements were performed at the Los Alamos Critical Experiment Facility in December 2002 to provide an estimate of the subcriticality of 237 Np reflected by various amounts of high-enriched uranium. This paper provides a description of the measurements and presents some preliminary results of the analysis of the measurements. The measured and calculated spectral ratios differ by 15% whereas the 'interpreted' and calculated k eff values differ by approximately 1%. (author)

  12. Transmutation of uranium and thorium in the particle field of the Quinta sub-critical assembly

    Science.gov (United States)

    Hashemi-Nezhad, S. R.; Asquith, N. L.; Voronko, V. A.; Sotnikov, V. V.; Zhadan, Alina; Zhuk, I. V.; Potapenko, A.; Husak, Krystsina; Chilap, V.; Adam, J.; Baldin, A.; Berlev, A.; Furman, W.; Kadykov, M.; Khushvaktov, J.; Kudashkin, I.; Mar'in, I.; Paraipan, M.; Pronskih, V.; Solnyshkin, A.; Tyutyunnikov, S.

    2018-03-01

    The fission rates of natural uranium and thorium were measured in the particle field of Quinta, a 512 kg natural uranium target-blanket sub-critical assembly. The Quinta assembly was irradiated with deuterons of energy 4 GeV from the Nuclotron accelerator of the Joint Institute for Nuclear Research (JINR), Dubna, Russia. Fission rates of uranium and thorium were measured using Gamma spectroscopy and fission track techniques. The production rate of 239Np was also measured. The obtained experimental results were compared with Monte Carlo predictions using the MCNPX 2.7 code employing the physics and fission-evaporation models of INCL4-ABLA, CEM03.03 and LAQGSM03.03. Some of the neutronic characteristics of the Quinta are compared with the "Energy plus Transmutation (EpT)" subcritical assembly, which is composed of a lead target and natU blanket. This comparison clearly demonstrates the importance of target material, neutron moderator and reflector types on the performance of a spallation neutron driven subcritical system. As the dimensions of the Quinta are very close to those of an optimal multi-rod-uranium target, the experimental and Monte Carlo calculation results presented in this paper provide insights on the particle field within a uranium target as well as in Accelerator Driven Systems in general.

  13. An experimental stack for the control of uranium enrichment; Empilement pour le controle de l'enrichissement de l'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Bailly du Bois, B; Raievski, V; Tretiakoff, O [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    An apparatus is described for controlling the enrichment of the uranium in a fuel element at all stages in its fabrication. The apparatus consists of a stack of graphite surrounded by cadmium, in which a thermal neutron density is maintained by a Ra-Be source placed in the stack. The fuel element to be controlled is placed in the stack at 16 cm from the cadmium wall and at 70 cm from the source. The fast neutrons produced in this fuel element are slowed down and detected by a BF{sub 3} counter placed in a second graphite stack adjacent to the first. The enrichment is obtained by comparing the counting rate measured with the element to be controlled, with that of a geometrically similar standard element, previously calibrated. (author) [French] On decrit un dispositif permettant de controler l'enrichissement de l'uranium contenu dans un element de combustible a tous les stades de sa fabrication. Le dispositif est constitue par un empilement de graphite entoure de cadnium dans lequel regne une densite de neutrons thermiques entretenus par une source de Ra-Be placee dans l'empilement. L'element de combustible a contr er est place dans cet empilement a 16 cm de la paroi de cadmium et a 70 cm de la source. Les neutrons rapides produits dans cet element de combustible sont ralentis et detectes par un compteur a BF{sub 3} place dans un deuxieme empilement de graphite contigu au premier. L'enrichissement est obtenu en comparant les taux de comptage mesures arec l'element a controler et un element standard geometriquement semblable, prealablement etalonne. (auteur)

  14. Criticality safety of storage barrels for enriched uranium fresh fuel at the RB research reactor

    International Nuclear Information System (INIS)

    Pesic, M. P.

    1997-01-01

    Study on criticality safety of fresh low and high enriched uranium (LEU and HEU) fuel elements in the storage/transport barrels at the RB research reactor is carried out by using the well-known MCNP computer code. It is shown that studied arrays of tightly closed fuel barrels, each entirely loaded with 100 fresh (HEU or LEU) fuel slugs, are far away from criticality, even in cases of an unexpected flooding by light water.(author)

  15. Experience with graphite in JET

    International Nuclear Information System (INIS)

    Pick, M.A.; Celentano, G.; Deksnis, E.; Dietz, K.J.; Shaw, R.; Sonnenberg, K.; Walravens, M.

    1987-01-01

    During the current operational period of JET more than 50% of the internal area of the machine is covered in graphite tiles. This includes the 15 m 2 of carbon tiles installed in the new toroidal limiter, the 40 poloidal belts of graphite tiles covering the U-joints and bellows as well as a two metre high ring (-- 20 m 2 ) or carbon tiles on the inner wall of the Torus. A ring of tiles in the equatorial plane (3 tiles high) consists of carbon-carbon fibre tiles. Test bed results indicated that the fine grained graphite tiles cracked at ∼ 1 kW/cm 2 for 2s of irradiation whereas the carbon-carbon fibre tiles were able to sustain a flux, limited by the irradiation facility, of 3.5 kW for 3s without any damage. The authors report on the generally positive experience they have had had with the installed graphite during the present and previous in-vessel configurations. This includes the physical integrity of the tiles under severe conditions such as high energy run-away electron beams, plasma disruptions and high heat fluxes. They report on the importance of the precise positioning of the inner wall and x-point tiles at the very high power fluxes of JET and the effect of deviations on both graphite and carbon-fibre tiles

  16. Los Alamos Critical Assemblies Facility

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1981-06-01

    The Critical Assemblies Facility of the Los Alamos National Laboratory has been in existence for thirty-five years. In that period, many thousands of measurements have been made on assemblies of 235 U, 233 U, and 239 Pu in various configurations, including the nitrate, sulfate, fluoride, carbide, and oxide chemical compositions and the solid, liquid, and gaseous states. The present complex of eleven operating machines is described, and typical applications are presented

  17. Disposition of PUREX facility tanks D5 and E6 uranium and plutonium solutions

    International Nuclear Information System (INIS)

    Harty, D.P.

    1993-12-01

    Approximately 9 kilograms of plutonium and 5 metric tons of uranium in a 1 molar nitric acid solution are being stored in two PUREX facility vessels, tanks D5 and E6. The plutonium was accumulated during cleanup activities of the plutonium product area of the PUREX facility. Personnel at PUREX recently completed a formal presentation to the Surplus Materials Peer Panel (SMPP) regarding disposition of the material currently in these tanks. The peer panel is a group of complex-wide experts who have been chartered by EM-64 (Office of Site and Facility Transfer) to provide a third party independent review of disposition decisions. The information presented to the peer panel is provided in the first section of this report. The panel was generally receptive to the information provided at that time and the recommendations which were identified

  18. Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project. Companion CD-ROM

    International Nuclear Information System (INIS)

    2016-05-01

    Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants

  19. Design Guide for Category I reactors critical facilities

    International Nuclear Information System (INIS)

    Brynda, W.J.; Powell, R.W.

    1978-08-01

    The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned critical facilities be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission

  20. Graphite moderator lifecycle behaviour. Proceedings of a specialists meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    The meeting provided the forum for graphite specialists representing operating and research organizations worldwide to exchange information in the following areas: the status of graphite development; operation and safety procedures for existing and future graphite moderated reactors; graphite testing techniques; review of the experiences gained and data acquired on the influence of neutron irradiation and oxidizing conditions on key graphite properties; and to exchange information useful for decommissioning activities. The participants provided twenty-seven papers on behalf of their countries and respective technical organizations. An open discussion followed each of the presentations. A consistently reoccurring theme throughout the specialists meeting was the noticeable reduction in the number of graphite experts remaining the nuclear power industry. Graphite moderated power reactors have provided a significant contribution to the generation of electricity throughout the past forty years and will continue to be a prominent energy source for the future. Yet, many of the renowned experts in the field of nuclear graphites are nearing the end of their careers without apparent replacement. This, coupled with changes in the focus on nuclear power by some industrialized countries, has prompted the IAEA to initiate an evaluation on the feasibility and interest by Member States of establishing a central archive facility for the storage of data on irradiated graphites. Refs, figs, tabs.

  1. Graphite moderator lifecycle behaviour. Proceedings of a specialists meeting

    International Nuclear Information System (INIS)

    1996-08-01

    The meeting provided the forum for graphite specialists representing operating and research organizations worldwide to exchange information in the following areas: the status of graphite development; operation and safety procedures for existing and future graphite moderated reactors; graphite testing techniques; review of the experiences gained and data acquired on the influence of neutron irradiation and oxidizing conditions on key graphite properties; and to exchange information useful for decommissioning activities. The participants provided twenty-seven papers on behalf of their countries and respective technical organizations. An open discussion followed each of the presentations. A consistently reoccurring theme throughout the specialists meeting was the noticeable reduction in the number of graphite experts remaining the nuclear power industry. Graphite moderated power reactors have provided a significant contribution to the generation of electricity throughout the past forty years and will continue to be a prominent energy source for the future. Yet, many of the renowned experts in the field of nuclear graphites are nearing the end of their careers without apparent replacement. This, coupled with changes in the focus on nuclear power by some industrialized countries, has prompted the IAEA to initiate an evaluation on the feasibility and interest by Member States of establishing a central archive facility for the storage of data on irradiated graphites. Refs, figs, tabs

  2. Validation of the Monte Carlo criticality program KENO IV and the Hansen-Roach sixteen-energy-group-cross sections for high-assay uranium systems

    International Nuclear Information System (INIS)

    Handley, G.R.; Masters, L.C.; Stachowiak, R.V.

    1981-01-01

    Validation of the Monte Carlo criticality code, KENO IV, and the Hansen-Roach sixteen-energy-group cross sections was accomplished by calculating the effective neutron multiplication constant, k/sub eff/, of 29 experimentally critical assemblies which had uranium enrichments of 92.6% or higher in the uranium-235 isotope. The experiments were chosen so that a large variety of geometries and of neutron energy spectra were covered. Problems, calculating the k/sub eff/ of systems with high-uranium-concentration uranyl nitrate solution that were minimally reflected or unreflected, resulted in the separate examination of five cases

  3. 77 FR 14837 - Bioassay at Uranium Mills

    Science.gov (United States)

    2012-03-13

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0057] Bioassay at Uranium Mills AGENCY: Nuclear Regulatory..., ``Bioassay at Uranium Mills.'' This guide describes a bioassay program acceptable to the NRC staff for uranium mills and applicable portions of uranium conversion facilities where the possibility of exposure...

  4. Depleted uranium processing and fluorine extraction

    International Nuclear Information System (INIS)

    Laflin, S.T.

    2010-01-01

    Since the beginning of the nuclear era, there has never been a commercial solution for the large quantities of depleted uranium hexafluoride generated from uranium enrichment. In the United States alone, there is already in excess of 1.6 billion pounds (730 million kilograms) of DUF_6 currently stored. INIS is constructing a commercial uranium processing and fluorine extraction facility. The INIS facility will convert depleted uranium hexafluoride and use it as feed material for the patented Fluorine Extraction Process to produce high purity fluoride gases and anhydrous hydrofluoric acid. The project will provide an environmentally friendly and commercially viable solution for DUF_6 tails management. (author)

  5. ORALLOY (93.2 235U) METAL CYLINDER WITH BERYLLIUM TOP REFLECTOR

    International Nuclear Information System (INIS)

    Bess, John D.; Montierth, Leland M.; Reed, Raymond L.; Mihalczo, John T.

    2010-01-01

    A variety of critical experiments were constructed of enriched uranium metal during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, one experiment was comprised of a stack of approximately 7-inch-diameter metal discs. The bottom of the stack consisted of uranium with an approximate height of 4-1/8 inches. The top of the stack consisted of beryllium with an approximate height of 5-9/16 inches. This experiment was performed on August 20, 1963 by J. T. Mihalczo and R. G. Taylor (Ref. 1) with accompanying logbook. Both detailed and simplified model specifications are provided in this evaluation. This fast-spectra experiment was determined to represent an acceptable benchmark. The calculated eigenvalues for both the detailed and simple models are within approximately 0.5% of the benchmark values, but significantly greater than 3s from the benchmark value because the uncertainty in the benchmark is very small: ±0.0002 (1s). There is significant variability between results using different neutron cross section libraries, the greatest being a k eff of ∼0.65% . Unreflected and unmoderated experiments with the same highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in HEU MET FAST 051. Thin graphite reflected (2 inches or less) experiments also using the same highly enriched uranium metal parts are evaluated

  6. Initial tank calibration at NUCEF critical facility. 2

    International Nuclear Information System (INIS)

    Yanagisawa, Hiroshi

    1994-07-01

    Analyses on initial tank calibration data were carried out for the purpose of the nuclear material accountancy and control for critical facilities in NUCEF: Nuclear Fuel Cycle Safety Engineering Research Facility. Calibration functions to evaluate volume of nuclear material solution in accountancy tanks were determined by regression analysis on the data considering dimension and shape of the tank. The analyses on dip-tube separation (probe separation), which are necessary to evaluate solution density in the tanks, were also carried out. As a result, regression errors of volume calculated with the calibration functions were within 0.05 lit. (0.01%) at a nominal level of Pu accountancy tanks. Errors of the evaluated dip-tube separations were also small, e.g. within 0.2mm (0.11%). Therefore, it was estimated that systematic errors of bulk measurements would satisfy the target value of NUCEF critical facilities (0.3% for Pu accountancy tanks). This paper summarizes the data analysis methods, results of analysis and evaluated errors. (author)

  7. Combined use of the RPI [Rensselaer Polytechnic Institute] reactor for training and critical experiments

    International Nuclear Information System (INIS)

    Harris, D.R.; Rohr, R.R.; Rodriguez-Vera, F.

    1990-01-01

    The Rensselaer Polytechnic Institute (RPI) reactor critical facility (RCF) has provided educational and research opportunities for RPI and other students for >25 yr. The RCF was built by the American Locomotive Company (ALCO) in the 1950s as a critical facility in support of the army package power reactor program, and, when ALCO went out of business in 1964, the RCF was acquired by RPI. Since that time, RPI has operated the RCF primarily in a teaching mode in the nuclear engineering department, although reactor research, activation analyses, and reactivity assays have been carried out as well. Until recently, the RCF was fueled by plates containing highly enriched uranium as a cermet in stainless steel. This highly enriched uranium (HEU) fuel was replaced recently by 4.81 wt% enriched UO 2 high-density pellets clad in stainless steel rods. The use of these SPERT (F1) fuel rods in the RCF provided a cost-effective method for conversion of the core from HEU to low-enriched uranium and for enhancement of the RCF training and research program. The RCF is the only facility in the United States that provides reactor training on a core containing fuel that is similar to that used in power industry light water reactors (LWRs). Moreover, the RCF is the only facility in the United States currently available for supplying critical experimental data in support of the LWR power industry. Thus, the RCF is in a unique position to carry out important training and research services consistent with RPI's nuclear engineering objectives

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

    International Nuclear Information System (INIS)

    Alvarez, A.; Correa, E.; Navarro, N.; Sancho, C.; Angeles, A.

    2000-01-01

    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)

  9. Influence of the reduction-crucible material on the uranium properties

    International Nuclear Information System (INIS)

    Braga, F.J.C.; Bose, A.; Freitas, C.T. de

    1979-01-01

    The uranium obtained by UF 4 reduction using Mg in bombs coated with different materials such as alumina, blast furnace slag, Zirconia and graphite was studied. The reduction process involves a reaction that altains temperatures of the order of 1600 0 C at tightly closed enclosure environment. Assuming in this process that the only possible influencial agent on the reaction main product, i.e., metallic uranium is the own bomb coaling, different properties, mechanical-metallurgical and phase-transformation characteristics were examined and the influences of the coating materials were compared. The comparison of these properties was also studied in uranium refined by arc fusion. (Author) [pt

  10. Removal of uranium from uranium-contaminated soils -- Phase 1: Bench-scale testing

    International Nuclear Information System (INIS)

    Francis, C.W.

    1993-09-01

    To address the management of uranium-contaminated soils at Fernald and other DOE sites, the DOE Office of Technology Development formed the Uranium in Soils Integrated Demonstration (USID) program. The USID has five major tasks. These include the development and demonstration of technologies that are able to (1) characterize the uranium in soil, (2) decontaminate or remove uranium from the soil, (3) treat the soil and dispose of any waste, (4) establish performance assessments, and (5) meet necessary state and federal regulations. This report deals with soil decontamination or removal of uranium from contaminated soils. The report was compiled by the USID task group that addresses soil decontamination; includes data from projects under the management of four DOE facilities [Argonne National Laboratory (ANL), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), and the Savannah River Plant (SRP)]; and consists of four separate reports written by staff at these facilities. The fundamental goal of the soil decontamination task group has been the selective extraction/leaching or removal of uranium from soil faster, cheaper, and safer than current conventional technologies. The objective is to selectively remove uranium from soil without seriously degrading the soil's physicochemical characteristics or generating waste forms that are difficult to manage and/or dispose of. Emphasis in research was placed more strongly on chemical extraction techniques than physical extraction techniques

  11. Safety of uranium enrichment plant

    International Nuclear Information System (INIS)

    Yonekawa, Shigeru; Morikami, Yoshio; Morita, Minoru; Takahashi, Tsukasa; Tokuyasu, Takashi.

    1991-01-01

    With respect to safety evaluation of the gas centrifuge enrichment facility, several characteristic problems are described as follows. Criticality safety in the cascade equipments can be obtained to maintain the enrichment of UF 6 below 5 %. External radiation dose equivalent rate of the 30B cylinder is low enough, the shield is not necessary. Penetration ratio of the two-stage HEPA filters for UF 6 aerosol is estimated at 10 -9 . From the experimental investigation, vacuum tightness is not damaged by destruction of gas centrifuge rotor. Carbon steel can be used for uranium enrichment equipments under the condition below 100degC. (author)

  12. The Pajarito Site operating procedures for the Los Alamos Critical Experiments Facility

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1991-12-01

    Operating procedures consistent with DOE Order 5480.6, and the American National Standard Safety Guide for the Performance of Critical Experiments are defined for the Los Alamos Critical Experiments Facility (LACEF) of the Los Alamos National Laboratory. These operating procedures supersede and update those previously published in 1983 and apply to any criticality experiment performed at the facility. 11 refs

  13. Reactor physics experiments in PURNIMA sub critical facility coupled with 14 MeV neutron source

    International Nuclear Information System (INIS)

    Kumar, Rajeev; Degweker, S.B.; Patel, Tarun; Bishnoi, Saroj; Adhikari, P.S.

    2011-01-01

    Accelerator Driven Sub-critical Systems (ADSS) are attracting increasing worldwide attention due to their superior safety characteristics and their potential for burning actinide and fission product waste and energy production. A number of countries around the world have drawn up roadmaps/programs for development of ADSS. Indian interest in ADSS has an additional dimension, which is related to the planned utilization of our large thorium reserves for future nuclear energy generation. A programme for development of ADSS is taken up at the Bhabha Atomic Research Centre (BARC) in India. This includes R and D activities for high current proton accelerator development, target development and Reactor Physics studies. As part of the ADSS Reactor Physics research programme, a sub-critical facility is coming up in BARC which will be coupled with an existing D-D/D-T neutron generator. Two types of cores are planned. In one of these, the sub-critical reactor assembly consists of natural uranium moderated by high density polyethylene (HDP) and reflected by BeO. The other consists of natural uranium moderated by light water. The maximum neutron yield of the neutron source with tritium target is around 10 10 neutron per sec. Various reactor physics experiments like measurement of the source strength, neutron flux distribution, buckling estimation and sub-critical source multiplication are planned. Apart from this, measurement of the total fission power and neutron spectrum will also be carried out. Mainly activation detectors will be used in all in-core neutron flux measurement. Measurement of the degree of sub-criticality by various deterministic and noise methods is planned. Helium detectors with advanced data acquisition card will be used for the neutron noise experiments. Noise characteristics of ADSS are expected to be different from that of traditional reactors due to the non-Poisson statistical features of the source. A new theory incorporating these features has been

  14. Progress in radioactive graphite waste management. Additional information

    International Nuclear Information System (INIS)

    2010-06-01

    Radioactive graphite constitutes a major waste stream which arises during the decommissioning of certain types of nuclear installations. Worldwide, a total of around 250 000 tonnes of radioactive graphite, comprising graphite moderators and reflectors, will require management solutions in the coming years. 14 C is the radionuclide of greatest concern in nuclear graphite; it arises principally through the interaction of reactor neutrons with nitrogen, which is present in graphite as an impurity or in the reactor coolant or cover gas. 3 H is created by the reactions of neutrons with 6 Li impurities in graphite as well as in fission of the fuel. 36 Cl is generated in the neutron activation of chlorine impurities in graphite. Problems in the radioactive waste management of graphite arise mainly because of the large volumes requiring disposal, the long half-lives of the main radionuclides involved and the specific properties of graphite - such as stored Wigner energy, graphite dust explosibility and the potential for radioactive gases to be released. Various options for the management of radioactive graphite have been studied but a generally accepted approach for its conditioning and disposal does not yet exist. Different solutions may be appropriate in different cases. In most of the countries with radioactive graphite to manage, little progress has been made to date in respect of the disposal of this material. Only in France has there been specific thinking about a dedicated graphite waste-disposal facility (within ANDRA): other major producers of graphite waste (UK and the countries of the former Soviet Union) are either thinking in terms of repository disposal or have no developed plans. A conference entitled 'Solutions for Graphite Waste: a Contribution to the Accelerated Decommissioning of Graphite Moderated Nuclear Reactors' was held at the University of Manchester 21-23 March 2007 in order to stimulate progress in radioactive graphite waste management

  15. Scrap uranium recycling via electron beam melting

    International Nuclear Information System (INIS)

    McKoon, R.

    1993-11-01

    A program is underway at the Lawrence Livermore National Laboratory (LLNL) to recycle scrap uranium metal. Currently, much of the material from forging and machining processes is considered radioactive waste and is disposed of by oxidation and encapsulation at significant cost. In the recycling process, uranium and uranium alloys in various forms will be processed by electron beam melting and continuously cast into ingots meeting applicable specifications for virgin material. Existing vacuum processing facilities at LLNL are in compliance with all current federal and state environmental, safety and health regulations for the electron beam melting and vaporization of uranium metal. One of these facilities has been retrofitted with an auxiliary electron beam gun system, water-cooled hearth, crucible and ingot puller to create an electron beam melt furnace. In this furnace, basic process R ampersand D on uranium recycling will be performed with the goal of eventual transfer of this technology to a production facility

  16. Uranium, the joint facilities, disarmament and peace

    International Nuclear Information System (INIS)

    Anon.

    1984-07-01

    The Australian Government recognises that the only outcome which is ultimately acceptable is total nuclear disarmament. One of the issues addressed in this document is that of the export of Australian uranium. It is argued that by refusing to supply uranium Australia would be making a symbolic gesture which would have the reverse effect to that which was intended. It would weaken the non-proliferation regime and Australia's voice in arms control and disarmament forums

  17. The experimental determination of the buckling in the bare heavy water natural uranium critical assembly 'RB'

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N M; Popovic, D D; Takac, S M; Djordjevic, M M [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1960-03-15

    The buckling in the bare heavy water natural uranium critical assembly was determined by measuring the thermal neutron flux distribution. The obtained value for the critical buckling at the temperature of 20 deg C is: B{sup 2} = (8.516 {+-} 0.02) m{sup -2}. The above error is a statistical one, obtained from several series of measurements. The possible systematic error was estimated as 0.1 m{sup -2}. (author)

  18. The experimental determination of the buckling in the bare heavy water natural uranium critical assembly 'RB'

    International Nuclear Information System (INIS)

    Raisic, N.M.; Popovic, D.D.; Takac, S.M.; Djordjevic, M.M.

    1960-01-01

    The buckling in the bare heavy water natural uranium critical assembly was determined by measuring the thermal neutron flux distribution. The obtained value for the critical buckling at the temperature of 20 deg C is: B 2 = (8.516 ± 0.02) m -2 . The above error is a statistical one, obtained from several series of measurements. The possible systematic error was estimated as 0.1 m -2 . (author)

  19. Remediation of former uranium mining and milling facilities in Germany - the WISMUT experience

    International Nuclear Information System (INIS)

    Gatzweiler, R.

    2000-01-01

    The former German Democratic Republic (East Germany) provided most of the natural uranium for the nuclear programmes of the former Soviet Union. Uranium mining and milling activities caused extensive devastation and resulted in large amounts of waste with serious impacts on the environment and unacceptable risks to human health. Production ceased in 1990-91 in the course of the reunification of Germany. At the same time a very large environmental remediation programme was initiated by the German Federal Government. WISMUT GmbH, the successor company of the former Soviet-German enterprise SDAG WISMUT, was designated to carry out this DM 13 billion programme. The programme is currently in its ninth year and will likely continue up to 2015. The initial assessment of the remediation and the cost estimates were based on closure plans for the mining and milling facilities, an extensive environmental database and basic concepts for site specific remediation in accordance with legal requirements and directives. The decision making process for individual remediation objects is based on risk analyses and the evaluation of remediation options. The methodologies used depend on the size and complexity of the individual object. For simple cases, an environmental assessment study is used. Remediation options for larger and more complex objects such as tailings facilities are evaluated by multi-attribute analysis with emphasis on sensitivity investigations. The general public is not formally involved in the decision making process but is informed on conceptual remediation plans for the individual sites. These plans are regularly updated. For several of the remediated facilities, future use cannot be unrestricted and therefore, some form of institutional control is needed. To date, criteria for close-out have not been clearly defined for all sites. Similarly, criteria concerning transfer of ownership and responsibilities for long term surveillance and maintenance have yet to be

  20. Activation of Persulfates by Graphitized Nanodiamonds for Removal of Organic Compounds.

    Science.gov (United States)

    Lee, Hongshin; Kim, Hyoung-Il; Weon, Seunghyun; Choi, Wonyong; Hwang, Yu Sik; Seo, Jiwon; Lee, Changha; Kim, Jae-Hong

    2016-09-20

    This study introduces graphited nanodiamond (G-ND) as an environmentally friendly, easy-to-regenerate, and cost-effective alternative catalyst to activate persulfate (i.e., peroxymonosulfate (PMS) and peroxydisulfate (PDS)) and oxidize organic compounds in water. The G-ND was found to be superior for persulfate activation to other benchmark carbon materials such as graphite, graphene, fullerene, and carbon nanotubes. The G-ND/persulfate showed selective reactivity toward phenolic compounds and some pharmaceuticals, and the degradation kinetics were not inhibited by the presence of oxidant scavengers and natural organic matter. These results indicate that radical intermediates such as sulfate radical anion and hydroxyl radical are not majorly responsible for this persulfate-driven oxidation of organic compounds. The findings from linear sweep voltammetry, thermogravimetric analysis, Fourier transform infrared spectroscopy, and electron paramagnetic resonance spectroscopy analyses suggest that the both persulfate and phenol effectively bind to G-ND surface and are likely to form charge transfer complex, in which G-ND plays a critical role in mediating facile electron transfer from phenol to persulfate.

  1. Criticality safety analysis for mockup facility

    International Nuclear Information System (INIS)

    Shin, Young Joon; Shin, Hee Sung; Kim, Ik Soo; Oh, Seung Chul; Ro, Seung Gy; Bae, Kang Mok

    2000-03-01

    Benchmark calculations for SCALE4.4 CSAS6 module have been performed for 31 UO 2 fuel, 15MOX fuel and 10 metal material criticality experiments and then calculation biases of the SCALE 4.4 CSAS6 module have been revealed to be 0.00982, 0.00579 and 0.02347, respectively. When CSAS6 is applied to the criticality safety analysis for the mockup facility in which several kinds of nuclear material components are included, the calculation bias of CSAS6 is conservatively taken to be 0.02347. With the aid of this benchmarked code system, criticality safety analyses for the mockup facility at normal and hypothetical accidental conditions have been carried out. It appears that the maximum K eff is 0.28356 well below than the critical limit, K eff =0.95 at normal condition. In a hypothetical accidental condition, the maximum K eff is found to be 0.73527 much lower than the subcritical limit. For another hypothetical accidental condition the nuclear material leaks out of container and spread or lump in the floor, it was assumed that the nuclear material is shaped into a slab and water exists in the empty space of the nuclear material. K eff has been calculated as function of slab thickness and the volume ratio of water to nuclear material. The result shows that the K eff increases as the water volume ratio increases. It is also revealed that the K eff reaches to the maximum value when water if filled in the empty space of nuclear material. The maximum K eff value is 0.93960 lower than the subcritical limit

  2. Hot pressing of U-UC cermets and stoichiometric uranium monocarbide; Preparation par frittage sous charge de cermets U-UC et de monocarbure stoechiometrique

    Energy Technology Data Exchange (ETDEWEB)

    Dubuisson, J; Houyvet, A; Le Boulbin, E; Lucas, R; Moranville, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    High density fuels, either in uranium monocarbide or in U-UC cermets have been prepared on laboratory-scale, by hot pressing of mixtures of uranium powder and graphite in suitable proportions. Uranium powder is prepared by calcium reduction of UO{sub 2} followed by an acetic leaching at low temperature. An adequate protection-treatment permits the manipulation of the powder in the open air. Uranium and Graphite powders are intimately mixed and then hot pressed in a double effect graphite die at a temperature of 900-1000 deg. C under a charge of 200 kg/cm{sup 2} during 3 hours. A special design of the die avoids the breaking of the graphite during the sintering. In this way, samples are prepared, the characteristics of which are: 1) {+-} 5 pour cent of homogeneity for a ratio height/diameter = 2. 2) almost theoretical density (98 pour cent) 3) low concentration of unreacted carbon (heat treatment of stoichiometric monocarbide can be useful for completion of reaction) 4) the micrographic examination shows: - a network of monocarbide surrounding uranium in the case of low concentration cermets (<2,5 per cent C) - two networks intimately mixed for high concentration cermets (<2,5 per cent C) - a fine grain structure for the monocarbide (10 u). 5) In every case, the X rays examinations show a fine grain structure without any orientation, and no UC{sub 2}. Some indications are given on the physical (thermal cycling, conductibility) and chemical properties (corrosion, reaction with cladding materials). (author)Fren. [French] Une methode de preparation de combustibles de haute densite, soit en monocarbure d'uranium, soit en cermets U-UC, a ete mise au point au laboratoire. Il s'agit du frittage sous charge de melanges de poudres d'uranium et de graphite en proportion convenable. La poudre d'uranium est elaboree par calciothermie de l'oxyde UO{sub 2} suivie d'un lavage acetique a basse temperature. Un traitement de protection adequat pe sa manipulation a l'air. Les poudres

  3. Analysis of fuel cycles with natural uranium; Analiza gorivnih ciklusa sa prirodnim uranom

    Energy Technology Data Exchange (ETDEWEB)

    Stojanovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-05-15

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

  4. Initial conceptual design study of self-critical nuclear pumped laser systems

    Science.gov (United States)

    Rodgers, R. J.

    1979-01-01

    An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.

  5. Critical review of uranium resources and production capability to 2020

    International Nuclear Information System (INIS)

    1998-08-01

    This report was prepared to assess the changing uranium supply and demand situation as well as the adequacy of uranium resources and the production capability to supply uranium concentrate to meet reactor demand through 2020. Uranium production has been meeting only 50 to 60 percent of the world requirements with the balance met from sale of excess inventory offered on the market at low prices. It is generally agreed by most specialists that the end of the excess inventory is approaching. With inventory no longer able to meet the production shortfall it is necessary to significantly expand uranium production to fill an increasing share of demand. Non-production supplies of uranium, such as the blending of highly enriched uranium (HEU) warheads to produce low enriched reactor fuel and reprocessing of spent fuel, are also expected to grow in importance as a fuel source. This analysis addresses three major concerns as follows: adequacy of resources to meet projected demand; adequacy of production capability to produce the uranium; and market prices to sustain production to fill demand. This analysis indicates uranium mine production to be the primary supply providing about 76 to 78 percent of cumulative needs through 2020. Alternative sources supplying the balance, in order of relative importance are: (1) low enriched uranium (LEU) blended from 500 tonnes of highly enriched uranium (HEU) Russian weapons, plus initial US Department of Energy (US DOE) stockpile sales (11 to 13%); (2) reprocessing of spent nuclear fuel (6%) and; (3) utility and Russian stockpiles. Further this report gives uranium production profiles by countries: CIS producers (Kazakhstan, Russian Federation, Ukraine, Uzbekistan) and other producers (Australia, Canada, China, Gabon, Mongolia, Namibia, Niger, South Africa, United States of America)

  6. Licensed fuel facility status report

    International Nuclear Information System (INIS)

    1990-04-01

    NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related NRC investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  7. Licensed fuel facility status report

    International Nuclear Information System (INIS)

    Joy, D.; Brown, C.

    1993-04-01

    NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related NRC investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  8. Standard test methods for arsenic in uranium hexafluoride

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 These test methods are applicable to the determination of total arsenic in uranium hexafluoride (UF6) by atomic absorption spectrometry. Two test methods are given: Test Method A—Arsine Generation-Atomic Absorption (Sections 5-10), and Test Method B—Graphite Furnace Atomic Absorption (Appendix X1). 1.2 The test methods are equivalent. The limit of detection for each test method is 0.1 μg As/g U when using a sample containing 0.5 to 1.0 g U. Test Method B does not have the complete collection details for precision and bias data thus the method appears as an appendix. 1.3 Test Method A covers the measurement of arsenic in uranyl fluoride (UO2F2) solutions by converting arsenic to arsine and measuring the arsine vapor by flame atomic absorption spectrometry. 1.4 Test Method B utilizes a solvent extraction to remove the uranium from the UO2F2 solution prior to measurement of the arsenic by graphite furnace atomic absorption spectrometry. 1.5 Both insoluble and soluble arsenic are measured when UF6 is...

  9. Detection of uranium enrichment activities using environmental monitoring techniques

    International Nuclear Information System (INIS)

    Belew, W.L.; Carter, J.A.; Smith, D.H.; Walker, R.L.

    1993-01-01

    Uranium enrichment processes have the capability of producing weapons-grade material in the form of highly enriched uranium. Thus, detection of undeclared uranium enrichment activities is an international safeguards concern. The uranium separation technologies currently in use employ UF 6 gas as a separation medium, and trace quantities of enriched uranium are inevitably released to the environment from these facilities. The isotopic content of uranium in the vegetation, soil, and water near the plant site will be altered by these releases and can provide a signature for detecting the presence of enriched uranium activities. This paper discusses environmental sampling and analytical procedures that have been used for the detection of uranium enrichment facilities and possible safeguards applications of these techniques

  10. The epidemiological studies of leukemia around nuclear facilities for children and young adults: critical review

    International Nuclear Information System (INIS)

    2008-01-01

    This objective of this report is to make a review of studies relative to the risk of leukemia among children and young adults less than twenty five years old near the nuclear facilities. the nuclear facilities considered in this report are nuclear power plants (electric power generation), the nuclear research centers, the fuel or weapons fabrication plants, reprocessing plants. This report does not describe the risk analysis near the sites of nuclear weapons test, the consequences of accidents on nuclear facilities such Chernobylsk or the the population near the military and industrial site of Mayak (Ural). The same is for the mining sites and the facilities of uranium extraction that are out of this report. (N.C.)

  11. Seismic evaluation of critical facilities at the Lawrence Livermore Laboratory

    International Nuclear Information System (INIS)

    Murray, R.C.; Tokarz, F.J.

    1976-01-01

    The performance of critical facilities at the Lawrence Livermore Laboratory (LLL) are being evaluated for severe earthquake loading. Facilities at Livermore, Site-300 and the Nevada Test Site are included in this study. These facilities are identified, the seismic criteria used for the analysis are indicated, the various methods used for structural analysis are discussed and a summary of the results of facilities analyzed to date are presented

  12. Spectrographic determination of impurities in uranium tetrafluoride

    International Nuclear Information System (INIS)

    Capdevila Perez, C.; Roca Adell, M.; Alvarez Gonzalez, F.

    1967-01-01

    A carrier distillation method for the determination of Ag, Al, As, B, Cd, Cr, Cu, Fe, Mg, Mn, Ni, Pb, and Si in uranium tetrafluoride was develop ped. the previous addition of 25% Y 2 3 prevents the excitation of uranium by conversion of the volatile UF 4 into U 3 0 8 during the arc discharge. NaCl or Ga 2 0 3 , containing Ge and V as internal standards, are used as carriers, and samples are arced in 10 Amp. d.c. arc in a graphite anode cup. 7 mm diameter, 10 mm deep, being the weight of charge 300 mg. (Author) 14 refs

  13. In situ observations of graphite formation during solidification of cast iron

    DEFF Research Database (Denmark)

    Bjerre, Mathias Karsten

    solidification and growth continues throughout solid state cooling and the eutectoid transformation. Years of research have greatly improved the understanding of the basic mechanisms that control graphite growth as well as the ability to control graphite morphology during industrial production of cast components......, the solidification of cast iron is studied with focus on formation and growth of spheroidal graphite. To this end, an experiment is conducted at the Diamond Light Source synchrotron facility in Harwell, UK: Employing an environmental cell devel-oped at the Manchester X-ray Imaging Facility at the University...... state growth presented in the present thesis. From the analysis it is clear that the presented data is of an unprecedented quality and that it represents a solid basis for validation of future models. Solidification simulations of a ductile cast iron component highlights the importance of the nucleation...

  14. Critical mass experiment using U-235 foils and lucite plates

    International Nuclear Information System (INIS)

    Sanchez, R.; Butterfield, K.; Kimpland, R.; Jaegers, P.

    1998-01-01

    The main objective of this experiment was to show how the multiplication of the system increases as moderated material is placed between highly enriched uranium foils. In addition, this experiment served to demonstrate the hand-stacking techniques, and approach to criticality by remote operation. This experiment was designed by Tom McLaughlin in the mid seventies as part of the criticality safety course that is taught at Los Alamos Critical Experiment Facility (LACEF). The W-U-235 ratio for this experiment was 215 which is where the minimum critical mass for this configuration occurs

  15. Critical mass experiment using 235U foils and lucite plates

    International Nuclear Information System (INIS)

    Sanchez, R.; Butterfield, K.; Kimpland, R.; Jaegers, P.

    1998-01-01

    This experiment demonstrated how the neutron multiplication of a system increases as moderated material is placed between highly enriched uranium foils. In addition, this experiment served to demonstrate the hand-stacking technique and approach to criticality be remote operation. This experiment was designed by McLaughlin in the mid-seventies as part of the criticality safety course that is taught at the Los Alamos Critical Experiments Facility. The H/ 235 U ratio for this experiment was 215, which is the ratio at which the minimum critical mass for this configuration occurs

  16. Uranium extraction history using pressure leaching

    International Nuclear Information System (INIS)

    Fraser, K.S.; Thomas, K.G.

    2010-01-01

    Over the past 60 years of uranium process development only a few commercial uranium plants have adopted a pressure leaching process in their flowsheet. The selection of acid versus alkaline pressure leaching is related to the uranium and gangue mineralogy. Tetravalent (U"+"4) uranium has to be oxidized to hexavalent (U"+"6) uranium to be soluble. Refractory tetravalent uranium requires higher temperature and pressure, as practised in pressure leaching, for conversation to soluble hexavalent uranium. This paper chronicles the history of these uranium pressure leaching facilities over the past 60 years, with specific details of each design and operation. (author)

  17. Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R M

    1976-01-01

    Evidence of expanding markets, improved prices and the short supply of uranium became abundantly clear in 1975, providing the much needed impetus for widespread activity in all phases of uranium operations. Exploration activity that had been at low levels in recent years in Canada was evident in most provinces as well as the Northwest Territories. All producers were in the process of expanding their uranium-producing facilities. Canada's Atomic Energy Control Board (AECB) by year-end had authorized the export of over 73,000 tons of U/sub 3/0/sub 8/ all since September 1974, when the federal government announced its new uranium export guidelines. World production, which had been in the order of 25,000 tons of U/sub 3/0/sub 8/ annually, was expected to reach about 28,000 tons in 1975, principally from increased output in the United States.

  18. Sensibility test for uranium ores from Qianjiadian sandstone type uranium deposit

    International Nuclear Information System (INIS)

    Zhang Mingyu

    2005-01-01

    Sensibility tests for uranium ores from Qianjiadian sandstone type uranium deposit in Songliao Basin which is suitable to in-situ leach are carried out, including water sensibility, velocity sensibility, salt sensibility, acid sensibility and alkaline sensibility. The sensibility critical value of this ore is determined. Some references on mining process and technical parameter are provided for in-situ leaching of uranium. (authors)

  19. Design of the 50 kW neutron converter for SPIRAL2 facility

    Energy Technology Data Exchange (ETDEWEB)

    Avilov, M.S. [Budker Institute of Nuclear Physics, 630090 Novosibirsk, SB RAS (Russian Federation); Tecchio, L.B., E-mail: tecchio@lnl.infn.i [Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Titov, A.T. [Boreskov Institute of Catalysis, 630090 Novosibirsk, SB RAS (Russian Federation); Tsybulya, V.S. [Trofimuk Institute of Geology, 630090 Novosibirsk, SB RAS (Russian Federation); Zhmurikov, E.I. [Budker Institute of Nuclear Physics, 630090 Novosibirsk, SB RAS (Russian Federation)

    2010-06-21

    SPIRAL2 is a facility for the study of fundamental nuclear physics and multidisciplinary research. SPIRAL2 represents a major advance for research on exotic nuclei. The radioactive ion beam (RIB) production system is comprised of a neutron converter, a target and an ion source. This paper is dedicated to the designing of the 50 kW neutron converter for the SPIRAL2 facility. Among the different variants of the neutron converter, the one based on a rotating solid disk seems quite attractive due to its safety, ease in production and relatively low cost. Dense graphite used as the converter's material allows the production of high-intensity neutron flux and, at the same time, the heat removal from the converter by means of radiation cooling. Thermo-mechanical simulations performed in order to determine the basic geometry and physical characteristics of the neutron production target for SPIRAL2 facility, to define the appropriate beam power distribution, and to predict the target behaviour under the deuteron beam of nominal parameters (40 MeV, 1.2 mA, 50 kW) are presented. To study the main physical and mechanical properties and serviceability under operating conditions, several kinds of graphite have been analyzed and tested. The paper reports the results of such measurements. Radiation damage is the most important issue for the application of graphite as neutron converter. It is well known that the thermal conductivity of the neutron-irradiated graphite is reduced by a factor of 10 from the initial value after irradiation. Difference in volume expansions between the matrix and the fiber results in serious damage of neutron-irradiated C/C composites. Calculations showed that at high temperature the effect of neutron radiation is not so critical and that the change in thermal conductivity does not prevent the use of graphite as neutron converter.

  20. Design of the 50 kW neutron converter for SPIRAL2 facility

    International Nuclear Information System (INIS)

    Avilov, M.S.; Tecchio, L.B.; Titov, A.T.; Tsybulya, V.S.; Zhmurikov, E.I.

    2010-01-01

    SPIRAL2 is a facility for the study of fundamental nuclear physics and multidisciplinary research. SPIRAL2 represents a major advance for research on exotic nuclei. The radioactive ion beam (RIB) production system is comprised of a neutron converter, a target and an ion source. This paper is dedicated to the designing of the 50 kW neutron converter for the SPIRAL2 facility. Among the different variants of the neutron converter, the one based on a rotating solid disk seems quite attractive due to its safety, ease in production and relatively low cost. Dense graphite used as the converter's material allows the production of high-intensity neutron flux and, at the same time, the heat removal from the converter by means of radiation cooling. Thermo-mechanical simulations performed in order to determine the basic geometry and physical characteristics of the neutron production target for SPIRAL2 facility, to define the appropriate beam power distribution, and to predict the target behaviour under the deuteron beam of nominal parameters (40 MeV, 1.2 mA, 50 kW) are presented. To study the main physical and mechanical properties and serviceability under operating conditions, several kinds of graphite have been analyzed and tested. The paper reports the results of such measurements. Radiation damage is the most important issue for the application of graphite as neutron converter. It is well known that the thermal conductivity of the neutron-irradiated graphite is reduced by a factor of 10 from the initial value after irradiation. Difference in volume expansions between the matrix and the fiber results in serious damage of neutron-irradiated C/C composites. Calculations showed that at high temperature the effect of neutron radiation is not so critical and that the change in thermal conductivity does not prevent the use of graphite as neutron converter.

  1. Validation of KENO V.a for highly enriched uranium systems with hydrogen and/or carbon moderation

    International Nuclear Information System (INIS)

    Elliott, E.P.; Vornehm, R.G.; Dodds, H.L. Jr.

    1993-01-01

    This paper describes the validation in accordance with ANSI/ANS-8.1-1983(R1988) of KENO V.a using the 27-group ENDF/B-IV cross-section library for systems containing highly-enriched uranium, carbon, and hydrogen and for systems containing highly-enriched uranium and carbon with high carbon to uranium (C/U) atomic ratios. The validation has been performed for two separate computational platforms: an IBM 3090 mainframe and an HP 9000 Model 730 workstation, both using the Oak Ridge Y-12 Plant Nuclear Criticality Safety Software (NCSS) code package. Critical experiments performed at the Oak Ridge Critical Experiments Facility, in support of the Rover reactor program, and at the Pajarito site at Los Alamos National Laboratory were identified as having the constituents desired for this validation as well as sufficient experimental detail to allow accurate construction of KENO V.a calculational models. Calculated values of k eff for the Rover experiments, which contain uranium, carbon, and hydrogen, are between 1.0012 ± 0.0026 and 1.0245 ± 0.0023. Calculation of the Los Alamos experiments, which contain uranium and carbon at high C/U ratios, yields values of k eff between 0.9746 ± 0.0028 and 0.9983 ± 0.0027. Safety criteria can be established using this data for both types of systems

  2. Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

    Energy Technology Data Exchange (ETDEWEB)

    Boussard, F.; Huillery, R. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Combustibles; Averseng, J.L.; Serpantie, J.P. [Novatome Industries, 92 - Le Plessis-Robinson (France)

    1994-12-31

    This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs.

  3. Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

    International Nuclear Information System (INIS)

    Boussard, F.; Huillery, R.

    1994-01-01

    This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs

  4. Criticality safety evaluation of a type B enriched uranium shipping container

    International Nuclear Information System (INIS)

    Hopper, C.M.

    1978-01-01

    The Oak Ridge Y-12 Plant Model DT-14 container was developed to replace and extend the enriched uranium shipping capabilities of the USA/5765/BF Vermiculite shipping container. This work was accomplished to comply with the DOE Immediate Action Directive Number 0529-02 for ''Phasing out the use of loose or bagged Vermiculite packaging material as a thermal shield and shock absorber in radioactive material packages''. The Model DT-14 is fabricated from a Specification 17H 30-gallon drum, cane fiberboard insulation, and a steel inner containment vessel (159 mm dia by 390 mm height). The single-package and array analyses are based upon results of the multigroup Monte Carlo criticality program, KENO, utilizing 16-energy-group Hansen-Roach, and Knight modified 238 U cross sections. The program and cross sections are considered well established on the basis of their success in calculating a large variety of critical experiments. Validation results show that a calculated neutron multiplication factor plus two standard deviations equal to 0.970 or greater must be considered critical, and all lower values may be considered safe

  5. Low-resolution gamma-ray measurements of uranium enrichment

    International Nuclear Information System (INIS)

    Sprinkle, J.K. Jr.; Christiansen, A.; Cole, R.; Collins, M.L.

    1996-01-01

    Facilities that process special nuclear material perform periodic inventories. In bulk facilities that process low-enriched uranium, these inventories and their audits are based primarily on weight and enrichment measurements. Enrichment measurements determine the 211 U weight fraction of the uranium compound from the passive gamma-ray emissions of the sample. Both international inspectors and facility operators rely on the capability to make in-field gamma-ray measurements of uranium enrichment. These users require rapid, portable measurement capability. Some in-field measurements have been biased, forcing the inspectors to resort to high-resolution measurements or mass spectrometry to accomplish their goals

  6. Production of nanodiamonds by high-energy ion irradiation of graphite at room temperature

    International Nuclear Information System (INIS)

    Daulton, T.L.; Kirk, M.A.; Lewis, R.S.; Rehn, L.E.

    2001-01-01

    It has previously been shown that graphite can be transformed into diamond by MeV electron and ion irradiation at temperatures above approximately 600 deg. C. However, there exists geological evidence suggesting that carbonaceous materials can be transformed to diamond by irradiation at substantially lower temperatures. For example, submicron-size diamond aggregates have been found in uranium-rich, Precambrian carbonaceous deposits that never experienced high temperature or pressure. To test if diamonds can be formed at lower irradiation temperatures, sheets of fine-grain polycrystalline graphite were bombarded at 20 deg. C with 350±50 MeV Kr ions to fluences of 6x10 12 cm -2 using the Argonne tandem linear accelerator system (ATLAS). Ion-irradiated (and unirradiated control) graphite specimens were then subjected to acid dissolution treatments to remove untransformed graphite and isolate diamonds that were produced; these acid residues were subsequently characterized by high-resolution and analytical electron microscopy. The acid residue of the ion-irradiated graphite was found to contain nanodiamonds, demonstrating that ion irradiation of graphite at ambient temperature can produce diamond. The diamond yield under our irradiation conditions is low, ∼0.01 diamonds/ion. An important observation that emerges from comparing the present result with previous observations of diamond formation during irradiation is that nanodiamonds form under a surprisingly wide range of irradiation conditions. This propensity may be related to the very small difference in the graphite and diamond free-energies coupled with surface-energy considerations that may alter the relative stability of diamond and graphite at nanometer sizes

  7. Conversion and Blending Facility highly enriched uranium to low enriched uranium as uranyl nitrate hexahydrate. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-05

    This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials to pure HEU uranyl nitrate (UNH) and (2) blend pure HEU UNH with depleted and natural UNH to produce HEU UNH crystals. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU Will be produced as a waste suitable for storage or disposal.

  8. Uranium nitride: a cubic antiferromagnet with anisotropic critical behavior

    International Nuclear Information System (INIS)

    Buyers, W.J.L.; Holden, T.M.; Svensson, E.C.; Lander, G.H.

    1977-11-01

    Highly anisotropic critical scattering associated with the transition at T/sub N/ = 49.5 K to the type-I antiferromagnetic structure has been observed in uranium nitride. The transverse susceptibility is found to be unobservably small. The longitudinal susceptibility diverges at T/sub N/ and its anisotropy shows that the spins within the (001) ferromagnetic sheets of the [001] domain are much more highly correlated than they are with the spins lying in adjacent (001) sheets. The correlation range within the sheets is much greater than that expected for a Heisenberg system with the same T/sub N/. The rod-like scattering extended along the spin and domain direction is reminiscent of two-dimensional behavior. The results are inconsistent with a simple localized model and may reflect the itinerant nature of the 5f electrons

  9. Research on the reactor physics using the Kyoto University Critical Assembly (KUCA)

    International Nuclear Information System (INIS)

    1986-10-01

    The Kyoto University Critical Assembly [KUCA] is a multi-core type critical assembly established in 1974, as a facility for the joint use study by researchers of all universities in Japan. Thereafter, many reactor physics experiments have been carried out using three cores (A-, B-, and C-cores) in the KUCA. In the A- and B-cores, solid moderator such as polyethylene or graphite is used, whereas light-water is utilized as moderator in the C-core. The A-core has been employed mainly in connection with the Cockcroft-Walton type accelerator installed in the KUCA, to measure (1) the subcriticality by the pulsed neutron technique for the critical safety research and (2) the neutron spectrum by the time-of-flight technique. Recently, a basic study on the tight lattice core has also launched using the A-core. The B-core has been employed for the research on the thorium fuel cycle ever since. The C-core has been employed (1) for the basic studies on the nuclear characteristics of light-water moderated high-flux research reactors, including coupled-cores, and (2) for a research related to reducing enrichment of uranium fuel used in research reactors. The C-core is being utilized in the reactor laboratory course experiment for students of ten universities in Japan. The data base of the KUCA critical experiments is generated so far on the basis of approximately 350 experimental reports accumulated in the KUCA. Besides, the assessed KUCA code system has been established through analyses on the various KUCA experiments. In addition to the KUCA itself, both of them are provided for the joint use study by researchers of all universities in Japan. (author)

  10. Recent developments in graphite. [Use in HTGR and aerospace

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, J.E.

    1983-01-01

    Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications.

  11. Facile fabrication of ordered mesoporous graphitic carbon nitride for RhB photocatalytic degradation

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Lei; Zhang, Anfeng [State Key Laboratory of Fine Chemicals, PSU-DUT Joint Center for Energy Research, School of Chemical Engineering, Dalian University of Technology, Dalian 116024 (China); Janik, Michael J. [EMS Energy Institute, PSU-DUT Joint Center for Energy Research and Department of Energy & Mineral Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Li, Keyan [State Key Laboratory of Fine Chemicals, PSU-DUT Joint Center for Energy Research, School of Chemical Engineering, Dalian University of Technology, Dalian 116024 (China); Song, Chunshan [State Key Laboratory of Fine Chemicals, PSU-DUT Joint Center for Energy Research, School of Chemical Engineering, Dalian University of Technology, Dalian 116024 (China); EMS Energy Institute, PSU-DUT Joint Center for Energy Research and Department of Energy & Mineral Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Guo, Xinwen, E-mail: guoxw@dlut.edu.cn [State Key Laboratory of Fine Chemicals, PSU-DUT Joint Center for Energy Research, School of Chemical Engineering, Dalian University of Technology, Dalian 116024 (China)

    2017-02-28

    Highlights: • Ordered mesoporous graphitic carbon nitrides with S{sub BET} = 279.3 m{sup 2}/g were prepared. • Enhanced photocatalytic activity and reusability were presented. • Improved S{sub BET} and charge carrier separation efficiency contribute to the activity. - Abstract: Ordered mesoporous graphitic carbon nitrides were prepared by directly condensing the uniform mixtures of melamine and KIT-6. After removal of the KIT-6 sacrificial template, the carbon nitrides were characterized with TEM, N{sub 2} physical adsorption, XRD, FT-IR, XPS, UV–vis and PL spectrometries, and tested for their RhB photocatalytic degradation activity. Together, these characterizations confirmed the as-prepared tunable mesoporous materials with enhanced charge separation efficiency and superior photocatalytic performance. Compared with a conventional bulk g-C{sub 3}N{sub 4}, ordered mesoporous g-C{sub 3}N{sub 4} exhibits a larger specific surface area of 279.3 m{sup 2}/g and a pore size distribution about 4.0 nm and 13.0 nm. Meanwhile, the reduced bandgap energy of 2.77 eV and lower photogenerated electron-hole pair recombination frequency were evidenced by UV–Vis and PL spectra. The RhB photocatalytic degradation activity maximizes with a mass ratio of KIT-6/melamine of 80% (KCN80), and the kinetic constant reaches 0.0760 min{sup −1} which is 16 times higher than that of the bulk sample. Reusability of KCN80 was demonstrated by a lack of evident deactivation after three consecutive reaction periods. The direct condensation of the KIT-6 and melamine mixture does not require pre-casting of the precursor into the pore system of the templates. Owing to its high product yield, improved S{sub BET}, reduced bandgap energy and limited charge recombination, the facile-prepared ordered mesoporous g-C{sub 3}N{sub 4} is a practical candidate for further modification.

  12. Disposal criticality analysis for aluminum-based DOE fuels

    International Nuclear Information System (INIS)

    Davis, J.W.; Gottlieb, P.

    1997-11-01

    This paper describes the disposal criticality analysis for canisters containing aluminum-based Department of Energy fuels from research reactors. Different canisters were designed for disposal of highly enriched uranium (HEU) and medium enriched uranium (MEU) fuel. In addition to the standard criticality concerns in storage and transportation, such as flooding, the disposal criticality analysis must consider the degradation of the fuel and components within the waste package. Massachusetts Institute of Technology (MIT) U-Al fuel with 93.5% enriched uranium and Oak Ridge Research Reactor (ORR) U-Si-Al fuel with 21% enriched uranium are representative of the HEU and MEU fuel inventories, respectively. Conceptual canister designs with 64 MIT assemblies (16/layer, 4 layers) or 40 ORR assemblies (10/layer, 4 layers) were developed for these fuel types. Borated stainless steel plates were incorporated into a stainless steel internal basket structure within a 439 mm OD, 15 mm thick XM-19 canister shell. The Codisposal waste package contains 5 HLW canisters (represented by 5 Defense Waste Processing Facility canisters from the Savannah River Site) with the fuel canister placed in the center. It is concluded that without the presence of a fairly insoluble neutron absorber, the long-term action of infiltrating water can lead to a small, but significant, probability of criticality for both the HEU and MEU fuels. The use of 1.5kg of Gd distributed throughout the MIT fuel and the use of carbon steels for the structural basket or 1.1 kg of Gd distributed in the ORR fuel will reduce the probability of criticality to virtually zero for both fuels

  13. Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

    International Nuclear Information System (INIS)

    Anfimov, S.S.

    2000-01-01

    As a result of decommissioning of water-cooled graphite-moderated reactors, a large amount of rad-waste in the form of graphite stack fragments is generated (on average 1500-2000 tons per reactor). That is why it is essentially important, although complex from the technical point of view, to develop advanced technologies based on up-to-date remotely-controlled systems for unmanned dismantling of the graphite stack containing highly-active long-lived radionuclides and for conditioning of irradiated graphite (IG) for the purposes of transportation and subsequent long term and ecologically safe storage either on NPP sites or in special-purpose geological repositories. The main characteristics critical for radiation and nuclear hazards of the graphite stack are as follows: the graphite stack is contaminated with nuclear fuel that has gotten there as a result of the accidents; the graphite mass is 992 tons, total activity -6?104 Ci (at the time of unit shutdown); the fuel mass in the reactor stack amounts to 100-140 kg, as estimated by IPPE and RDIPE, respectively; γ-radiation dose rate in the stack cells varies from 4 to 4300 R/h, with the prevailing values being in the range from 50 to 100 R/h. In this paper the traditional methods of rad-waste handling as bituminization technology, cementing technology are discussed. In terms of IG handling technology two lines were identified: long-term storage of conditioned IG and IG disposal by means of incineration. The specific cost of graphite immobilization in a radiation-resistant polymeric matrix amounts to -2600 USD per 1 t of graphite, whereas the specific cost of immobilization in slag-stone containers with an inorganic binder (cement) is -1400 USD per 1 t of graphite. On the other hand, volume of conditioned IG rad-waste subject for disposal, if obtained by means of the first technology, is 2-2.5 times less than the volume of rad-waste generated by means of the second technology. It can be concluded from the above that

  14. Comparison of the dose evaluation methods for criticality accident

    International Nuclear Information System (INIS)

    Shimizu, Yoshio; Oka, Tsutomu

    2004-01-01

    The improvement of the dose evaluation method for criticality accidents is important to rationalize design of the nuclear fuel cycle facilities. The source spectrums of neutron and gamma ray of a criticality accident depend on the condition of the source, its materials, moderation, density and so on. The comparison of the dose evaluation methods for a criticality accident is made. Some methods, which are combination of criticality calculation and shielding calculation, are proposed. Prompt neutron and gamma ray doses from nuclear criticality of some uranium systems have been evaluated as the Nuclear Criticality Slide Rule. The uranium metal source (unmoderated system) and the uranyl nitrate solution source (moderated system) in the rule are evaluated by some calculation methods, which are combinations of code and cross section library, as follows: (a) SAS1X (ENDF/B-IV), (b) MCNP4C (ENDF/B-VI)-ANISN (DLC23E or JSD120), (c) MCNP4C-MCNP4C (ENDF/B-VI). They have consisted of criticality calculation and shielding calculation. These calculation methods are compared about the tissue absorbed dose and the spectrums at 2 m from the source. (author)

  15. Radio-Ecological Conditions of Groundwater in the Area of Uranium Mining and Milling Facility - 13525

    International Nuclear Information System (INIS)

    Titov, A.V.; Semenova, M.P.; Seregin, V.A.; Isaev, D.V.; Metlyaev, E.G.; Glagolev, A.V.; Klimova, T.I.; Sevtinova, E.B.; Zolotukhina, S.B.; Zhuravleva, L.A.

    2013-01-01

    Manmade chemical and radioactive contamination of groundwater is one of damaging effects of the uranium mining and milling facilities. Groundwater contamination is of special importance for the area of Priargun Production Mining and Chemical Association, JSC 'PPMCA', because groundwater is the only source of drinking water. The paper describes natural conditions of the site, provides information on changes of near-surface area since the beginning of the company, illustrates the main trends of contaminators migration and assesses manmade impact on the quality and mode of near-surface and ground waters. The paper also provides the results of chemical and radioactive measurements in groundwater at various distances from the sources of manmade contamination to the drinking water supply areas. We show that development of deposits, mine water discharge, leakages from tailing dams and cinder storage facility changed general hydro-chemical balance of the area, contributed to new (overlaid) aureoles and flows of scattering paragenetic uranium elements, which are much smaller in comparison with natural ones. However, increasing flow of groundwater stream at the mouth of Sukhoi Urulyungui due to technological water infiltration, mixing of natural water with filtration streams from industrial reservoirs and sites, containing elevated (relative to natural background) levels of sulfate-, hydro-carbonate and carbonate- ions, led to the development and moving of the uranium contamination aureole from the undeveloped field 'Polevoye' to the water inlet area. The aureole front crossed the southern border of water inlet of drinking purpose. The qualitative composition of groundwater, especially in the southern part of water inlet, steadily changes for the worse. The current Russian intervention levels of gross alpha activity and of some natural radionuclides including 222 Rn are in excess in drinking water; regulations for fluorine and manganese concentrations are also in excess

  16. Radio-Ecological Conditions of Groundwater in the Area of Uranium Mining and Milling Facility - 13525

    Energy Technology Data Exchange (ETDEWEB)

    Titov, A.V.; Semenova, M.P.; Seregin, V.A.; Isaev, D.V.; Metlyaev, E.G. [FSBU SRC A.I.Burnasyan Federal Medical Biophysical Center of FMBA of Russia, Zhivopisnaya Street, 46, Moscow (Russian Federation); Glagolev, A.V.; Klimova, T.I.; Sevtinova, E.B. [FSESP ' Hydrospecgeologiya' (Russian Federation); Zolotukhina, S.B.; Zhuravleva, L.A. [FSHE ' Centre of Hygiene and Epidemiology no. 107' under FMBA of Russia (Russian Federation)

    2013-07-01

    Manmade chemical and radioactive contamination of groundwater is one of damaging effects of the uranium mining and milling facilities. Groundwater contamination is of special importance for the area of Priargun Production Mining and Chemical Association, JSC 'PPMCA', because groundwater is the only source of drinking water. The paper describes natural conditions of the site, provides information on changes of near-surface area since the beginning of the company, illustrates the main trends of contaminators migration and assesses manmade impact on the quality and mode of near-surface and ground waters. The paper also provides the results of chemical and radioactive measurements in groundwater at various distances from the sources of manmade contamination to the drinking water supply areas. We show that development of deposits, mine water discharge, leakages from tailing dams and cinder storage facility changed general hydro-chemical balance of the area, contributed to new (overlaid) aureoles and flows of scattering paragenetic uranium elements, which are much smaller in comparison with natural ones. However, increasing flow of groundwater stream at the mouth of Sukhoi Urulyungui due to technological water infiltration, mixing of natural water with filtration streams from industrial reservoirs and sites, containing elevated (relative to natural background) levels of sulfate-, hydro-carbonate and carbonate- ions, led to the development and moving of the uranium contamination aureole from the undeveloped field 'Polevoye' to the water inlet area. The aureole front crossed the southern border of water inlet of drinking purpose. The qualitative composition of groundwater, especially in the southern part of water inlet, steadily changes for the worse. The current Russian intervention levels of gross alpha activity and of some natural radionuclides including {sup 222}Rn are in excess in drinking water; regulations for fluorine and manganese

  17. Results of Active Test of Uranium-Plutonium Co-denitration Facility at Rokkasho Reprocessing Plant

    International Nuclear Information System (INIS)

    Numao, Teruhiko; Nakayashiki, Hiroshi; Arai, Nobuyuki; Miura, Susumu; Takahashi, Yoshiharu; Nakamura, Hironobu; Tanaka, Izumi

    2007-01-01

    In the U-Pu co-denitration facility at Rokkasho Reprocessing Plant (RRP), Active Test which composes of 5 steps was performed by using uranium-plutonium nitrate solution that was extracted from spent fuels. During Active Test, two kinds of tests were performed in parallel. One was denitration performance test in denitration ovens, and expected results were successfully obtained. The other was validation and calibration of non-destructive assay (NDA) systems, and expected performances were obtained and their effectiveness as material accountancy and safeguards system was validated. (authors)

  18. Special graphites

    International Nuclear Information System (INIS)

    Leveque, P.

    1964-01-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [fr

  19. Construction of the two-phase critical flow test facility

    International Nuclear Information System (INIS)

    Chung, C. H.; Chang, S. K.; Park, H. S.; Min, K. H.; Choi, N. H.; Kim, C. H.; Lee, S. H.; Kim, H. C.; Chang, M. H.

    2002-03-01

    The two-phase critical test loop facility has been constructed in the KAERI engineering laboratory for the simulation of small break loss of coolant accident entrained with non-condensible gas of SMART. The test facility can operate at 12 MPa of pressure and 0 to 60 C of sub-cooling with 0.5 kg/s of non- condensible gas injection into break flow, and simulate up to 20 mm of pipe break. Main components of the test facility were arranged such that the pressure vessel containing coolant, a test section simulating break and a suppression tank inter-connected with pipings were installed vertically. As quick opening valve opens, high pressure/temperature coolant flows through the test section forming critical two-phase flow into the suppression tank. The pressure vessel was connected to two high pressure N2 gas tanks through a control valve to control pressure in the pressure vessel. Another N2 gas tank was also connected to the test section for the non-condensible gas injection. The test facility operation was performed on computers supported with PLC systems installed in the control room, and test data such as temperature, break flow rate, pressure drop across test section, gas injection flow rate were all together gathered in the data acquisition system for further data analysis. This test facility was classified as a safety related high pressure gas facility in law. Thus the loop design documentation was reviewed, and inspected during construction of the test loop by the regulatory body. And the regulatory body issued permission for the operation of the test facility

  20. Critical review of uranium resources and production capability to 2020

    International Nuclear Information System (INIS)

    Underhill, D.H.

    2002-01-01

    Even with a modest forecast of nuclear power growth for the next 25 years, it is expected that the world uranium requirements will increase. This analysis indicates uranium mine production will continue to be the primary supply of requirements through 2020. Secondary supplies, such as low enriched uranium blended from highly enriched uranium, reprocessing of spent fuel would have to make-up the remaining balance, although the contribution of US and Russian strategic stockpiles is not well known at this time. (author)

  1. Nuclear Criticality Calculation for Determining the Bach Size in a Pyroprocessing Facility

    International Nuclear Information System (INIS)

    Ko, Won Il; Lee, Ho Hee; Chang, Hong Rae; Song, Dae Yong; Kwon, Eun Ha; Jung, Chang Jun; Yoon, Suk Kyun

    2009-01-01

    The criticality analysis in a pyroprocessing facility is very important element for the R and D and the facility design in terms of the determination of batch size of the sub-processes as well as facility safety. Particularly, the determining the batch size is essential at the beginning stage of the R and D. In this report, the criticality analysis was carried out for the subprocesses such as voloxidation, electrolytic reduction, electrorefining and electrowinning process in order to estimate the maximum batch size of each process by using Monte Carlo code (MCNP4/C2). On the whole, the criticality problem could not give a big effect on the batch sizes in the voloxidation, electrolytic reduction and electrorefining. However, it was resulted that permissible amount of nuclear material to prevent the criticality accident in the electrowinning process was about 10kgHM

  2. Nuclear Criticality Calculation for Determining the Bach Size in a Pyroprocessing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Lee, Ho Hee; Chang, Hong Rae; Song, Dae Yong; Kwon, Eun Ha; Jung, Chang Jun; Yoon, Suk Kyun [KAERI, Daejeon (Korea, Republic of)

    2009-01-15

    The criticality analysis in a pyroprocessing facility is very important element for the R and D and the facility design in terms of the determination of batch size of the sub-processes as well as facility safety. Particularly, the determining the batch size is essential at the beginning stage of the R and D. In this report, the criticality analysis was carried out for the subprocesses such as voloxidation, electrolytic reduction, electrorefining and electrowinning process in order to estimate the maximum batch size of each process by using Monte Carlo code (MCNP4/C2). On the whole, the criticality problem could not give a big effect on the batch sizes in the voloxidation, electrolytic reduction and electrorefining. However, it was resulted that permissible amount of nuclear material to prevent the criticality accident in the electrowinning process was about 10kgHM

  3. The chemical industry of uranium in France; L'industrie chimique de l'uranium en France

    Energy Technology Data Exchange (ETDEWEB)

    Goldschmidt, B [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires

    1955-07-01

    The actual CEA program is concerned with the construction of two large graphite reactors, each of those containing at least one hundred tons of uranium metal with nuclear purity. The uranium for these two reactors will be regularly supplied by new resources discovered in France and Madagascar in the last five years. The working and treatment of such ore have led to the creation of an important french industry of which the general outline and principle are described. The operated ores have got different natures and concentration, individual characteristics are described for the main ores.The most high-grade ore are transported to a central plant in Bouchet near Paris; the low-grade ore are concentrated by physical methods or chemical processes of which principles and economy are studied with constancy. The acid processes are the only used until now, although the carbonated alkaline processes has been studied in France. The next following steps after the acid process until the obtention of uranium rich concentrate are described. The purification steps of uranium compounds to nuclear purity material are described as well as the steps to elaborate metal of which the purity grade will be specify. Finally, the economic aspects of uranium production difficulty will be considered in relation with technical progresses which we can expect to achieve in the future. (M.P.)

  4. Studies on the development of special graphite for use in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Bhatia, G.; Aggarwal, R.K.; Saha, M.; Sengupta, P.R.; Mishra, A. [National Physical Lab., New Delhi (India). Carbon Technology Unit

    2002-07-01

    Special graphite is considered as a critical component of the present-day tokamaks wherein it acts as the armour material for plasma-facing components. This graphite is required to possess, besides other characteristics, high values of bulk density, bending strength and electrical and thermal conductivities and a low value of ash content. Since such graphite was not commercially available in the country, efforts to develop it were initiated at the National Physical Laboratory, New Delhi. The basic approach to develop this graphite was based on green coke method of making the high density graphite, wherein the green coke was modified by incorporating in it small amounts of conducting carbon materials, i.e. needle coke, synthetic graphite and natural graphite. The resulting graphites were characterized with respect to various physical characteristics, namely, green density, weight loss, volume shrinkage, linear shrinkage, bulk density, bending strength, Young's modulus and electrical resistivity, etc. The results are described and discussed in the present paper. 6 refs., 2 tabs.

  5. Continued Multicolumns Bioleaching for Low Grade Uranium Ore at a Certain Uranium Deposit

    OpenAIRE

    Gongxin Chen; Zhanxue Sun; Yajie Liu

    2016-01-01

    Bioleaching has lots of advantages compared with traditional heap leaching. In industry, bioleaching of uranium is still facing many problems such as site space, high cost of production, and limited industrial facilities. In this paper, a continued column bioleaching system has been established for leaching a certain uranium ore which contains high fluoride. The analysis of chemical composition of ore shows that the grade of uranium is 0.208%, which is lower than that of other deposits. Howev...

  6. Administrative Aspects of the Criticality Controls Used in Programmes for Basic Criticality Research, Reactor Development and Materials Processing

    Energy Technology Data Exchange (ETDEWEB)

    Wood, D. P.; Giessing, D. F. [Operational Safety Division, USAEC Albuquerque Operations Office, NM (United States)

    1966-05-15

    This paper describes the administrative and procedural aspects of criticality controls used by a field office of the United States Atomic Energy Commission in programmes that include reactor criticals, research and materials testing reactors, and power reactor development. Situations encountered include handling, storing, and processing large quantities of uranium-235 and plutonium-239 of various configurations and compositions in laboratories and operations which gather basic criticality data, processing of fissile material, and varied reactor research and development, programmes including fuel materials. Similar situations exist for uranium-233 and plutonium-238 on a smaller laboratory scale. The administrative controls and interactions of the USAEC field office and the operating contractors, who operate these installations for the USAEC, are outlined. Also, the purpose and scope of the direct examination by USAEC personnel of these contractor facilities are analysed. The programme has been in effect for three years and is believed to be successful in maintaining efficient operations and an acceptable low level of risk of inadvertent criticality. Success of this programme is in good measure due to the close working relationship between the staffs of the USAEC field office and the operating contractors. (author)

  7. Development and engineering plan for graphite spent fuels conditioning program

    International Nuclear Information System (INIS)

    Bendixsen, C.L.; Fillmore, D.L.; Kirkham, R.J.; Lord, D.L.; Phillips, M.B.; Pinto, A.P.; Staiger, M.D.

    1993-09-01

    Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated

  8. A comparative study between transport and criticality safety indexes for fissile uranium nuclearly pure

    Energy Technology Data Exchange (ETDEWEB)

    Moraes da Silva, T. de; Sordi, G.M.A.A. [Instituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN (Brazil)]. e-mail: tmsilva@ipen.br

    2006-07-01

    The international and national standards determine that during the transport of radioactive materials the package to be sent should be identified by labels of risks specifying content, activity and the transport index. The result of the monitoring of the package to 1 meter identifies the transport index, TI, which represents the dose rate to 1 meter of this. The transport index is, by definition, a number that represents a gamma radiation that crosses the superficial layer the radioactive material of the package to 1 meter of distance. For the fissile radioactive material that is the one in which a neutron causes the division of the atom, the international standards specify criticality safety index CSI, which is related with the safe mass of the fissile element. In this work it was determined the respective safe mass for each considered enrichment for the compounds of uranium oxides UO{sub 2}, U{sub 3}O{sub 8} and U{sub 3}Si{sub 2}. In the study of CSI it was observed that the value 50 of the expression 50/N being N the number of packages be transported in subcriticality conditions it represents a fifth part of the safe mass of the element uranium or 9% of the smallest mass critical for a transport not under exclusive use. As conclusion of the accomplished study was observed that the transport index starting from 7% of enrichment doesn't present contribution and that criticality safety index is always greater than the transport index. Therefore what the standards demand to specify, the largest value between both indexes, was clearly identified in this study as being the criticality safety index. (Author)

  9. Sorption and bioreduction of hexavalent uranium at a military facility by the Chesapeake Bay

    International Nuclear Information System (INIS)

    Dong Wenming; Xie Guibo; Miller, Todd R.; Franklin, Mark P.; Oxenberg, Tanya Palmateer; Bouwer, Edward J.; Ball, William P.; Halden, Rolf U.

    2006-01-01

    Directly adjacent to the Chesapeake Bay lies the Aberdeen Proving Ground, a U.S. Army facility where testing of armor-piercing ammunitions has resulted in the deposition of >70,000 kg of depleted uranium (DU) to local soils and sediments. Results of previous environmental monitoring suggested limited mobilization in the impact area and no transport of DU into the nation's largest estuary. To determine if physical and biological reactions constitute mechanisms involved in limiting contaminant transport, the sorption and biotransformation behavior of the radionuclide was studied using geochemical modeling and laboratory microcosms (500 ppb U(VI) initially). An immediate decline in dissolved U(VI) concentrations was observed under both sterile and non-sterile conditions due to rapid association of U(VI) with natural organic matter in the sediment. Reduction of U(VI) to U(IV) occurred only in non-sterile microcosms. In the non-sterile samples, intrinsic bioreduction of uranium involved bacteria of the order Clostridiales and was only moderately enhanced by the addition of acetate (41% vs. 56% in 121 days). Overall, this study demonstrates that the migration of depleted uranium from the APG site into the Chesapeake Bay may be limited by a combination of processes that include rapid sorption of U(VI) species to natural organic matter, followed by slow, intrinsic bioreduction to U(IV). - At the Aberdeen Proving Ground in Maryland, USA, migration of depleted uranium into the Chesapeake Bay is limited by rapid sorption of the radionuclide to natural organic matter followed by slow biological reduction of water-soluble U(VI) to the insoluble and less toxic U(IV) species

  10. Consensus standards utilized and implemented for nuclear criticality safety in Japan

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Okuno, Hiroshi; Naito, Yoshitaka

    1996-01-01

    The fundamental framework for the criticality safety of nuclear fuel facilities regulations is, in many advanced countries, generally formulated so that technical standards or handbook data are utilized to support the licensing safety review and to implement its guidelines. In Japan also, adequacy of the safety design of nuclear fuel facilities is checked and reviewed on the basis of licensing safety review guides. These guides are, first, open-quotes The Basic Guides for Licensing Safety Review of Nuclear Fuel Facilities,close quotes and as its subsidiaries, open-quotes The Uranium Fuel Fabrication Facility Licensing Safety Review Guidesclose quotes and open-quotes The Reprocessing Facility Licensing Safety Review Guides.close quotes The open-quotes Nuclear Criticality Safety Handbook close-quote of Japan and the Technical Data Collection are published and utilized to supply related data and information for the licensing safety review, such as for the Rokkasho reprocessing plant. The well-established technical standards and data abroad such as those by the American Nuclear Society and the American National Standards Institute are also utilized to complement the standards in Japan. The basic principles of criticality safety control for nuclear fuel facilities in Japan are duly stipulated in the aforementioned basic guides as follows: 1. Guide 10: Criticality control for a single unit; 2. Guide 11: Criticality control for multiple units; 3. Guide 12: Consideration for a criticality accident

  11. Method for producing dustless graphite spheres from waste graphite fines

    Science.gov (United States)

    Pappano, Peter J [Oak Ridge, TN; Rogers, Michael R [Clinton, TN

    2012-05-08

    A method for producing graphite spheres from graphite fines by charging a quantity of spherical media into a rotatable cylindrical overcoater, charging a quantity of graphite fines into the overcoater thereby forming a first mixture of spherical media and graphite fines, rotating the overcoater at a speed such that the first mixture climbs the wall of the overcoater before rolling back down to the bottom thereby forming a second mixture of spherical media, graphite fines, and graphite spheres, removing the second mixture from the overcoater, sieving the second mixture to separate graphite spheres, charging the first mixture back into the overcoater, charging an additional quantity of graphite fines into the overcoater, adjusting processing parameters like overcoater dimensions, graphite fines charge, overcoater rotation speed, overcoater angle of rotation, and overcoater time of rotation, before repeating the steps until graphite fines are converted to graphite spheres.

  12. Kinetics of the chain uranium decay

    International Nuclear Information System (INIS)

    Zel'dovich, Ya.B.; Khariton, Yu.B.

    1984-01-01

    The development of chain reaction in uranium mass, when passing through the mass critical value, is studied. It is shown that thermal expanson is a powerful regulating factor, making the limit passing quite safe. The case of convergence of two uranium masses of precritical dimensions is considered in particular. The critical mass value is assessed and it is shown that its gradual increase above the critical value results in the reaction vibrational regime with the period inversely proportional to the square root of uranium supply rate. The role of delaying neutrons in the fission chain reaction development is pointed out

  13. Adverse Condition and Critical Event Prediction in Cranfield Multiphase Flow Facility

    DEFF Research Database (Denmark)

    Egedorf, Søren; Shaker, Hamid Reza

    2017-01-01

    , or even to the environment. To cope with these, adverse condition and critical event prediction plays an important role. Adverse Condition and Critical Event Prediction Toolbox (ACCEPT) is a tool which has been recently developed by NASA to allow for a timely prediction of an adverse event, with low false...... alarm and missed detection rates. While ACCEPT has shown to be an effective tool in some applications, its performance has not yet been evaluated on practical well-known benchmark examples. In this paper, ACCEPT is used for adverse condition and critical event prediction in a multiphase flow facility....... Cranfield multiphase flow facility is known to be an interesting benchmark which has been used to evaluate different methods from statistical process monitoring. In order to allow for the data from the flow facility to be used in ACCEPT, methods such as Kernel Density Estimation (KDE), PCA-and CVA...

  14. Social Licensing in uranium mining: Experiences from the IAEA review of planned Mukju River Uranium Project, Tanzania

    International Nuclear Information System (INIS)

    Schnell, Henry

    2014-01-01

    The IAEA Uranium Production Site Appraisal Team (UPSAT) programme is designed to assist Member States to enhance the operational performance and the occupational, public and environmental health and safety of uranium mining and processing facilities across all phases of the uranium production cycle. These include exploration, resource assessment, mining, processing, waste management, site management and remediation, and final closure.

  15. Western states uranium resource survey

    International Nuclear Information System (INIS)

    Tinney, J.F.

    1977-01-01

    ERDA's National Uranium Resource Evaluation (NURE) program was established to provide a comprehensive description of uranium resources in the United States. To carry out this task, ERDA has contracted with various facilities, including universities, private companies, and state agencies, to undertake projects such as airborne radiometric surveys, geological and geochemical studies, and the development of advanced geophysical technology. LLL is one of four ERDA laboratories systematically studying uranium distribution in surface water, groundwater, and lake and stream sediments. We are specifically responsible for surveying seven western states. This past year we have designed and installed facilities for delayed-neutron counting and neutron-activation analysis, completed seven orientation surveys, and analyzed several thousand field samples. Full-scale reconnaissance surveys began last fall

  16. Occupational exposures to uranium: processes, hazards, and regulations

    International Nuclear Information System (INIS)

    Stoetzel, G.A.; Fisher, D.R.; McCormack, W.D.; Hoenes, G.R.; Marks, S.; Moore, R.H.; Quilici, D.G.; Breitenstein, B.D.

    1981-04-01

    The United States Uranium Registry (USUR) was formed in 1978 to investigate potential hazards from occupational exposure to uranium and to assess the need for special health-related studies of uranium workers. This report provides a summary of Registry work done to date. The history of the uranium industry is outlined first, and the current commercial uranium industry (mining, milling, conversion, enrichment, and fuel fabrication) is described. This description includes information on basic processes and areas of greatest potential radiological exposure. In addition, inactive commercial facilities and other uranium operations are discussed. Regulation of the commercial production industry for uranium fuel is reported, including the historic development of regulations and the current regulatory agencies and procedures for each phase of the industry. A review of radiological health practices in the industry - facility monitoring, exposure control, exposure evaluation, and record-keeping - is presented. A discussion of the nonradiological hazards of the industry is provided, and the final section describes the tissue program developed as part of the Registry

  17. Application of 3D coupled code ATHLET-QUABOX/CUBBOX for RBMK-1000 transients after graphite block modernization

    Energy Technology Data Exchange (ETDEWEB)

    Samokhin, Aleksei [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS), Moscow (Russian Federation); Zilly, Matias [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    This work describes the application and the results of transient calculations for the RBMK-1000 with the coupled code system ATHLET 2.2A-QUABOX/CUBBOX which was developed in GRS. Within these studies the planned modernization of the graphite blocks of the RBMK-1000 reactor is taken into account. During the long-term operation of the uranium-graphite reactors RBMK-1000, a change of physical and mechanical properties of the reactor graphite blocks is observed due to the impact of radiation and temperature effects. These have led to a deformation of the reactor graphite columns and, as a result, a deformation of the control and protection system (CPS) and of fuel channels. Potentially, this deformation can lead to problems affecting the smooth movement of the control rods in the CPS channels and problems during the loading and unloading of fuel assemblies. The present paper analyzes two reactivity insertion transients, each taking into account three graphite removal scenarios. The presented work is directly connected with the modernization program of the RBMK- 1000 reactors and has an important contribution to the assessment of the safety-relevant parameters after the modification of the core graphite blocks.

  18. Structural disorder of graphite and implications for graphite thermometry

    Science.gov (United States)

    Kirilova, Martina; Toy, Virginia; Rooney, Jeremy S.; Giorgetti, Carolina; Gordon, Keith C.; Collettini, Cristiano; Takeshita, Toru

    2018-02-01

    Graphitization, or the progressive maturation of carbonaceous material, is considered an irreversible process. Thus, the degree of graphite crystallinity, or its structural order, has been calibrated as an indicator of the peak metamorphic temperatures experienced by the host rocks. However, discrepancies between temperatures indicated by graphite crystallinity versus other thermometers have been documented in deformed rocks. To examine the possibility of mechanical modifications of graphite structure and the potential impacts on graphite thermometry, we performed laboratory deformation experiments. We sheared highly crystalline graphite powder at normal stresses of 5 and 25 megapascal (MPa) and aseismic velocities of 1, 10 and 100 µm s-1. The degree of structural order both in the starting and resulting materials was analyzed by Raman microspectroscopy. Our results demonstrate structural disorder of graphite, manifested as changes in the Raman spectra. Microstructural observations show that brittle processes caused the documented mechanical modifications of the aggregate graphite crystallinity. We conclude that the calibrated graphite thermometer is ambiguous in active tectonic settings.

  19. Asymptomatic Intracorneal Graphite Deposits following Graphite Pencil Injury

    OpenAIRE

    Philip, Swetha Sara; John, Deepa; John, Sheeja Susan

    2012-01-01

    Reports of graphite pencil lead injuries to the eye are rare. Although graphite is considered to remain inert in the eye, it has been known to cause severe inflammation and damage to ocular structures. We report a case of a 12-year-old girl with intracorneal graphite foreign bodies following a graphite pencil injury.

  20. Evaluation of activities of carbons in chemical equilibrium with uranium carbonitride

    International Nuclear Information System (INIS)

    Katsura, Masahiro; Hirota, Masayuki; Miyake, Masanobu; Hamada, Kazuo.

    1992-01-01

    A mixture of uranium sesquinitride and carbon was prepared by the reaction of UC of UC 2 with N 2 in the temperature range from 700 to 1400degC. When the mixture of uranium sesquinitride and carbon is kept at temperatures above 1200degC in the atmosphere of N 2 at low pressure, the state where uranium carbonitride (UC 1-x N x ) and carbon are present together in chemical equilibrium will be established. A thermodynamic analysis suggests that, in the equilibrium state, the composition of UC 1-x N x is determined by the chemical activity of carbon, a c , which is related to the chemical potential of the carbon, μ c , by the equation, μ c = μ c deg + RT 1n a c . Here μ c deg refers to graphite, which is usually taken as the standard state of carbon (a c = 1). Mixtures of U 2 N 3 and carbon with several degrees of graphitization were heat-treated at 1400degC, and the composition of UC 1-x N x in the reaction product was determined. From these experimental results and the thermodynamic analysis, values of the activity of the carbon coexisting with UC 1-x N x were estimated. (author)

  1. A Summary Report on Assembly 3 of FR0

    International Nuclear Information System (INIS)

    Andersson, T.L.; Brunfelter, B.; Cecchi, P.F.; Hellstrand, E.; Kockum, J.; Londen, S.O.; Tiren, L.I.

    1966-06-01

    The third core of the zero energy fast reactor FR0 consisted of 20 % enriched uranium diluted to 29 vol. % with graphite and had a volume of 50 litres. Like previous cores it was surrounded by a thick copper reflector. The report summarizes measurements of critical mass, control rod reactivities, fine structure flux variations and conversion ratio. In particular, effects associated with the heterogeneous arrangement of the uranium and graphite plates are examined

  2. A Summary Report on Assembly 3 of FR0

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, T L; Brunfelter, B; Cecchi, P F; Hellstrand, E; Kockum, J; Londen, S O; Tiren, L I

    1966-06-15

    The third core of the zero energy fast reactor FR0 consisted of 20 % enriched uranium diluted to 29 vol. % with graphite and had a volume of 50 litres. Like previous cores it was surrounded by a thick copper reflector. The report summarizes measurements of critical mass, control rod reactivities, fine structure flux variations and conversion ratio. In particular, effects associated with the heterogeneous arrangement of the uranium and graphite plates are examined.

  3. Selective leaching of uranium from uranium-contaminated soils: Progress report 1

    International Nuclear Information System (INIS)

    Francis, C.W.; Mattus, A.J.; Farr, L.L.; Elless, M.P.; Lee, S.Y.

    1993-02-01

    Three soils and a sediment contaminated with uranium were used to determine the effectiveness of sodium carbonate and citric acid leaching to decontaminated or remove uranium to acceptable regulatory levels. Two of the soils were surface soils from the DOE facility formerly called the Feed Materials Production Center (FMPC) at Fernald, Ohio. This facility is presently called the Femald Environmental Management Project (FEMP). Carbonate extractions generally removed from 70 to 90% of the uranium from the Fernald storage pad soil. Uranium was slightly more difficult to extract from the Fernald incinerator and the Y-12 landfarm soils. Very small amounts of uranium could be extracted from the storm sewer sediment. Extraction with carbonate at high solution-to-soil ratios were as effective as extractions at low solution-to-soil ratios, indicating attrition by the paddle mixer was not significantly different than that provided in a rotary extractor. Also, pretreatments such as milling or pulverizing the soil sample did not appear to increase extraction efficiency when carbonate extractions were carried out at elevated temperatures (60 degree C) or long extraction times (23 h). Adding KMnO 4 in the carbonate extraction appeared to increase extraction efficiency from the Fernald incinerator soil but not the Fernald storage pad soil. The most effective leaching rates (> 90 % from both Fernald soils) were obtained using a citrate/dithionite extraction procedure designed to remove amorphous (noncrystalline) iron/aluminum sesquioxides from surfaces of clay minerals. Citric acid also proved to be a very good extractant for uranium

  4. From USA operation experience of industrial uranium-graphite reactors

    International Nuclear Information System (INIS)

    Burdakov, N.S.

    1996-01-01

    The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors

  5. Uranium material removing and recovering device

    International Nuclear Information System (INIS)

    Takita, Shin-ichi.

    1997-01-01

    A uranium material removing and recovering device for use in removing surplus uranium heavy metal (UO 2 ) generated in a uranium handling facility comprises a uranium material removing device and a uranium material recovering device. The uranium material removing device comprises an adsorbing portion filled with a uranium adsorbent, a control portion for controlling the uranium adsorbent of the uranium adsorbing portion by a controlling agent, a uranium adsorbing device connected thereto and a jetting device for jetting the adsorbing liquid to equipments deposited with uranium. The recovering device comprises a recovering apparatus for recovering uranium materials deposited with the adsorbent liquid removed by the jetting device and a recovering tank for storing the recovered uranium materials. The device of the present invention can remove surplus uranium simply and safely, mitigate body's load upon removing and recovering operations, facilitate the processing for the exchange of the adsorbent and reduces the radioactive wastes. (T.M.)

  6. Nuclear fuel cycle head-end enriched uranium purification and conversion into metal

    International Nuclear Information System (INIS)

    Bonini, A.; Cabrejas, J.; Lio, L. de; Dell'Occhio, L.; Devida, C.; Dupetit, G.; Falcon, M.; Gauna, A.; Gil, D.; Guzman, G.; Neuringer, P.; Pascale, A.; Stankevicius, A.

    1998-01-01

    The CNEA (Comision Nacional de Energia Atomica - Argentina) operated two facilities at the Ezeiza Atomic Center which supply purified enriched uranium employed in the production of nuclear fuels. At one of those facilities, the Triple Height Laboratory scraps from the production of MTR type fuel elements (mainly out of specification U 3 O 8 plates or powder) are purified to nuclear grade. The purification is accomplished by a solvent extraction process. The other facility, the Enriched Uranium Laboratory produces 90% enriched uranium metal to be used in Mo 99 production (originally the uranium was used for the manufacture of MTR fuel elements made of aluminium-uranium alloy). This laboratory also provided metallic uranium with a lower enrichment (20%) for a first uranium-silicon testing fuel element, and in the near future it is going to recommence 20% enriched uranium related activities in order to provide the metal for the silicon-based fuel elements production (according to the policy of enrichment reduction for MTR reactors). (author)

  7. Radiological aspects of in situ uranium recovery

    International Nuclear Information System (INIS)

    BROWN, STEVEN H.

    2007-01-01

    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), have enabled commercial scale mining and milling of relatively small ore pockets of lower grade, and may 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 1975. Solution mining involves the pumping 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. Processing involves ion exchange and may also include precipitation, drying or calcining and packaging operations depending on facility specifics. This paper presents an overview of the ISR process and the health physics monitoring programs developed at a number of commercial scale ISL / ISR Uranium recovery and production facilities as a result of the radiological character of these processes. Although many radiological aspects of the process are similar to that of conventional mills, conventional-type tailings as such are not generated. However, liquid and solid byproduct materials may be generated and impounded. The quantity and radiological character of these by products are related to facility specifics. Some special monitoring considerations are presented which are required due to the manner in which Radon gas is evolved in

  8. Study of reactions for the production of uranium titrafluoride and uranium hexafluoride

    International Nuclear Information System (INIS)

    Guzella, M.F.R.

    1985-01-01

    The main production processes of uranium hexafluoride in pilot plants and industrial facilities are described. The known reactions confirmed in laboratory experiments that lead to Uf 6 or other intermediate fluorides are discussed. For the purpose of determining a thermodinamically feasible reaction involving the sulfur hexafluoride as fluorinating agent, a mock-up facility was designed and constructed as a part of the R and D work planned at the CDTN (Nuclebras Center for Nuclear Technology Development). IN the uranium tatrafluoride synthesis employing U 3 O 8 and SF 6 several experimental parameters are studied. The reaction time, gasflow, temperature and stoechiometic relations among reagents are described in detail. (Author) [pt

  9. Criticality safety evaluations - a {open_quotes}stalking horse{close_quotes} for integrated safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R.A. [Westinghouse Electric Corp., Columbia, SC (United States)

    1995-12-31

    The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility`s criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE.

  10. Benchmark experiments at ASTRA facility on definition of space distribution of 235U fission reaction rate

    International Nuclear Information System (INIS)

    Bobrov, A. A.; Boyarinov, V. F.; Glushkov, A. E.; Glushkov, E. S.; Kompaniets, G. V.; Moroz, N. P.; Nevinitsa, V. A.; Nosov, V. I.; Smirnov, O. N.; Fomichenko, P. A.; Zimin, A. A.

    2012-01-01

    Results of critical experiments performed at five ASTRA facility configurations modeling the high-temperature helium-cooled graphite-moderated reactors are presented. Results of experiments on definition of space distribution of 235 U fission reaction rate performed at four from these five configurations are presented more detail. Analysis of available information showed that all experiments on criticality at these five configurations are acceptable for use them as critical benchmark experiments. All experiments on definition of space distribution of 235 U fission reaction rate are acceptable for use them as physical benchmark experiments. (authors)

  11. Starting up a programme of atomic piles using compressed gas

    International Nuclear Information System (INIS)

    Horowitz, J.; Yvon, J.

    1959-01-01

    1) An examination of the intellectual and material resources which have directed the French programme towards: a) the natural uranium and plutonium system, b) the use of compressed gas as heat transfer fluid (primary fluid). 2) The parts played in exploring the field by the pile EL2 and G1, EL2 a natural uranium, heavy water and compressed gas pile, G1 a natural uranium, graphite and atmospheric air pile. 3) Development of the neutronics of graphite piles: physical study of G1. 4) The examination of certain problem posed by centres equipped with natural uranium, graphite and compressed carbon dioxide piles: structure, special materials, fluid circuits, maximum efficiency. Economic aspects. 5) Aids to progress: a) piles for testing materials and for tests on canned fuel elements, b) laboratory and calculation facilities. 6) Possible new orientations of compressed gas piles: a) raising of the pressure, b) enriched fuel, c) higher temperatures, d) use of heavy water. (author) [fr

  12. Structural disorder of graphite and implications for graphite thermometry

    Directory of Open Access Journals (Sweden)

    M. Kirilova

    2018-02-01

    Full Text Available Graphitization, or the progressive maturation of carbonaceous material, is considered an irreversible process. Thus, the degree of graphite crystallinity, or its structural order, has been calibrated as an indicator of the peak metamorphic temperatures experienced by the host rocks. However, discrepancies between temperatures indicated by graphite crystallinity versus other thermometers have been documented in deformed rocks. To examine the possibility of mechanical modifications of graphite structure and the potential impacts on graphite thermometry, we performed laboratory deformation experiments. We sheared highly crystalline graphite powder at normal stresses of 5 and 25  megapascal (MPa and aseismic velocities of 1, 10 and 100 µm s−1. The degree of structural order both in the starting and resulting materials was analyzed by Raman microspectroscopy. Our results demonstrate structural disorder of graphite, manifested as changes in the Raman spectra. Microstructural observations show that brittle processes caused the documented mechanical modifications of the aggregate graphite crystallinity. We conclude that the calibrated graphite thermometer is ambiguous in active tectonic settings.

  13. Organic free radicals and micropores in solid graphitic carbonaceous matter at the Oklo natural fission reactors, Gabon

    International Nuclear Information System (INIS)

    Rigali, M.J.; Nagy, B.

    1997-01-01

    The presence, concentration, and distribution of organic free radicals as well as their association with specific surface areas and microporosities help characterize the evolution and behavior of the Oklo carbonaceous matter. Such information is necessary in order to evaluate uranium mineralization, liquid bitumen solidification, and radio nuclide containment at Oklo. In the Oklo ore deposits and natural fission reactors carbonaceous matter is often referred to as solid graphitic bitumen. The carbonaceous parts of the natural reactors may contain as much as 65.9% organic C by weight in heterogeneous distribution within the clay-rich matrix. The solid carbonaceous matter immobilized small uraninite crystals and some fission products enclosed in this uraninite and thereby facilitated radio nuclide containment in the reactors. Hence, the Oklo natural fission reactors are currently the subjects of detailed studies because they may be useful analogues to support performance assessment of radio nuclide containment at anthropogenic radioactive waste repository sites. Seven carbonaceous matter rich samples from the 1968 ± 50 Ma old natural fission reactors and the associated Oklo uranium ore deposit were studied by electron spin resonance (ESR) spectroscopy and by measurements of specific surface areas (BET method). Humic acid, fulvic acid, and fully crystalline graphite standards were also examined by ESR spectroscopy for comparison with the Oklo solid graphitic bitumens. With one exception, the ancient Oklo bitumens have higher organic free radical concentrations than the modem humic and fulvic acid samples. The presence of carbon free radicals in the graphite standard could not be determined due to the conductivity of this material. 72 refs., 7 figs., 1 tab

  14. Fluid inclusion and oxygen isotope studies of the Nabarlek and Jabiluka uranium deposits, Northern Territory, Australia

    International Nuclear Information System (INIS)

    Ypma, P.J.M.; Fuzikawa, K.

    1980-01-01

    We lack a basic understanding of the solutions producing the uranium deposits of the Alligator Rivers Uranium Field (ARUF). Several theories have been proposed ranging from syngenetic, epigenetic hydrothermal, epigenetic metamorphogenic, surficial origin (Ferguson et al., this volume), and mobilization by evaporite deposits. As for a precipitation mechanism, we do not seem to find much beyond the presence of graphite in some ore-bearing and intra-formational strata, and pre-uranium sulphides, none of which reducing factors are common throughout all ore bodies. This study was initiated with the aim of obtaining direct fluid inclusion evidence of the solution transport and precipitation of uranium

  15. Criticality assessment of LLRWDF closure

    International Nuclear Information System (INIS)

    Sarrack, A.G.; Weber, J.H.; Woody, N.D.

    1992-01-01

    During the operation of the Low Level Radioactive Waste Disposal Facility (LLRWDF), large amounts (greater than 100 kg) of enriched uranium (EU) were buried. This EU came primarily from the closing and decontamination of the Naval Fuels Facility in the time period from 1987--1989. Waste Management Operations (WMO) procedures were used to keep the EU boxes separated to prevent possible criticality during normal operation. Closure of the LLRWDF is currently being planned, and waste stabilization by Dynamic Compaction (DC) is proposed. Dynamic compaction will crush the containers in the LLRWDF and result in changes in their geometry. Research of the LLRWDF operations and record keeping practices have shown that the EU contents of trenches are known, but details of the arrangement of the contents cannot be proven. Reviews of the trench contents, combined with analysis of potential critical configurations, revealed that some portions of the LLRWDF can be expected to be free of criticality concerns while other sections have credible probabilities for the assembly of a critical mass, even in the uncompacted configuration. This will have an impact on the closure options and which trenches can be compacted

  16. Structural performance of a graphite blanket in fusion reactors

    International Nuclear Information System (INIS)

    Wolfer, W.G.; Watson, R.D.

    1978-01-01

    Irradiation of graphite in a fusion reactor causes dimensional changes, enhanced creep, and changes in elastic properties and fracture strength. Temperature and flux gradients through the graphite blanket structure produce differential distortions and stress gradients. An inelastic stress analysis procedure is described which treats these variations of the graphite properties in a consistent manner as dictated by physical models for the radiation effects. Furthermore, the procedure follows the evolution of the stress and fracture strength distributions during the reactor operation as well as for possible shutdowns at any time. The lifetime of the graphite structure can be determined based on the failure criterion that the stress at any location exceeds one-half of the fracture strength. This procedure is applied to the most critical component of the blanket module in the SOLASE design

  17. Actinides in irradiated graphite of RBMK-1500 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Plukienė, R., E-mail: rita@ar.fi.lt; Plukis, A.; Barkauskas, V.; Gudelis, A.; Gvozdaitė, R.; Duškesas, G.; Remeikis, V.

    2014-10-01

    type reactors, especially for {sup 244}Cm estimation which is a critical parameter contributing to the total alpha activity of the irradiated reactor graphite for approximately 200 years.

  18. Radiochronological Age of a Uranium Metal Sample from an Abandoned Facility

    Energy Technology Data Exchange (ETDEWEB)

    Meyers, L A; Williams, R W; Glover, S E; LaMont, S P; Stalcup, A M; Spitz, H B

    2012-03-16

    A piece of scrap uranium metal bar buried in the dirt floor of an old, abandoned metal rolling mill was analyzed using multi-collector inductively coupled plasma mass spectroscopy (MC-ICP-MS). The mill rolled uranium rods in the 1940s and 1950s. Samples of the contaminated dirt in which the bar was buried were also analyzed. The isotopic composition of uranium in the bar and dirt samples were both the same as natural uranium, though a few samples of dirt also contained recycled uranium; likely a result of contamination with other material rolled at the mill. The time elapsed since the uranium metal bar was last purified can be determined by the in-growth of the isotope {sup 230}Th from the decay of {sup 234}U, assuming that only uranium isotopes were present in the bar after purification. The age of the metal bar was determined to be 61 years at the time of this analysis and corresponds to a purification date of July 1950 {+-} 1.5 years.

  19. Radiochronological age of a uranium metal sample from an abandoned facility

    International Nuclear Information System (INIS)

    Meyers, L.A.; Stalcup, A.M.; Glover, S.E.; Spitz, H.B.

    2013-01-01

    A piece of scrap uranium metal bar buried in the dirt floor of an old, abandoned metal rolling mill was analyzed using multi-collector inductively coupled plasma mass spectroscopy (MC-ICP-MS). The mill rolled uranium rods in the 1940 and 1950s. Samples of the contaminated dirt in which the bar was buried were also analyzed. The isotopic composition of uranium in the bar and dirt samples were both the same as natural uranium, though a few samples of dirt also contained recycled uranium; likely a result of contamination with other material rolled at the mill. The time elapsed since the uranium metal bar was last purified can be determined by the in-growth of the isotope 230 Th from the decay of 234 U, assuming that only uranium isotopes were present in the bar after purification. The age of the metal bar was determined to be 61 years at the time of this analysis and corresponds to a purification date of July 1950 ± 1.5 years. (author)

  20. Radiochronological Age of a Uranium Metal Sample from an Abandoned Facility

    International Nuclear Information System (INIS)

    Meyers, L.A.; Williams, R.W.; Glover, S.E.; LaMont, S.P.; Stalcup, A.M.; Spitz, H.B.

    2012-01-01

    A piece of scrap uranium metal bar buried in the dirt floor of an old, abandoned metal rolling mill was analyzed using multi-collector inductively coupled plasma mass spectroscopy (MC-ICP-MS). The mill rolled uranium rods in the 1940s and 1950s. Samples of the contaminated dirt in which the bar was buried were also analyzed. The isotopic composition of uranium in the bar and dirt samples were both the same as natural uranium, though a few samples of dirt also contained recycled uranium; likely a result of contamination with other material rolled at the mill. The time elapsed since the uranium metal bar was last purified can be determined by the in-growth of the isotope 230 Th from the decay of 234 U, assuming that only uranium isotopes were present in the bar after purification. The age of the metal bar was determined to be 61 years at the time of this analysis and corresponds to a purification date of July 1950 ± 1.5 years.

  1. SYNER-G buildings, lifelines, transportation networks and critical facilities

    CERN Document Server

    Pitilakis, K; Kaynia, AM

    2014-01-01

    This volume collates serviceable fragility functions from an international selection of seismic risk probability research, developing and adapting them to a European context to improve damage forecasts for buildings, infrastructure, and critical facilities.

  2. Chemical thermodynamics of uranium

    International Nuclear Information System (INIS)

    Grenthe, I.; Fuger, J.; Lemire, R.J.; Muller, A.B.; Nguyen-Trung Cregu, C.; Wanner, H.

    1992-01-01

    A comprehensive overview on the chemical thermodynamics of those elements that are of particular importance in the safety assessment of radioactive waste disposal systems is provided. This is the first volume in a series of critical reviews to be published on this subject. The book provides an extensive compilation of chemical thermodynamic data for uranium. A description of procedures for activity corrections and uncertainty estimates is given. A critical discussion of data needed for nuclear waste management assessments, including areas where significant gaps of knowledge exist is presented. A detailed inventory of chemical thermodynamic data for inorganic compounds and complexes of uranium is listed. Data and their uncertainty limits are recommended for 74 aqueous complexes and 199 solid and 31 gaseous compounds containing uranium, and on 52 aqueous and 17 solid auxiliary species containing no uranium. The data are internally consistent and compatible with the CODATA Key Values. The book contains a detailed discussion of procedures used for activity factor corrections in aqueous solution, as well as including methods for making uncertainty estimates. The recommended data have been prepared for use in environmental geochemistry. Containing contributions written by experts the chapters cover various subject areas such a s: oxide and hydroxide compounds and complexes, the uranium nitrides, the solid uranium nitrates and the arsenic-containing uranium compounds, uranates, procedures for consistent estimation of entropies, gaseous and solid uranium halides, gaseous uranium oxides, solid phosphorous-containing uranium compounds, alkali metal uranates, uncertainties, standards and conventions, aqueous complexes, uranium minerals dealing with solubility products and ionic strength corrections. The book is intended for nuclear research establishments and consulting firms dealing with uranium mining and nuclear waste disposal, as well as academic and research institutes

  3. Long-term fate and transport of arsenic in an in-pit uranium mine tailings facility

    International Nuclear Information System (INIS)

    Moldovan, B.; Hendry, M.J.

    2006-01-01

    An important environmental issue facing the uranium mining industry in Saskatchewan is the quantification of the long-term migration of arsenic from its tailings facilities to the adjacent groundwater system. Decommissioning of these arsenic-rich tailings requires that the long-term arsenic source term for the tailings to the groundwater be defined. To meet this need, arsenic-rich uranium mine tailings from one in-pit tailings facility (tailings emplaced in a mined out open pit) were studied in detail. The tailings facility selected for study was the Rabbit Lake in-pit tailings management facility (RLITMF) in northern Saskatchewan, Canada. The tailings body in the RLITMF is 425 m long x 300 m wide x 100 m deep at its center and mill tailings were deposited in layers between 1985 (base) and 2004 (top). Associated with the low-level radioactive tailings is approximately 23,000 tonnes of arsenic. The in-pit design limits solute transport in these fine-grained tailings to diffusion. Because the layers of tailings have varying chemical characteristics (controlled by the ore being milled at the time), the total arsenic concentrations in the layers and their associated pore fluids range from 56 to 9,871 μ/g and 0.24 to 140 mg/l, respectively. As was the case for arsenic, the concentration of iron present in the layers was also variable (ranging from 8,967 to 30,247 μ/g). Synchrotron-based studies show that the arsenic in these tailings is strongly attenuated by adsorption to secondary 2-line ferrihydrite through inner sphere bidentate linkages. Single reservoir diffusion cell testing shows that the effective diffusion coefficient for arsenic in the tailings is 4.5 x 10 -10 m 2 s- 1 . Based on results from our field- and laboratory-based studies, the redistribution (via diffusion) and attenuation (via adsorption) of arsenic in the RLITMF was modelled using a one-dimensional geochemical reactive transport model to provide a source term for arsenic migration from the

  4. Decommissioning an uranium and thorium facility: a radiation protection approach

    International Nuclear Information System (INIS)

    Feijo Vasques, Francisco Mario; Saburo Todo, Alberto; Mestre, Paulo Antonio

    2008-01-01

    Decommissioning means actions taken at the end of the useful life of a facility in retiring it from service with adequate regard for the health and safety of workers and members of the public. In the present work, we introduce a radiation protection approach for the removal of radioactive material to the extent that the facility or site becomes available for use without restriction. The facility to be decommissioned is a fuel cycle pilot plant that operated with natural uranium and thorium for almost two decades and then, kept inactive for about 10 years at the Nuclear and Energy Research Institute - IPEN. Even after this long period of inactivity, it has presented significant levels of radiation and contamination spread over the floor, walls, windows, doors and ceiling. The fuel cycle pilot plant was completely dismantled, remaining only the walls and the concrete structures. In this work we present the job done to restore the area. According to each step of dismantling a continuous monitoring of the contaminated surfaces was carried out including the survey of the deep material from the floor and walls. The material identified as radioactive waste was stored into appropriated metal drums. A radiation protection team guided this stage of the work, prescribing the tasks, and the amount of material that should be removed from floors, windows and ceiling. For this, repetitive surveys had to be done. The results of monitoring and contamination levels were analysed, thus guiding the next steps of the job. In this way radiation protection team took over the tasks, running the work with the purpose of achieving acceptable levels of radiation, restoring the area for unrestricted use. (author)

  5. Draft Environmental Impact Statement for Construction and Operation of a Depleted Uranium Hexafluoride Conversion Facility at the Paducah, Kentucky, Site

    International Nuclear Information System (INIS)

    2003-01-01

    This document is a site-specific environmental impact statement (EIS) for construction and operation of a proposed depleted uranium hexafluoride (DUF 6 ) conversion facility at the U.S. Department of Energy (DOE) Paducah site in northwestern Kentucky (Figure S-1). The proposed facility would convert the DUF 6 stored at Paducah to a more stable chemical form suitable for use or disposal. In a Notice of Intent (NOI) published in the ''Federal Register'' (FR) on September 18, 2001 (''Federal Register'', Volume 66, page 48123 [66 FR 48123]), DOE announced its intention to prepare a single EIS for a proposal to construct, operate, maintain, and decontaminate and decommission two DUF 6 conversion facilities at Portsmouth, Ohio, and Paducah, Kentucky, in accordance with the National Environmental Policy Act of 1969 (NEPA) (''United States Code'', Title 42, Section 4321 et seq. [42 USC 4321 et seq.]) and DOE's NEPA implementing procedures (''Code of Federal Regulations'', Title 10, Part 1021 [10 CFR Part 1021]). Subsequent to award of a contract to Uranium Disposition Services, LLC (hereafter referred to as UDS), Oak Ridge, Tennessee, on August 29, 2002, for design, construction, and operation of DUF 6 conversion facilities at Portsmouth and Paducah, DOE reevaluated its approach to the NEPA process and decided to prepare separate site-specific EISs. This change was announced in a ''Federal Register'' Notice of Change in NEPA Compliance Approach published on April 28, 2003 (68 FR 22368); the Notice is included as Attachment B to Appendix C of this EIS. This EIS addresses the potential environmental impacts from the construction, operation, maintenance, and decontamination and decommissioning (DandD) of the proposed conversion facility at three alternative locations within the Paducah site; from the transportation of depleted uranium conversion products to a disposal facility; and from the transportation, sale, use, or disposal of the fluoride-containing conversion products

  6. Inert annealing of irradiated graphite by inductive heating

    International Nuclear Information System (INIS)

    Botzem, W.; Woerner, J.

    2001-01-01

    Fission neutrons change physical properties of graphite being used in nuclear reactors as moderator and as structural material. The understanding of these effects on an atomic model is expressed by dislocations of carbon atoms within the graphite and the thereby stored energy is known as Wigner Energy. The dismantling of the Pile 1 core may necessitate the thermal treatment of the irradiated but otherwise undamaged graphite. This heat treatment - usually called annealing - initiates the release of stored Wigner Energy in a controlled manner. This energy could otherwise give rise to an increase in temperature under certain conditions during transport or preparation for final storage. In order to prevent such an effect it is intended to anneal the major part of Pile 1 graphite before it is packed into boxes suitable for final disposal. Different heating techniques have been assessed. Inductive heating in an inert atmosphere was selected for installation in the Pile 1 Waste Processing Facility built for the treatment and packaging of the dismantled Pile 1 waste. The graphite blocks will be heated up to 250 deg. C in the annealing ovens, which results in the release of significant amount of the stored energy. External heat sources in a final repository will never heat up the storage boxes to such a temperature. (author)

  7. DOE's Stewardship of Government-Owned Uranium Materials

    International Nuclear Information System (INIS)

    Jackson, J. Dale; Donaldson, Dale E.

    2002-01-01

    Beginning in the 1980's, a significant number of Department of Energy facilities have been shut down and are in the decommissioning process. The shutdown of additional facilities is planned. In addition, during the past several decades, the Department of Energy has loaned nuclear material to a wide variety of private and governmental institutions for research and educational purposes. Subsequent changes in the Department's priorities have reduced the need for nuclear materials to support the Department's programs. Similarly, there has been a reduction in the need for borrowed nuclear materials by organizations and institutions using nuclear materials 'on loan' from the Department. As a result, inventories of uranium material from the Department's facilities and 'on loan' must be removed and returned to the Department. This material is in the form of low enriched uranium (LEU), normal uranium (NU), and depleted uranium (DU) in various forms. This uranium material is located at over one hundred sites within the United States and overseas, including universities and laboratories. Much of this uranium is not needed to support national priorities and programs. The Department of Energy has assumed a stewardship role in managing nuclear materials throughout their life cycle, from acquisition to storage. Surplus uranium has created challenges for DOE in managing and storing the material as well as identifying opportunities for its further use. On behalf of the Department, the Oak Ridge Operations Office has been given the responsibility to implement the Department responsibilities in meeting these challenges and managing the Department's uranium materials. To support this effort, the Office of Nuclear Fuel Security and Uranium Technology within the ORO complex coordinates uranium management functions across the Department of Energy. This coordination provides DOE with a number of important benefits, among which are: consolidated management and storage of uranium; improved

  8. Critical and sub-critical experiments on U-BeO lattices

    International Nuclear Information System (INIS)

    Benoist, P.; Gourdon, Ch.; Martelly, J.; Sagot, M.; Wanner, G.

    1958-01-01

    Sub-critical experiments have allowed us to measure the material buckling of uranium natural oxide of beryllium lattices with a grid of 15 cm, and made up of uranium bars measuring 2.60 - 2.92 - 3.56 and 4.40 cm of diameter. A critical experiment has then been conducted with hollow 1.35 per cent enriched uranium bars. A study of U-BeO 18.03 cm grid lattices is at present being conducted. (author) [fr

  9. Uranium mill tailings management

    International Nuclear Information System (INIS)

    1982-01-01

    Facilities for the disposal of uranium mill tailings will invariably be subjected to geomorphological and climatological influences in the long-term. Proceedings of a workshop discuss how the principles of geomorphology can be applied to the siting, design, construction, decommissioning and rehabilitation of disposal facilities in order to provide for long-term containment and stability of tailings. The characteristics of tailings and their behaviour after disposal influence the potential impacts which might occur in the long-term. Proceedings of another workshop examine the technologies for uranium ore processing and tailings conditioning with a view to identifying improvements that could be made in such characteristics

  10. In-situ electric resistance measurements and annealing effects of graphite exposed to swift heavy ions

    International Nuclear Information System (INIS)

    Fernandes, Sandrina; Pellemoine, Frederique; Tomut, Marilena; Avilov, Mikhail; Bender, Markus; Boulesteix, Marine; Krause, Markus; Mittig, Wolfgang; Schein, Mike; Severin, Daniel; Trautmann, Christina

    2013-01-01

    To study the suitability of using graphite as material for high-power targets for rare isotope production at the future Facility for Rare Isotope Beams (FRIB) in the USA and at the Facility for Antiproton and Ion Research (FAIR) in Germany, thin foils of polycrystalline graphite were exposed to 8.6-MeV/u Au ions reaching a maximum fluence of 1 × 10 15 ions/cm 2 . Foil irradiation temperatures of up to 1800 °C were obtained by ohmic heating. In-situ monitoring of the electrical resistance of the graphite foils during and after irradiation provided information on beam-induced radiation damage. The rate of electrical resistance increase as a function of fluence was found to decrease with increasing irradiation temperature, indicating a more efficient annealing of the irradiation-produced defects. This is corroborated by the observation that graphite foils irradiated at temperatures below about 800 °C showed cracks and pronounced deformations, which did not appear on the samples irradiated at higher temperatures

  11. Uranium mining

    International Nuclear Information System (INIS)

    2008-01-01

    Full text: The economic and environmental sustainability of uranium mining has been analysed by Monash University researcher Dr Gavin Mudd in a paper that challenges the perception that uranium mining is an 'infinite quality source' that provides solutions to the world's demand for energy. Dr Mudd says information on the uranium industry touted by politicians and mining companies is not necessarily inaccurate, but it does not tell the whole story, being often just an average snapshot of the costs of uranium mining today without reflecting the escalating costs associated with the process in years to come. 'From a sustainability perspective, it is critical to evaluate accurately the true lifecycle costs of all forms of electricity production, especially with respect to greenhouse emissions, ' he says. 'For nuclear power, a significant proportion of greenhouse emissions are derived from the fuel supply, including uranium mining, milling, enrichment and fuel manufacture.' Dr Mudd found that financial and environmental costs escalate dramatically as the uranium ore is used. The deeper the mining process required to extract the ore, the higher the cost for mining companies, the greater the impact on the environment and the more resources needed to obtain the product. I t is clear that there is a strong sensitivity of energy and water consumption and greenhouse emissions to ore grade, and that ore grades are likely to continue to decline gradually in the medium to long term. These issues are critical to the current debate over nuclear power and greenhouse emissions, especially with respect to ascribing sustainability to such activities as uranium mining and milling. For example, mining at Roxby Downs is responsible for the emission of over one million tonnes of greenhouse gases per year and this could increase to four million tonnes if the mine is expanded.'

  12. Urea-assisted liquid-phase exfoliation of natural graphite into few-layer graphene

    Science.gov (United States)

    Hou, Dandan; Liu, Qinfu; Wang, Xianshuai; Qiao, Zhichuan; Wu, Yingke; Xu, Bohui; Ding, Shuli

    2018-05-01

    The mass production of graphene with high quality is desirable for its wide applications. Here, we demonstrated a facile method to exfoliate natural graphite into graphene in organic solvent by assisting of urea. The exfoliation of graphite may originate from the "molecular wedge" effect of urea, which can intercalate into the edge of natural graphite, thus facilitating the production of graphene dispersion with a high concentration up to 1.2 mg/mL. The obtained graphene is non-oxidized with negligible defects. Therefore, this approach has great promise in bulk production of graphene with superior quality for a variety of applications.

  13. Geology and exploration of the Rum Jungle Uranium Field

    International Nuclear Information System (INIS)

    Fraser, W.J.

    1980-01-01

    The Rum Jungle Uranium Field was discovered by a private prospector in 1949. A total of 3530 tonnes of uranium oxide was mined and treated from four ore-bodies by Territory Enterprises Pty. Limited who managed the Rum Jungle Project on behalf of the Australian Atomic Energy Commission until the closure of operations in 1971. One small low grade uranium orebody remains to be developed. Lead, zinc, copper, cobalt and nickel were found zoned sub vertically with uranium at one deposit. One medium sized lead, zinc, copper, cobalt and nickel deposit remains to be developed and one small copper deposit with minor uranium was mined. The basemetal deposits show a regional zoning relationship with the known uranium mineralization. Uranium and basemetal mineralization is hosted by graphitic or chloritic, pyritic shales at the contact with a magnesite. These rocks are in the lower part of a sequence of Lower Proterozoic sediments which unconformably overlie Archaean basement complexes. The sediments and complexes are displaced by Giants Reef Fault and sub-parallel shears and linears may further control mineralization. Nearly 50km of the prospective shale/magnesite contact was tested by total count radiometric surveys, various electrical methods, auger, rotary percussion and diamond drilling. The source for the uranium mineralization was probably the Archaean basement complexes from which uranium was initially deposited as protore by either chemical precipitation or clay adsorption in the shale units or as detrital placers in quartz pebble conglomerates immediately overlying the basement complexes. (author)

  14. Utilization of the MCNP-3A code for criticality safety analysis

    International Nuclear Information System (INIS)

    Maragni, M.G.; Moreira, J.M.L.

    1996-01-01

    In the last decade, Brazil started to operate facilities for processing and storing uranium in different forms. The necessity of criticality safety analysis appeared in the design phase of the uranium pilot process plants and also in the licensing of transportation and storage of fissile materials. The 2-MW research reactor and the Angra I power plant also required criticality safety assessments because their spent-fuel storage was approaching full-capacity utilization. The criticality safety analysis in Brazil has been based on KENO IV code calculations, which present some difficulties for correct geometry representation. The MCNP-3A code is not reported to be used frequently for criticality safety analysis in Brazil, but its good geometry representation makes it a possible tool for treating problems of complex geometry. A set of benchmark tests was performed to verify its applicability for criticality safety analysis in Brazil. This paper presents several benchmark tests aimed at selecting a set of options available in the MCNP-3A code that would be adequate for criticality safety analysis. The MCNP-3A code is also compared with the KENO-IV code regarding its performance for criticality safety analysis

  15. The progress in the researches for uranium mill tailings cleaning treatment and no-waste uranium ore milling processes

    International Nuclear Information System (INIS)

    Wang Jintang

    1990-01-01

    The production of uranium mill tailings and their risk assessment are described. The moethods of uranium mill tailings disposal and management are criticized and the necessity of the researches for uranium mill tailings cleaning treatment and no-wasle uranium ore milling process are demonstrated. The progress for these researches in China and other countries with uranium production is reviewed, and the corresponding conclusions are reported

  16. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    Directory of Open Access Journals (Sweden)

    Krása Antonín

    2017-01-01

    Full Text Available VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector. Discrepancies between experiments and Monte Carlo calculations (MCNP5 of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2 are presented.

  17. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    Science.gov (United States)

    Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

    2017-09-01

    VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

  18. Criticality safety evaluations - a open-quotes stalking horseclose quotes for integrated safety assessment

    International Nuclear Information System (INIS)

    Williams, R.A.

    1995-01-01

    The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility's criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE

  19. A Study of Critical Flowrate in the Integral Effect Test Facilities

    International Nuclear Information System (INIS)

    Kim, Yeongsik; Ryu, Sunguk; Cho, Seok; Yi, Sungjae; Park, Hyunsik

    2014-01-01

    In earlier studies, most of the information available in the literature was either for a saturated two-phase flow or a sub-cooled water flow at medium pressure conditions, e. g., up to about 7.0 MPa. The choking is regarded as a condition of maximum possible discharge through a given orifice and/or nozzle exit area. A critical flow rate can be achieved at a choking under the given thermo-hydraulic conditions. The critical flow phenomena were studied extensively in both single-phase and two-phase systems because of its importance in the LOCA analyses of light water reactors and in the design of other engineering areas. Park suggested a modified correlation for predicting the critical flow for sub-cooled water through a nozzle. Recently, Park et al. performed an experimental study on a two-phase critical flow with a noncondensable gas at high pressure conditions. Various experiments of critical flow using sub-cooled water were performed for a modeling of break simulators in thermohydraulic integral effect test facilities for light water reactors, e. g., an advanced power reactor 1400MWe (APR1400) and a system-integrated modular advanced reactor (SMART). For the design of break simulators of SBLOCA scenarios, the aspect ratio (L/D) is considered to be a key parameter to determine the shape of a break simulator. In this paper, an investigation of critical flow phenomena was performed especially on break simulators for LOCA scenarios in the integral effect test facilities of KAERI, such as ATLAS and FESTA. In this study, various studies on the critical flow models for sub-cooled and/or saturated water were reviewed. For a comparison among the models for the selected test data, discussions of the comparisons on the effect of the diameters, predictions of critical flow models, and break simulators for SBLOCA in the integral effect test facilities were presented

  20. A Study of Critical Flowrate in the Integral Effect Test Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeongsik; Ryu, Sunguk; Cho, Seok; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In earlier studies, most of the information available in the literature was either for a saturated two-phase flow or a sub-cooled water flow at medium pressure conditions, e. g., up to about 7.0 MPa. The choking is regarded as a condition of maximum possible discharge through a given orifice and/or nozzle exit area. A critical flow rate can be achieved at a choking under the given thermo-hydraulic conditions. The critical flow phenomena were studied extensively in both single-phase and two-phase systems because of its importance in the LOCA analyses of light water reactors and in the design of other engineering areas. Park suggested a modified correlation for predicting the critical flow for sub-cooled water through a nozzle. Recently, Park et al. performed an experimental study on a two-phase critical flow with a noncondensable gas at high pressure conditions. Various experiments of critical flow using sub-cooled water were performed for a modeling of break simulators in thermohydraulic integral effect test facilities for light water reactors, e. g., an advanced power reactor 1400MWe (APR1400) and a system-integrated modular advanced reactor (SMART). For the design of break simulators of SBLOCA scenarios, the aspect ratio (L/D) is considered to be a key parameter to determine the shape of a break simulator. In this paper, an investigation of critical flow phenomena was performed especially on break simulators for LOCA scenarios in the integral effect test facilities of KAERI, such as ATLAS and FESTA. In this study, various studies on the critical flow models for sub-cooled and/or saturated water were reviewed. For a comparison among the models for the selected test data, discussions of the comparisons on the effect of the diameters, predictions of critical flow models, and break simulators for SBLOCA in the integral effect test facilities were presented.