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Sample records for fuels fbr analytical

  1. Sintered-to-size FBR fuel

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Schaus, P.S.

    1984-04-01

    Fabrication of sintered-to-size PuO 2 -UO 2 fuel pellets was completed for testing of proposed FBR product specifications. Approximately 6000 pellets were fabricated to two nominal diameters and two densities by cold pressing and sintering to size. Process control and correlation between test and production batches are discussed

  2. Part 6. Internationalization and collocation of FBR fuel cycle facilities

    International Nuclear Information System (INIS)

    Stevenson, M.G.; Abramson, P.B.; LeSage, L.G.

    1980-01-01

    This report examines some of the non-proliferation, technical, and institutional aspects of internationalization and/or collocation of major facilities of the Fast Breeder Reactor (FBR) fuel cycle. The national incentives and disincentives for establishment of FBR Fuel Cycle Centers are enumerated. The technical, legal, and administrative considerations in determining the feasibility of FBR Fuel Cycle Centers are addressed by making comparisons with Light Water Reactor (LWR) centers which have been studied in detail by the IAEA and UNSRC

  3. Research on CDA for advanced fuel FBR

    International Nuclear Information System (INIS)

    Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime.

    1997-03-01

    For the purpose of evaluating possibility of the re-criticality of a metallic fueled reactor, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'Research on CDA for advanced fuel FBR'. The results of this year are the following. The accident initiator considered is a loss-of-flow accident with ATWS. The LOF analysis was performed for the metallic fueled 600 MWe homogeneous two region reactors, both for a metallic fuel only and for a metallic fuel core with ZrH pin. The SAS3D CDA initiation phase analysis code was used to investigate the re-criticality potential at the severe accident. The change mainly in the constants was necessary to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. LOF with flow decay half time of t 1/2 =0.5(s) (all blackout case) and 5.5(s) (ordinary LOF case) were analyzed. Independent of the conditions of the analysis, the results show all the cases could avoid to become prompt-critical. Depending on the analysis condition, it becomes necessary to transfer to the transient phase, it is also shown there is a possibility to avoid re-criticality due to the motion of molten fuel both for the metallic fuel and for the metallic fuel with ZrH moderator. However, because of the constants used for the material property the results might overestimate the fuel motion. It is shown that the moderator is effective to terminate the accident at an early stage. The behavior of metallic fueled reactors at CDA was analyzed with SAS3D code by modifying the constants of material properties to be applied to the reactor. It is shown that a metallic fueled reactor has a possibility to avoid re-criticality at CDA. (J.P.N.)

  4. Design study on metal fuel FBR cores

    International Nuclear Information System (INIS)

    Yokoo, T.; Tanaka, Y.; Ogata, T.

    1991-01-01

    A design approach for metal fuel FBR core to maintain fuel integrity during transient events by limiting eutectic/liquid phase formation is proposed based on the current status of metallic fuel development. Its impact as the limitation on the core outlet temperature is assessed through its application to two of CRIEPI's core concepts, high linear power 1000 MWe homogeneous design and medium linear power 300 MWe radially heterogeneous design. SESAME/SALT code is used in this study to analyze steady state and transient fuel behavior. SE2-FA code is developed based on SUPERENERGY-2 and used to analyze core thermal-hydraulics with uncertainties. As the result, the core outlet temperatures of both designs are found to be limited to ≤500degC if it is required to prevent eutectic/liquid phase formation during operational transients in order to guarantee the fuel integrity. Additional assessment is made assuming an advanced limiting condition that allows small liquid phase formation based on the liquid phase penetration rate derived from existing experimental results. The result indicates possibility of raising core outlet temperature to ∼ 530degC. Also, it is found that core design technology improvements such as hot spot factors reduction can contribute to the core outlet temperature extension by 10 ∼ 20degC. (author)

  5. Examination of fast reactor fuels, FBR analytical quality assurance standards and methods, and analytical methods development: irradiation tests. Progress report, April 1--June 30, 1976, and FY 1976

    International Nuclear Information System (INIS)

    Baker, R.D.

    1976-08-01

    Characterization of unirradiated and irradiated LMFBR fuels by analytical chemistry methods will continue, and additional methods will be modified and mechanized for hot cell application. Macro- and microexaminations will be made on fuel and cladding using the shielded electron microprobe, emission spectrograph, radiochemistry, gamma scanner, mass spectrometers, and other analytical facilities. New capabilities will be developed in gamma scanning, analyses to assess spatial distributions of fuel and fission products, mass spectrometric measurements of burnup and fission gas constituents and other chemical analyses. Microstructural analyses of unirradiated and irradiated materials will continue using optical and electron microscopy and autoradiographic and x-ray techniques. Analytical quality assurance standards tasks are designed to assure the quality of the chemical characterizations necessary to evaluate reactor components relative to specifications. Tasks include: (1) the preparation and distribution of calibration materials and quality control samples for use in quality assurance surveillance programs, (2) the development of and the guidance in the use of quality assurance programs for sampling and analysis, (3) the development of improved methods of analysis, and (4) the preparation of continuously updated analytical method manuals. Reliable analytical methods development for the measurement of burnup, oxygen-to-metal (O/M) ratio, and various gases in irradiated fuels is described

  6. Method of planning fuel exchanges in FBR type reactors

    International Nuclear Information System (INIS)

    Urushihara, Hiroshi.

    1979-01-01

    Purpose: To simplify fuel exchange planning and ensure satisfactory fuel burning by simulating nuclear properties required for the fuel exchange in FBR type reactors with multi-variable algebraic expressions in integer linear planning method and conducting the decision for the optimum replacing positions capable of satisfying the aimed functions. Constitution: Number of new fuels to be loaded in each of the regions in the reactor core is determined previously for each of the fuel cycles based on a long term planning, positions for loading succeeding new fuel assemblies are forecast, nuclear properties and number of fuels to be exchanged required for the fuel exchange are respectively simulated with relevant algebraic expressions based on respective nuclear property data such as for neutron flux distribution, burning reactivity and uniform exchange for each of the forecast arrangement and present arrangement, and the loading positions for the new fuels are determined by the use of the integer linear planning method. (Horiuchi, T.)

  7. Description of JNC's analytical method and its performance for FBR cores

    International Nuclear Information System (INIS)

    Ishikawa, M.

    2000-01-01

    The description of JNC's analytical method and its performance for FBR cores includes: an outline of JNC's Analytical System Compared with ERANOS; a standard data base for FBR Nuclear Design in JNC; JUPITER Critical Experiment; details of Analytical Method and Its Effects on JUPITER; performance of JNC Analytical System (effective multiplication factor k eff , control rod worth, and sodium void reactivity); design accuracy of a 600 MWe-class FBR Core. JNC developed a consistent analytical system for FBR core evaluation, based on JENDL library, f-table method, and three dimensional diffusion/transport theory, which includes comprehensive sensitivity tools to improve the prediction accuracy of core parameters. JNC system was verified by analysis of JUPITER critical experiment, and other facilities. Its performance can be judged quite satisfactory for FBR-core design work, though there is room for further improvement, such as more detailed treatment of cross-section resonance regions

  8. Fuel exchanger in FBR type reactor

    International Nuclear Information System (INIS)

    Shinden, Kazuhiko; Tanaka, Osamu.

    1990-01-01

    The present invention concerns a fuel exchanger for exchanging fuels in an LMFBR type reactor using liquid metals as coolants. An outer gripper cylinder rotating device for rotating an outer gripper cylinder that holds a gripper is driven, to lower the gripper driving portion and the outer gripper cylinder, fuels are caught by the finger at the top end of the outer gripper cylinder and elevated to extract the fuels from the reactor core. Then, the gripper driving portion casing and the outer gripper cylinder are rotated to rotate the fuels caught by the gripper. Subsequently, the gripper driving portion and the outer gripper cylinder are lowered to charge the fuels in the reactor core. This can directly shuffle the fuels in the reactor core without once transferring the fuels into a reactor storing pot and replacing with other fuels, thereby shortening the shuffling time. (I.N.)

  9. Fuel exchange device for FBR type reactor

    International Nuclear Information System (INIS)

    Onuki, Koji.

    1993-01-01

    The device of the present invention can provide fresh fuels with a rotational angle aligned with the direction in the reactor core, so that the fresh fuels can be inserted being aligned with apertures of the reactor core even if a self orientation mechanism should fail to operate. That is, a rotational angle detection means (1) detects the rotational angle of fresh fuels before insertion to the reactor core. A fuel rotational angle control means (2) controls the rotational angle of the fresh fuels by comparing the detection result of the means (1) and the data for the insertion position of the reactor core. A fuel rotation means (3) compensates the rotational angel of the fresh fuels based on the control signal from the means (2). In this way, when the fresh fuels are inserted to the reactor core, the fresh fuels set at the same angle as that for the aperture of the reactor core. Accordingly, even if the self orientation mechanism should not operate, the fresh fuels can be inserted smoothly. As a result, it is possible to save loss time upon fuel exchange and mitigate operator's burden during operation. (I.S.)

  10. The strategic lines for developing FBR fuel

    International Nuclear Information System (INIS)

    Dievoet, J. van; Kunsch, P.L.

    1988-01-01

    In several European countries, there is agreement that fast breeder reactors will one day become the major electronuclear power system. This opinion is unaffected by the present slowing down of economic activities in Europe, which is responsible for the decrease in the electricity growth rate. As a result of this slowing down, however, investment decisions on fast reactor programs have been postponed. This transition period should be used to study ways to obtain more economic fast breeder reactors. Cost reduction can be tackled from many different angles. Regarding fuel it seems possible to simultaneously reduce three aspects of generation costs: capital cost, fuel cycle costs and operation/maintenance costs. Both fuel designer and manufacturer have an important part to play in promoting commercial breeders. Many new fuel design ideas are being discussed in the search for a common European model. They must be qualified by extensive R and D programs followed by commercial demonstration reactors. It is of vital interest to the industry that decisions regarding ordering the first demonstration reactor (after Superphenix) should be taken early enough to start these qualification programs. 3 refs., 1 fig

  11. Dissolution of mixed oxide spent fuel from FBR

    International Nuclear Information System (INIS)

    Sanyoshi, H.; Nishina, H.; Toyota, O.; Yamamoto, R.; Nemoto, S.; Okamoto, F.; Togashi, A.; Kawata, T.; Hayashi, S.

    1991-01-01

    At the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation (PNC), the Chemical Processing Facility (CPF) has been continuing operation since 1982 for laboratory scale hot experiments on reprocessing of FBR mixed oxide fuel. As a part of these experiments, dissolution experiments have been performed to define the key parameters affecting dissolution rates such as concentration of nitric acid, temperature and burnup and also to confirm the amount of insoluble residue. The dissolution rate of the irradiated fuel was determined to be in proportion to the 1.7 power of the nitric acid concentration. The activation energy determined from the experiments varied from 6 to 11 kcal/mol depending on the method of dissolution. The dissolution rate decreased as the fuel burnup increased in low nitric acid media below 5 mol/l. However, it was found that the effect of the burnup became negligible in a high concentration of nitric acid media. The amount of insoluble residue and its constituents were evaluated by changing the dissolution condition. (author)

  12. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    International Nuclear Information System (INIS)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-01-01

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  13. Advanced hybrid process with solvent extraction and pyro-chemical process of spent fuel reprocessing for LWR to FBR

    International Nuclear Information System (INIS)

    Fujita, Reiko; Mizuguchi, Koji; Fuse, Kouki; Saso, Michitaka; Utsunomiya, Kazuhiro; Arie, Kazuo

    2008-01-01

    Toshiba has been proposing a new fuel cycle concept of a transition from LWR to FBR. The new fuel cycle concept has better economical process of the LWR spent fuel reprocessing than the present Purex Process and the proliferation resistance for FBR cycle of plutonium with minor actinides after 2040. Toshiba has been developing a new Advanced Hybrid Process with Solvent Extraction and Pyrochemical process of spent fuel reprocessing for LWR to FBR. The Advanced Hybrid Process combines the solvent extraction process of the LWR spent fuel in nitric acid with the recovery of high pure uranium for LWR fuel and the pyro-chemical process in molten salts of impure plutonium recovery with minor actinides for metallic FBR fuel, which is the FBR spent fuel recycle system after FBR age based on the electrorefining process in molten salts since 1988. The new Advanced Hybrid Process enables the decrease of the high-level waste and the secondary waste from the spent fuel reprocessing plants. The R and D costs in the new Advanced Hybrid Process might be reduced because of the mutual Pyro-chemical process in molten salts. This paper describes the new fuel cycle concept of a transition from LWR to FBR and the feasibility of the new Advanced Hybrid Process by fundamental experiments. (author)

  14. Study on the FBR cycle introduction scenario. 4. Evaluation of the FBR cycle introduction scenario from the viewpoints of the fuel cycle requirements

    International Nuclear Information System (INIS)

    Ono, Kiyoshi; Shiotani, Hiroki; Hirao, Kazunori

    2003-07-01

    This report is intended to explain the outline of the scenario studies on FBR (Fast Breeder Reactor) cycle introduction. Recently, people value the reduction of environmental impact in addition to the recycle of energy resources and the energy security in these scenario studies. This report summarizes the analysis about the necessity of plutonium recycling in LWR (Light water Reactor) from short-term view and about the necessity of FBR cycle introduction from a long-term view in Japan, by comparing 'FBR scenario' with 'LWR once-through scenario' and 'Pu recycle in LWR scenario', from the viewpoints of cumulative uranium demand, spent fuel storage, radioactive waste arising, etc. It becomes clear that the plutonium recycling in LWR has a good effect on the reduction of spent fuel storage and the cumulative natural uranium demand before FBR cycle introduction, from short-term view (20-30 years). On the other hand, this analysis also shows that there is much effect of FBR deployment not only on saving amount of uranium use and energy security but also on reduction of high-level radioactive waste (spent fuels and vitrified waste) and minor actinide arising, from long-term view (100-200 years). (author)

  15. Studies in the dissolver off-gas system for a spent FBR fuel reprocessing plant

    International Nuclear Information System (INIS)

    Heinrich, E.; Huefner, R.; Weirich, F.

    1982-01-01

    Investigations of possible modifications of the process steps of a dissolver off-gas (DOG) system for a spent FBR fuel reprocessing plant are reported. The following operations are discussed: iodine removal from the fuel solution; behaviour of NOsub(x) and iodine in nitric acid off-gas scrubbers at different temperatures and nitric acid concentrations; iodine desorption from the scrub acid; selective absorption of noble gases in refrigerant-12; cold traps. The combination of suitable procedures to produce a total DOG system is described. (U.K.)

  16. Study on the nitride fuel fabrication for FBR cycle (1)

    International Nuclear Information System (INIS)

    Shinkai, Yasuo; Ono, Kiyoshi; Tanaka, Kenya

    2002-07-01

    In the phase-II of JNC's 'Feasibility Study on Commercialized Fuel Reactor Cycle System (the F/S)', the nitride fuels are selected as candidate for fuels for heavy metal cooled reactor, gas cooled reactor, and small scale reactor. In particular, the coated fuel particles are a promising concept for gas cooled reactor. In addition, it is necessary to study in detail the application possibility of pellet nitride fuel and vibration compaction nitride fuel for heavy metal cooled reactor and small scale reactor in the phase-II. In 2001, we studied more about additional equipments for the nitride fuel fabrication in processes from gelation to carbothermic reduction in the vibration compaction method. The result of reevaluation of off-gas mass flow around carbothermic reduction equipment in the palletizing method, showed that quantity of off-gas flow reduced and its reduction led the operation cost to decrease. We studied the possibility of fabrication of large size particles in the coated fuel particles for helium gas cooled reactor and we made basic technical issues clear. (author)

  17. Heavy metal inventory and fuel sustainability of recycling TRU in FBR design

    Science.gov (United States)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-01

    Nuclear fuel materials from spent fuel of light water reactors have a potential to be used for destructive devices with very huge energy release or in the same time, it can be utilized as a peaceful energy or civil applications, for generating electricity, desalination of water, medical application and others applications. Several research activities showed some recycled spent fuel can be used as additional fuel loading for increasing fuel breeding capability as well as improving intrinsic aspect of nuclear non-proliferation. The present investigation intends to evaluate the composition of heavy metals inventories and fuel breeding capability in the FBR design based on the loaded fuel of light water reactor (LWR) spent fuel (SF) of 33 GWd/t with 5 years cooling time by adopting depletion code of ORIGEN. Whole core analysis of FBR design is performed by adopting and coupling codes such as SLAROM code, JOINT and CITATION codes. Nuclear data library, JFS-3-J-3.2R which is based on the JENDL 3.2 has been used for nuclear data analysis. JSFR design is the basis design reference which basically adopted 800 days cycle length for 4 batches system. Higher inventories of plutonium of MOX fuel and TRU fuel types at equilibrium composition than initial composition have been shown. Minor actinide (MA) inventory compositions obtain a different inventory trends at equilibrium composition for both fuel types. Higher Inventory of MA is obtained by MOX fuel and less MA inventory for TRU fuel at equilibrium composition than initial composition. Some different MA inventories can be estimated from the different inventory trend of americium (Am). Higher americium inventory for MOX fuel and less americium inventory for TRU fuel at equilibrium condition. Breeding ratio of TRU fuel is relatively higher compared with MOX fuel type. It can be estimated from relatively higher production of Pu-238 (through converted MA) in TRU fuel, and Pu-238 converts through neutron capture to produce Pu-239

  18. Optimization of FBR fuel element for high burnup

    International Nuclear Information System (INIS)

    Marbach, G.; Millet, P.

    1985-03-01

    After a brief historical background showing evolution of the French fast reactor fuel element from RAPSODIE to PHENIX and SUPER PHENIX we have examined the main points which have permitted to increase irradiation performance of the subassembly

  19. Fast reactor core design studies to cope with TRU fuel composition changes in the LWR-to-FBR transition period

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Maruyama, Shuhei; Ohki, Shigeo; Mizuno, Tomoyasu

    2009-01-01

    As part of the Fast Reactor Cycle Technology Development Project (FaCT Project), sodium-cooled fast reactor core design efforts have been made to cope with the TRU fuel composition changes expected during LWR-to-FBR transition period, in which a various kind of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and a way of recycling them. A 750 MWe mixed-oxide fuel core is firstly defined as a FaCT medium-size reference core and its neutronics characteristics are determined. The core is a high internal conversion type and has an average burnup of 150 GWD/T. The reference TRU fuel composition is assumed to come from the FBR equilibrium state. Compared to the LWR-to-FBR transition period, the TRU fuels in the FBR equilibrium period are multi-recycled through fast reactors and have a different composition. An available TRU fuel composition is determined by fast reactor spent fuel multi-recycling scenarios. Then the FaCT core corresponding to the TRU fuel with different compositions is set according to the TRU fuel composition changes in LWR-to-FBR transition period, and the key core neutronics characteristics are assessed. It is shown that among the core neutronics characteristics, the burnup reactivity and the safety parameters such as sodium void reactivity and Doppler coefficient are significantly influenced by the TRU fuel composition changes. As a result, a general characteristic in the FaCT core design to cope with TRU fuel composition changes is grasped and the design envelopes are identified in terms of the burnup reactivity and the safety parameters. (author)

  20. Assemblies and fuel pin behaviour under irradiation in FBR-350

    International Nuclear Information System (INIS)

    Karaulov, V.N.; Blynski, A.P.; Yakovlev, I.L.; Kononova, E.V.

    1998-01-01

    The efficiency of all types of assemblies and fuel pins of the reactor BN-350 is considered in detail. The factors limiting the efficiency are indicated. The behaviour of assemblies with stainless steel ducts is studied. It is shown that the efficiency is restricted in this case by shape and dimensional changes of hexagonal ducts due to radiation swelling and radiation creep of structural materials. The problem of dimensional changes of ducts was solved for cores of reactors BN-350 and BN-600 after testing in BN-350 of experimental assemblies with ducts made of ferritic-martensitic steel 12Cr13Mo2NbVB. For fuel pins of the second type of loading with clad made of stainless steel 0Cr16Ni15Mo3Nb the efficiency is limited by burnup of 13% h.a. and damage dose of 90 dpa. To increase the burnup the core of the BN-350 is supplied by assemblies of modernized type with pins having a large gas plenum and clad made of the stainless steel 0Cr16Ni15Mo2Mn2TiVB that has a good resistance to irradiation. The efficiency of fuel pins of modernized assemblies in the reactor BN-350 core conditions is provided up to 15% h.a. and damage dose 105 dpa. (author)

  1. A basic research on the transient behavior for a metallic fuel FBR

    International Nuclear Information System (INIS)

    Baba, Mamoru; Hirano, Go; Kawada, Ken-ichi; Niwa, Hajime

    1999-03-01

    A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'A basic research on the transient behavior for a metallic fuel FBR'. The results are the following. (1) Target and Results of analysis: The accident initiator considered is a LOF accident without scram. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transition phase. It is shown that the moderator is quite elective to mitigate the accident at the initiation phase. However, it is necessary to analyze the transition phase to know if the re-criticality is totally avoided after the initiation phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large scale material movement in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics, and examined

  2. Recent R/D towards aqueous reprocessing of FBR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Mallika, C.; Pandey, N.K.; Kumar, S.; Kamachi Mudali, U. [Materials, Process and Equipment Development Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2016-07-01

    The mixed Pu-rich carbide spent fuel with a burn up of 155 GWd/t from the Fast Breeder Test Reactor is being reprocessed in a hot-cell facility by PUREX process. Based on the input from the operation of this facility, research and development activities were carried out to improve the recovery, decontamination factors, economy and to reduce the waste volumes. Reduction of uranyl ions in a continuous flow electrochemical reactor and electrolytic as well as chemical reduction of 4 M HNO{sub 3} from liquid waste could be performed in continuous mode. Using the optimized parameters, suitable electrolytic cells/experimental setups were designed for the plant capacity of 6 L/h. Studies on the extraction kinetics of Ru with 30% TBP (tributyl phosphate) in NPH revealed that better decontamination factor with respect to Ru can be achieved using fast contactors like centrifugal extractors (CEs). Towards developing a spent solvent recovery system to reduce organic waste volumes, a pilot plant was set up, which could recover diluent as top product of distillation column and 40% TBP as bottom product from inactive degraded solvent. A solvent recovery system using short path distillation was also developed for installation in hot cells. (authors)

  3. Storage and disposal of high-level radioactive waste from advanced FBR fuel cycle

    International Nuclear Information System (INIS)

    Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

    2011-01-01

    Waste management of fast breeder reactor (FBR) fuel cycle with and without partitioning and transmutation (P and T) technology was investigated by focusing on thermal constraints due to heat deposition from waste in storage and disposal facilities including economics aspects of those facilities. Partitioning of minor actinides (MAs) and heat-generating fission products in high-level waste can enlarge the containment ratio of waste elements in the glass waste forms and shorten predisposal storage period. Though MAs can be transmuted in FBRs or dedicated transmuters, heat-generating fission products are difficult to be transmuted; they are partitioned and stored for a long time before disposal. The disposal concepts for heat-generating fission products and remainders such as rare-earth elements depend on storage period that ranges from several years to several hundreds of years. Short-term storage results in small size of storage facilities and large size of repositories, and vice versa for long-term storage. This trade-off relation was analyzed by estimating repository size as a function of storage period. The result shows that transmutation of MAs is essentially effective to reduce repository size regardless to storage period, and a combination of P and T can provide a smaller repository than the conventional one by two orders of magnitude. The cost analysis for waste management was also made based on rough assumptions on storage, transportation and repository excluding cost for introducing P and T that are still under evaluation. Cost of waste management for FBR without P and T is 0.25 Yen/kWh that is slightly smaller than that for LWR without P and T, 0.30 Yen/kWh. The introduction of MA transmutation to the FBR results in cost of 0.20 Yen/kWh, and full introduction of P and T provides the smallest cost of 0.08 Yen/kWh. (author)

  4. Method of determining the composition of fuels for FBR type reactors

    International Nuclear Information System (INIS)

    Tsutsumi, Kiyoshi.

    1981-01-01

    Purpose: To improve the core safety of FBR type reactors by determining the composition of fuels composed of oxide mixture of plutonium and uranium, using a relation between specific plutonium seed and plutonium enrichment degree. Method: Relation is determined between the ratio of a specific plutonium seed for constituting plutonium oxide, for example 239 U ratio and a plutonium enrichment degree required for setting the assembly power to a constant level. The ratio of 239 U is plutonium having a given isotopic ratio is also determined. The accuracy of the 239 U ratio can be improved by the correction using the density coefficient. Then, the plutonium enrichment degree is determined using the relation determined as above based on the thus determined 239 U ratio. The composition of the fuel using oxide mixture of plutonium and uranium is determined by utilizing the thus obtained plutonium enrichment degree. (Moriyama, K.)

  5. Development of computer code SIMPSEX for simulation of FBR fuel reprocessing flowsheets: II. additional benchmarking results

    International Nuclear Information System (INIS)

    Shekhar Kumar; Koganti, S.B.

    2003-07-01

    Benchmarking and application of a computer code SIMPSEX for high plutonium FBR flowsheets was reported recently in an earlier report (IGC-234). Improvements and recompilation of the code (Version 4.01, March 2003) required re-validation with the existing benchmarks as well as additional benchmark flowsheets. Improvements in the high Pu region (Pu Aq >30 g/L) resulted in better results in the 75% Pu flowsheet benchmark. Below 30 g/L Pu Aq concentration, results were identical to those from the earlier version (SIMPSEX Version 3, code compiled in 1999). In addition, 13 published flowsheets were taken as additional benchmarks. Eleven of these flowsheets have a wide range of feed concentrations and few of them are β-γ active runs with FBR fuels having a wide distribution of burnup and Pu ratios. A published total partitioning flowsheet using externally generated U(IV) was also simulated using SIMPSEX. SIMPSEX predictions were compared with listed predictions from conventional SEPHIS, PUMA, PUNE and PUBG. SIMPSEX results were found to be comparable and better than the result from above listed codes. In addition, recently reported UREX demo results along with AMUSE simulations are also compared with SIMPSEX predictions. Results of the benchmarking SIMPSEX with these 14 benchmark flowsheets are discussed in this report. (author)

  6. Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

    Science.gov (United States)

    Permana, Sidik; Novitrian, Waris, Abdul; Ismail, Suzuki, Mitsutoshi; Saito, Masaki

    2014-09-01

    Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by convertion rasio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissile material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loding scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.

  7. The basic research on the CDA initiation phase for a metallic fuel FBR

    International Nuclear Information System (INIS)

    Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime

    1998-03-01

    A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku University and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled. (1) Target and Results of analysis: The accident initiator considered is a LOF accident with ATWS. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transient phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large material movement to in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics. The results of a sample case, that is a metallic fueled core at the beginning of cycle, show this improvement is appropriate. (3) Conclusion: The behavior at CDA of a metallic fueled core of a fast reactor was analyzed using the CDA initiation phase analysis code and the knowledge of the important characteristics at the CDA initiation phase was obtained

  8. Shielding test of a model for FBR irradiation fuel transport cask

    International Nuclear Information System (INIS)

    Ohashi, A.; Ueki, K.; Iyori, I.; Uruwashi, S.; Iwanaga, S.; Takahashi, S.

    1993-01-01

    Shielding experiments and their Monte Carlo analyses of the half scale model of the PIE cask were carried out to 1. obtain the dose rate distributions around the model and deduce the characteristics of the actual cask dose distribution from the results, 2. examine the propriety of calculational techniques and the accuracy of Monte Carlo codes to be used in the shielding design or the analysis of the actual cask. The following remarks were obtained through the present study. 1. The C/E values were good for almost all the detector locations except for a few particular points. 2. The calculation geometry with scatterable materials around the cask is necessary to derive the details of dose distribution. 3. The solution in the thermal expansion room must be taken into account and added in the calculation geometry. 4. The magnitude of the dose rates of secondary gamma rays is approximately one-fifth of those of the neutron of the half model. Two peaks must be paid attention at 2.2 MeV and 7.6 MeV due to the (n,γ) reaction of hydrogen and iron, respectively. Hence the calculational techniques that were employed in these analyses can be applied to the design or to the safety analysis of the actual cask, which included the computer codes and the nuclear data. In future for the full scale model of the PIE cask, its shielding effect will be calculated by means of the replacement of the source spectrum from 252 Cf to a FBR fuel assembly of post-irradiation-experiment. (J.P.N.)

  9. Method of detecting fuel failure in FBR type reactor and method of estimating fuel failure position

    International Nuclear Information System (INIS)

    Sonoda, Yukio; Tamaoki, Tetsuo

    1989-01-01

    Noise components in a normal state contained in detection signals from delayed neutron monitors disposed to a coolant inlet, etc. of an intermediate heat exchanger are forecast by self-recurring model and eliminated, and resultant detection signals are monitored thereby detecting fuel failure high sensitivity. Subsequently, the reactor is controlled to a low power operation state and a new self-recurring model to the detection signals from the delayed neutron monitors are prepared. Then, noise components in this state are removed and control rods near the delayed neutron monitors are extracted in a short stroke successively to examine the change of response of the delayed neutron monitors. Accordingly, the failed position for each of the fuels can be estimated at a level of one fuel assembly or a level of several assemblies containing the above-mentioned fuel assembly. Since the fuel failure can be detected at a high sensitivity and the position can be estimated, diffusion of abnormality can be prevented and plant shutdown for fuel exchange can be minimized. (I.S.)

  10. Change of fuel-to-cladding gap width with the burn-up in FBR MOX fuel irradiated to high burn-up

    International Nuclear Information System (INIS)

    Maeda, Koji; Asaga, Takeo

    2004-01-01

    In order to study the dependence of the gap width change on the burn-up, the fuel-to-cladding gap widths were investigated by ceramography in a large number of FBR MOX fuel pins irradiated to high burn-up. The dependence of gap widths on the burn-up was closely connected with the formations of JOG (joint oxyde-gaine) and rim structure. The gap widths decreased gradually due to the fuel swelling until ∼30 GWd/t, but beyond this burn-up the dependence showed two different tendencies. With the increase of burn-up, the gap widths decreased due to the increase of fuel swelling in the low fuel temperature region where the rim structure was observed, but they increased in the high fuel temperature region where the JOG rich in Cs and Mo formed in the gap

  11. Investigation on fuel-cladding chemical interaction in metal fuel for FBR

    International Nuclear Information System (INIS)

    Inagaki, Kenta; Nakamura, Kinya; Ogata, Takanari; Uwaba, Tomoyuki

    2013-01-01

    During steady-state irradiation of metallic fuel in fast reactors, rare-earth fission products can react with stainless steel cladding at the fuel-cladding interface. The authors conducted isothermal annealing tests with some diffusion couples to investigate the structure of the wastage layer formed at the interface. Candidate cladding alloys, ferritic-martensitic steel (PNC-FMS) and oxide-dispersion-strengthened (ODS) steel were assembled with rare-earth alloys, RE5 : La-Ce-Pr-Nd-Sm, which simulate the fission yield of rare-earth fission products. The diffusion couples were isothermally annealed in the temperature range of 500-650°C for up to 170 h. In both RE5/ODS-steel and RE5/PNC-FMS couples, the wastage layer of the two-phase region of the (Fe, Cr) 17 RE 2 matrix phase with the precipitation of the (Fe, RE, Cr) phase was formed. The structure was similar to that formed in RE5/Fe-12Cr and RE5/HT9 couples, which implies that the reaction between REs and steel is not significantly influenced by the minor alloying elements within the candidate cladding materials. It was also clarified that the increase in the wastage layer thickness was diffusion-controlled. The temperature dependence of the reaction rate constants were formulated, which can be the basis for the quantification of the wastage layer growth. (author)

  12. Universal high-temperature heat treatment furnace for FBR mixed uranium and plutonium carbide fuel

    International Nuclear Information System (INIS)

    Handa, Muneo; Takahashi, Ichiro; Watanabe, Hitoshi

    1978-10-01

    A universal high-temperature heat treatment furnace for LMFBR advanced fuels was installed in Plutonium Fuel Laboratory, Oarai Research Establishment. Design, construction and performance of the apparatus are described. With the apparatus, heat treatment of the fuel under a controlled gas atmosphere and quenching of the fuel with blowing helium gas are possible. Equipment to measure impurity gas release of the fuel is also provided. Various plutonium enclosure techniques, e.g., a gas line filter with new exchange mechanics, have been developed. In performance test, results of the enclosure techniques are described. (author)

  13. Analysis of an out-of-pile experiment for fuel relocation under CDA condition of FBR

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Ninokata, Hisashi; Shimizu, Akinao

    2004-01-01

    Calculation of one of the SIMBATH experiments was performed using the SIMMER-II code. The experiments were intended to simulate the fuel pin disintegration and the molten materials relocation that can occur during core disruptive accidents assumed in fast breeder reactors. The calculation by SIMMER-II showed that the incorporated step-wise fuel pin disintegration model and the modified particle jamming model were capable of reproducing the course of pin disintegration and materials relocation within the identified effective value ranges of the parameters which govern the blockages formation, i.e., the characteristic radius of solid particles jamming in the flow and the particle viscosity. As a result, the materials redistribution calculated by SIMMER-II well reproduced the experiment. This fact made it possible to interpret the mechanisms of flow blockages formation and related materials redistribution. (author)

  14. Role of analytical chemistry in the development of nuclear fuels

    International Nuclear Information System (INIS)

    Ramakumar, K.L.

    2012-01-01

    Analytical chemistry is indispensable and plays a pivotal role in the entire gamut of nuclear fuel cycle activities starting from ore refining, conversion, nuclear fuel fabrication, reactor operation, nuclear fuel reprocessing to waste management. As the fuel is the most critical component of the reactor where the fissions take place to produce power, extreme care should be taken to qualify the fuel. For example, in nuclear fuel fabrication, depending upon the reactor system, selection of nuclear fuel has to be made. The fuel for thermal reactors is normally uranium oxide either natural or slightly enriched. For research reactors it can be uranium metal or alloy. The fuel for FBR can be metal, alloy, oxide, carbide or nitride. India is planning an advanced heavy water reactor for utilization of vast resources of thorium in the country. Also research is going on to identify suitable metallic/alloy fuels for our future fast reactors and possible use in fast breeder test reactor. Other advanced fuel materials are also being investigated for thermal reactors for realizing increased performance levels. For example, advanced fuels made from UO 2 doped with Cr 2 O 3 and Al 2 O 3 are being suggested in LWR applications. These have shown to facilitate pellet densification during sintering and enlarge the pellet grain size. The chemistry of these materials has to be understood during the preparation to the stringent specification. A number of analytical parameters need to be determined as a part of chemical quality control of nuclear materials. Myriad of analytical techniques starting from the classical to sophisticated instrumentation techniques are available for this purpose. Insatiable urge of the analytical chemist enables to devise and adopt new superior methodologies in terms of reduction in the time of analysis, improvement in the measurement precision and accuracy, simplicity of the technique itself etc. Chemical quality control provides a means to ensure that the

  15. Fabrication of 0.5-inch diameter FBR mixed oxide fuel pellets

    International Nuclear Information System (INIS)

    Rasmussen, D.E.; Benecke, M.W.; McCord, R.B.

    1979-01-01

    Large diameter (0.535 inch) mixed oxide fuel pellets for Fast Breeder Reactor application were successfully fabricated by the cold-press-and-sinter technique. Enriched UO 2 , PuO 2 -UO 2 , and PuO 2 -ThO 2 compositions were fabricated into nominally 90% theoretical density pellets for the UO 2 and PuO 2 -UO 2 compositions, and 88% and 93% T.D. for the PuO 2 -ThO 2 compositions. Some processing adjustments were required to achieve satisfactory pellet quality and density. Furnace heating rate was reduced from 200 to 50 0 C/h for the organic binder burnout cycle for the large, 0.535-inch diameter pellets to eliminate pellet cracking during sintering. Additional preslugging steps and die wall lubrication during pressing were used to eliminate pressing cracks in the PuO 2 -ThO 2 pellets

  16. Application of non-reductive partial partitioning in FBR fuel reprocessing: a simulation study

    International Nuclear Information System (INIS)

    Shekhar Kumar; Koganti, S.B.

    2002-06-01

    The observed performance of conventional partitioning contactor in the Purex process in the Purex process is seldom satisfactory due to over-consumption of reductant and poor U-Pu decontamination factors. Contemporary trends indicate gradually move-over to MOX fuels for FBRs. In this scenario, it is not necessary to separate uranium and plutonium completely. By controlling the acid concentration and flow rates, it is possible to selectively strip essentially all the plutonium and part of the uranium in the aqueous stream. Therefore, a mixed product enriched in plutonium is obtained which can be precipitated and denitrated. Alternatively direct denitration by microwave heating can be used. The idea is particularly attractive for the flowsheet meant for the first core of FBTR where Pu/(U+Pu) ratio of 0.7 (in the discharged fuel) is diluted by addition of uranium to 0.3. By partial partitioning in the 2B contactor, a product enriched in plutonium (Pu/(U+Pu) ratio ∼0.6) can be obtained. After a minor uranium addition, this product will be suitable for the direct fabrication of II core of FBTR where a Pu/(U+Pu) ratio of 0.55 will be required. In a similar fashion, the enriched product can be used for multiple core zones of proposed PFBR by selective additions of uranium for each zone. To explore the feasibility of such partial partitioning step, an exhaustive simulation study was made using the in-house developed computer code SIMPSEX. 1300 simulations runs were completed for different combinations of parameters and results were analyzed. In this report, the results of this study and a possible flowsheet step have been discussed. Simultaneous variation in the flow rates has been considered and safe operating limits for the partial partitioning step have been established. (author)

  17. Improvement of the computing speed of the FBR fuel pin bundle deformation analysis code 'BAMBOO'

    International Nuclear Information System (INIS)

    Ito, Masahiro; Uwaba, Tomoyuki

    2005-04-01

    JNC has developed a coupled analysis system of a fuel pin bundle deformation analysis code 'BAMBOO' and a thermal hydraulics analysis code ASFRE-IV' for the purpose of evaluating the integrity of a subassembly under the BDI condition. This coupled analysis took much computation time because it needs convergent calculations to obtain numerically stationary solutions for thermal and mechanical behaviors. We improved the computation time of the BAMBOO code analysis to make the coupled analysis practicable. 'BAMBOO' is a FEM code and as such its matrix calculations consume large memory area to temporarily stores intermediate results in the solution of simultaneous linear equations. The code used the Hard Disk Drive (HDD) for the virtual memory area to save Random Access Memory (RAM) of the computer. However, the use of the HDD increased the computation time because Input/Output (I/O) processing with the HDD took much time in data accesses. We improved the code in order that it could conduct I/O processing only with the RAM in matrix calculations and run with in high-performance computers. This improvement considerably increased the CPU occupation rate during the simulation and reduced the total simulation time of the BAMBOO code to about one-seventh of that before the improvement. (author)

  18. Development of FBR technology in the FBR 'Joyo'

    International Nuclear Information System (INIS)

    Nara, Yoshihiko; Akiyama, Takao; Sato, Isao; Mizoo, Nobutatsu; Yoshimi, Hirotaka; Shimada, Takashi

    1986-01-01

    Power Reactor and Nuclear Fuel Development Corp. has advanced the construction of the prototype FBR ''Monju'', and the ground breaking ceremony was held on October 28, 1985. For the design and construction of Monju, the experience, achievement, and the results of development by the own effort and international cooperation gained by the experimental FBR ''Joyo'' have been reflected. It is important to develop the core management technology, operation-supporting system, the techniques of regular inspection, maintenance and repair, the reduction of radiation exposure and so on, to accumulate the experience, and to reflect those accurately to Monju. The operation history of the experimental FBR ''Joyo'', the international joint research on FBRs using the Joyo, the results regarding the characteristic technology of FBRs such as the reactor core, fuel and control rods, sodium technology, the construction of machinery and equipment, and the plant system the plan of developing the high grade technology of FBRs such as the development of fuel and materials, the improvement of reliability and the development of operation management techniques, the verifying test of new technology such as spent fuel storage, the new system for sodium purification and the techniques for analyzing earthquake response, and the international cooperation are reported. (Kako, I.)

  19. FBR Plant Engineering Center annual report 2012

    International Nuclear Information System (INIS)

    2013-12-01

    This annual report shows the last year's R and D activities of currently-reorganized FBR Plant Engineering Center, which was established on April 1, 2009. FBR Safety Technology Center was founded on April 1, 2013 by the consolidation of both the activities of 'former FBR Plant Engineering Center' and a portion of 'FBR Safety Evaluation Unit, Advanced Nuclear System Research and Development Directorate', especially concentrating on safety evaluations and analyses for severe accidents. As for FBR safety technology, it is necessary to continuously make an effort for compliance with new safety regulations in preparation for 'Monju' to restart, for safety enhancement evaluation and for safety technology upgrading. In this context, the new organization was founded in order to reinforce the safety evaluation capability, which will surely and steadily promote FBR safety-technology related activities. As a result, FBR Plant Engineering Center was abolished. This report summarizes the R and D activities at the former FBR Plant Engineering Center, aiming at contributing to the commercialization by using operation experiences and technology development results derived from the actual reactor 'Monju'. The activities are divided into five areas of operation-and-maintenance engineering, sodium engineering, reactor-core-and-fuel engineering, plant engineering, and safety engineering. This annual report is intended for a report of the activities of individual researcher in the center rather than that of the progress of the center as a whole. This will clarify the individual themes, progresses and problems of each researcher, which will, hopefully, facilitate communication with the outside researchers. (author)

  20. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  1. Preparation of a thermal-hydraulic design method for driver core fuel pins of a new in-pile experimental reactor for FBR safety research

    International Nuclear Information System (INIS)

    Mizuno, Masahiro; Yamaguchi, Katsuhisa; Uto, Nariaki

    1999-07-01

    A design study of a new in-pile experimental reactor, SERAPH (Safety Engineering Reactor for Accident PHenomenology), for FBR safety research has progressed at JNC (Japan Nuclear Cycle Development Institute). SERAPH is intended for various in-pile experiments to be performed under quasi-steady state and various transient operation modes. In order to evaluate the driver core performance in conducting such experiments, clarify the relating design issues to be resolved and refine the experimental needs, it is indispensable to comprehend the allowable margin for the thermal-hydraulic fuel pin design since it largely affects the strategy for the driver core design. This report presents a thermal-hydraulic design method for the driver core fuel pins, which is a combination of a two-dimensional time-dependent heat transfer analysis code TAC-2D and a general non-linear finite-element structural analysis code FINAS. In TAC-2D, the allowable spatial mesh and the time step sizes are evaluated. The code is modified so as to treat time-dependent thermal properties, include an improved gap heat-transfer model and treat the change of intra-pin gap width under transient modes, for the purpose of improving the accuracy of evaluating heat transfer characteristics which gives a significant impact on the thermal-hydraulic design. As for FINAS, the number of element nodes and spatial meshes required to obtain adequate accuracy for the thermal stress characteristics of a fuel pellet during transient modes are investigated. In addition, post-processing tools are newly developed to process the calculation results obtained from these codes. The results of this work contribute to advancing the fuel pin design study for SERAPH as well with the investigation on the technique of manufacturing fuel pins. (author)

  2. TRU composition changes and their influence on FBR core characteristics in the LWR-to-FBR transition period

    International Nuclear Information System (INIS)

    Maruyama, Shuhei; Ohki, Shigeo; Mizuno, Tomoyasu

    2009-01-01

    In the conceptual core and fuel design studies in Fast Reactor Cycle Technology Development Project (FaCT Project) in Japan, much interest has been taken in the fuel nuclide compositions for a transition period from light water reactor to fast breeder reactor (FBR). In this paper, the range of transuranic (TRU) nuclide composition to be provided to FBR is evaluated with extended recycling scenario calculations. The influence of TRU composition changes on FBR core characteristics are also discussed with explanations of major contributing factors. (author)

  3. Challenges in development of matrices for vitrification of old legacy waste and high-level radioactive waste generated from reprocessing of AHWR and FBR spent fuel

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2012-01-01

    Majority of radioactivity in entire nuclear fuel cycle is concentrated in HLW. A three step strategy for management of HLW has been adopted in India. This involves immobilization of waste oxides in stable and inert solid matrices, interim retrievable storage of the conditioned waste product under continuous cooling and disposal in deep geological formations. Glass has been accepted as most suitable matrix world-wide for immobilization of HLW, because of its attractive features like ability to accommodate wide range of waste constituents, modest processing temperatures, adequate chemical, thermal and radiation stability. Borosilicate glass matrix developed by BARC in collaboration with CGCRI has been adopted in India for immobilization of HLW. In view of compositional variation of HLW from site to site, tailor make changes in the glass formulations are often necessary to incorporate all the waste constituents and having the product of desirable characteristics. The vitrified waste products made with different glass formulations and simulated waste need to be characterized for chemical durability, thermal stability, homogeneity etc. before finalizing a suitable glass formulation. The present extended abstract summarises the studies carried out for development of glass formulations for vitrification of legacy waste and futuristic waste likely to be generated from AHWR and FBR having wide variations in their compositions. The presently stored HLW at Trombay is characterized by significant concentrations of uranium, sodium and sulphate in addition to fission products, corrosion products and small amount of other actinides

  4. Cost-benefit analysis on FBR cycle R and D for the world

    International Nuclear Information System (INIS)

    Kawasaki, Hirotsugu

    2006-01-01

    This analysis was estimated on the assumption that the nuclear power generation will be changed by FBR and both LWR and FBR indicate same nuclear power generation cost and the environmental load. The cost-benefit analysis results on FBR cycle R and D in the world showed that increase of power generation cost with increase of uranium fuel cost will be avoided and decrease of power generation cost by introducing FBR. The cost-benefit analysis results on FBR cycle R and D in Japan showed that about 9 billions yen will be obtained by the above two economic effects. Cost-benefit effects by introducing FBR, economic estimation method of cost-benefit effect, range and contents of cost-benefit effect on FBR R and D, preconditions of evaluation, and evaluation results are explained. (S.Y.)

  5. Feedback on FBR fuel fabrication at ATPu facility, as a support to the design of a future facility

    International Nuclear Information System (INIS)

    Geneves, T.; Favet, D.; Audubert, F.; Paret, L.

    2013-01-01

    Conclusions: → Within 40 years, AREVA and CEA gained a large experience in the domain of Fast Breeder fuel fabrication: – Standard and experimental fuels from different designs; – Throughput from 500 kgHM/year to 18 tHM/year; – Technological improvements all along the 120tHM manufactured. → MELOX plant operates industrial methods and technologies since 17 years: – 2000 tHM LWR fuels produced; → The basic design of the workshop for ASTRID fuel fabrication is based on this large experience

  6. Study on coated layer material performance of coated particle fuel FBR (2). High temperature property and capability of coating to thick layer of TiN

    International Nuclear Information System (INIS)

    Naganuma, Masayuki; Mizuno, Tomoyasu

    2002-08-01

    'Helium Gas Cooled Coated Particle Fuel FBR' is one of attractive core concepts in the Feasibility Study on Commercialized Fast Reactor Cycle System in Japan, and the design study is presently proceeded. As one of key technologies of this concept, the coated layer material is important, and ceramics is considered to be a candidate material because of the superior refractory. Based on existing knowledge, TiN is regarded to be a possible candidate material, to which some property tests and evaluations have been conducted. In this study, preliminary tests about the high temperature property and the capability of thick layer coating of TiN have been conducted. Results of these tests come to the following conclusions. Heating tests of two kinds of TiN layer specimens coated by PVD (Physical Vapor Deposition) and CVD (Chemical Vapor Deposition) were conducted. As a result, as for CVD coating specimens, remarkable charge was not observed on the layer up to 2,000degC, therefore we concluded that the layer by CVD had applicability up to high temperature of actual operation level. On the other hand, as for PVD coating specimens, an unstable behavior that the layer changed to a mesh like texture was observed on a 2,000degC heated specimen, therefore the applied PVD method is not considered to be promising as the coating technique. The surface conditions of some parts inside CVD device were investigated in order to evaluate possibility of TiN thick coating (∼100 μm). As a result, around 500 μm of TiN coating layer was observed on the condition of multilayer. Therefore, we conclude that CVD has capability of coating up to thick layer in actual coated particle fuel fabrication. (author)

  7. Sodium boiling and mixed oxide fuel thermal behavior in FBR undercooling transients; W-1 SLSF experiment results

    International Nuclear Information System (INIS)

    Henderson, J.M.; Wood, S.A.; Knight, D.D.

    1981-01-01

    The W-1 Sodium Loop Safety Facility (SLSF) Experiment was conducted to study fuel pin heat release characteristics during a series of LMFBR Loss-of-Piping Integrity (LOPI) transients and to investigate a regime of coolant boiling during a second series of transients at low, medium and high bundle power levels. The LOPI transients produced no coolant boiling and showed only small changes in coolant temperatures as the test fuel microstructure changed from a fresh, unrestructured to a low burnup, restructured condition. During the last of seven boiling transients, intense coolant boiling produced inlet flow reversal, cladding dryout and moderate cladding melting

  8. FBR/VHTR deployment scenarios in Japan

    International Nuclear Information System (INIS)

    Richards, Matt; Kunitomi, Kazuhiko

    2008-01-01

    Co-deployment of Fast Breeder Reactors (FBRs) and Very High Temperature Reactors (VHTRs) can be used as the nuclear technologies to meet a significant portion of Japan's future energy demands. The FBR provides the fissile fuel for energy security and sustainability, and can be used to provide a significant portion of the electricity demand. The VHTR can provide flexible energy outputs (electricity, hydrogen, and high-temperature heat) with high efficiency, can operate with a wide variety of fuel cycles, and can be sited at locations that have limited availability of cooling water. These features, combined with its passive safety and high degree of proliferation resistance, make the VHTR an ideal complement for co-deployment with the FBR in Japan and also a very low-risk technology of export to foreign countries. In addition to hydrogen production, the high-temperature thermal energy produced by the VHTR fleet can be used for a wide variety of process-heat applications, and the VHTR can play a key role for significantly reducing greenhouse-gas emissions. This paper describes assessments for deploying FBRs and VHTRs in Japan using a closed fuel cycle, with the FBRs supplying the fissile material to sustain the combined FBR/VHTR fleet. (author)

  9. The Swiss contribution to the FBR development

    International Nuclear Information System (INIS)

    Hudina, M.

    1981-01-01

    The program of fast breeder reactors development in Switzerland is considered from two points of view: energy self-sufficiency and optimization of fuel cycle. Research and development program covers: safety features of LMFBRs, development of mixed carbide fuel elements, study of steam generators transient behaviour, influence of various cooling concepts on thermal efficiency, techniques for detecting cover gas bubbles in the primary sodium circuit. This paper includes cost od the research and development activities as well as description of the future aims of the FBR projects

  10. Method of operating FBR type reactors

    International Nuclear Information System (INIS)

    Arie, Kazuo.

    1984-01-01

    Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

  11. Investigation on fuel-cladding chemical interaction in metal fuel for FBR. Reaction of rare earth elements with Fe-Cr alloy

    International Nuclear Information System (INIS)

    Inagaki, Kenta; Ogata, Takanari

    2010-01-01

    Rare-earth fission product (FP) elements generated in the metal fuel interact with cladding alloy and result in the wastage of the cladding (Fuel-Cladding Chemical Interaction (FCCI)). To evaluate FCCI quantitatively, several influential factors must be considered. They are temperature, temperature gradient, time, composition of the cladding and the behavior of rare-earth FP. In this research, the temperature and time dependencies are investigated with tests in the simplified system. Fe-12wt%Cr was used as stimulant material of cladding and rare-earth alloy 13La -24Ce -12Pr -39Nd -12Sm (RE) as a rare-earth FP. A diffusion couple Fe-Cr/RE was made and annealed at 923K, 853K, 773K or 693K. The structures of reaction layers were analyzed with Electron Probe Micro Analyzer (EPMA) and the details of the structures were clarified. The width of the reaction layer in the Fe-Cr alloy grew in proportion to the square root of time. The reaction rate constants K=(square of the width of reaction layer / time) were evaluated. It was confirmed that the relation between K and the inverse of the temperature showed linearity above 773 K. (author)

  12. Structural dynamics in FBR

    International Nuclear Information System (INIS)

    Bhoje, S.B.

    2003-01-01

    In view of thin walled large diameter shell structures with associated fluid effects, structural dynamics problems are very critical in a fast breeder reactor. Structural characteristics and consequent structural dynamics problems in typical pool type Fast Breeder Reactor are highlighted. A few important structural dynamics problems are pump induced as well as flow induced vibrations, seismic excitations, pressure transients in the intermediate heat exchangers and pipings due to a large sodium water reaction in the steam generator, and core disruptive accident loadings. The vibration problems which call for identification of excitation forces, formulation of special governing equations and detailed analysis with fluid structure interaction and sloshing effects, particularly for the components such as PSP, inner vessel, CP, CSRDM and TB are elaborated. Seismic design issues are presented in a comprehensive way. Other transient loadings which are specific to FBR, resulting from sodium-water reaction and core disruptive accident are highlighted. A few important results of theoretical as well as experimental works carried out for 500 MWe Prototype Fast Breeder Reactor (PFBR), in the domain of structural dynamics are presented. (author)

  13. Evaluation of the commercial FBR introduction date

    International Nuclear Information System (INIS)

    White, M.K.; Merrill, E.T.

    1981-09-01

    This report examines one criterion for introducing a commercial FBR: economic competitiveness with a Light Water Reactor (LWR). For this analysis, the commercial FBR is assumed to be the fifth-of-a kind replicate which represents an economically mature plant. This FBR is deemed economically competitive when its life-cycle energy cost is less than or equal to that of an LWR. Results of this analysis are used in a comparative analysis of alternative FBR development stategies. The strategies evaluated in these studies assume both 1000- and 1457-MWe FBRs. Since the capital costs per kilowatt, and therefore the energy costs, for these two FBR sizes are different, they will become economically competitive at different times. The probability density function for the 1457-MW(e) FBR has an expected value date or weighted average date of 2030, compared with 2033 for the probability density function for the 1000-MW(e) FBR

  14. FBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Azekura, Kazuo; Inoue, Kotaro.

    1981-01-01

    Purpose: To decrease power fluctuations due to burning of blanket fuel element clusters by partially replacing the fertile materials in the blanket fuel element clusters with fissile materials. Constitution: Fertile materials in the radial blanket fuel element clusters disposed to the outside or inside of the reactor core are partially replaced with fissile materials. Since the power density of the fissile materials is at the maximum in the initial burning stage and decreases as the burning proceeds, the power density of the materials which is smaller in the initial burning stage and becomes greater with the burning by the neutron-accumulated plutonium is offset. Accordingly, the power fluctuations in the blanket fuel element clusters due to the burning made smaller thereby enable to form a reactor core with less power fluctuations due to burning under the constant coolant flow rate depending on the power in the final burning stage where the blanket power is maximum. (Moriyama, K.)

  15. Analytical methodology and facility description spent fuel policy

    International Nuclear Information System (INIS)

    1978-08-01

    Three generic environmental impact statements (GEISs) on domestic fuels, foreign fuels, and storage charges are being prepared to provide environmental input into decisions on whether, and if so how the 1977 Presidential policy on spent fuel storage should be implmented. This report provides background information for two of these environmental impact statements: Storage of U.S. Spent Power Reactor Fuel and Storage of Foreign Spent Power Reactor Fuel. It includes the analytical methodology used in GEISs to assess the environmental effects and a description of the facilities used in the two GEISs

  16. Analytical methodology and facility description spent fuel policy

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-01

    Three generic environmental impact statements (GEISs) on domestic fuels, foreign fuels, and storage charges are being prepared to provide environmental input into decisions on whether, and if so how the 1977 Presidential policy on spent fuel storage should be implmented. This report provides background information for two of these environmental impact statements: Storage of U.S. Spent Power Reactor Fuel and Storage of Foreign Spent Power Reactor Fuel. It includes the analytical methodology used in GEISs to assess the environmental effects and a description of the facilities used in the two GEISs.

  17. Analytic models for fuel pin transient performance

    International Nuclear Information System (INIS)

    Bard, F.E.; Fox, G.L.; Washburn, D.F.; Hanson, J.E.

    1976-09-01

    HEDL's ability to analyze various mechanisms that operate within a fuel pin has progressed substantially through development of codes such as PECTCLAD, which solves cladding response, and DSTRESS, which solves fuel response. The PECTCLAD results show good correlation with a variety of mechanical tests on cladding material and also demonstrate the significance of cladding strength when applying the life fraction rule. The DSTRESS results have shown that fuel deforms sufficiently during overpower transient tests that available volumes are filled, whether in the form of a central cavity or start-up cracks

  18. Method of controlling power distribution in FBR type reactors

    International Nuclear Information System (INIS)

    Sawada, Shusaku; Kaneto, Kunikazu.

    1982-01-01

    Purpose: To attain the power distribution flattening with ease by obtaining a radial power distribution substantially in a constant configuration not depending on the burn-up cycle. Method: As the fuel burning proceeds, the radial power distribution is effected by the accumulation of fission products in the inner blancket fuel assemblies which varies the effect thereof as the neutron absorbing substances. Taking notice of the above fact, the power distribution is controlled in a heterogeneous FBR type reactor by varying the core residence period of the inner blancket assemblies in accordance with the charging density of the inner blancket assemblies in the reactor core. (Kawakami, Y.)

  19. Analytical study of stress and deformation of HTR fuel blocks

    International Nuclear Information System (INIS)

    Tanaka, M.

    1982-01-01

    A two-dimensional finite element computer code named HANS-GR has been developed to predict the mechanical behavior of the graphite fuel blocks with realistic material properties and core environment. When graphite material is exposed to high temperature and fast neutron flux of high density, strains arise due to thermal expansion, irradiation-induced shrinkage and creep. Thus stresses and distortions are induced in the fuel block in which there are spatial variation of these strains. The analytical method used in the program to predcit these induced stresses and distortions by finite element method is discussed. In order to illustrate the versatility of the computer code, numerical results of two example analyses of the multi-hole type fuel elements in the VHTR Reactor are given. Two example analyses presented are those concerning the stresses in fuel blocks with control rod holes and distortions of the fuel blocks at the periphery of the reactor core. It is considered these phenomena should be carefully examined when the multi-hole type fuel elements are applied to VHTR. It is assured that the predicted mechanical behavior of the graphite components is strongly dependent on the material properties used and obtaining the reliable material property is important to make the analytical prediction a reliable one

  20. The Cs2MoO4 / Na reaction: application to the fuel / sodium interaction during a high burnup clad failure in a FBR

    International Nuclear Information System (INIS)

    Tete, F.

    1999-01-01

    The use of sodium as coolant in fast neutron reactors (FBRs) has led to analyze the effects of an accidental input of sodium inside a fuel rod with a tightness defect. This work presents the study of the fuel/sodium interaction, in particular in high burnup conditions, and the study of the equilibrium of the seal/oxide/clad (SOC) / Na system using a simplified model: the Cs 2 MoO 4 / Na reaction, cesium molybdate being the main SOC constituent. In a first step, a bibliographic study about clad failure modeling is presented. It shows the main phenomena controlling the behaviour of a broken low burnup rod and makes a status of the present day knowledge about the consequences of Na penetration in the SOC parts, at high burnup. Then an experimental, thermodynamical and structural study is carried out on the Cs 2 MoO 4 compound in order to determine its physico-chemical properties and its behaviour without the presence of Na. The fundamental study of the compatibility of the Cs 2 MoO 4 / Na system is performed, first theoretically using a classical thermodynamic approach, and then experimentally using the differential thermal analysis and the scanning electron microscopy. The hypothesis of Cs substitution by Na, commonly accepted in the past, is rejected and replaced by the formation of cesium, sodium and molybdenum oxides as indicated by the thermodynamical results (reaction temperature and enthalpy variation). Finally, the study of the Cs 2 MoO 4 / Na system in the concrete case of the nuclear fuel is performed by the direct observation of the SOC / Na reaction zone on a rod broken in experimental reactor and using the Sage thermochemical calculation software. This work proposes, first, some new thermodynamical information, and second, confirms the results previously obtained and gives some elements about the behaviour of other fission products with respect to sodium, like the formation of sodium telluride Na 2 Te and sodium iodides NaI and Na 2 I 2 . (J.S.)

  1. Summary: analysis of alternative FBR development strategies

    International Nuclear Information System (INIS)

    Burnham, J.B.

    1981-12-01

    This report summarizes the comparative evaluation of alternative strategies for the development of the commercial fast breeder reactor (FBR) in the United States. For planning purposes, a range of possible FBR development paths called strategies were selected for evaluation. These strategies, designed to be technically and economically feasible, were expressed in terms of the timing and nature of facilities/research and development programs required to reach full power operation of the first commercial FBR. Four of the seven strategies resulted in a large (1457 MWe) FBR as an end point, the other three in a 1000-MWe plant. Probability distributions were calculated for total strategy costs and time to completion. For the seven strategies analyzed, the costs (discounted 1980 dollars) ranged from $1.8 billion to $4.9 billion; the completion times ranged from 24 to 55 years

  2. Study on commercial FBR concepts by combining innovative technologies

    International Nuclear Information System (INIS)

    Miura, M.; Inagaki, T.; Kuroha, M.; Hida, T.

    1992-01-01

    A study was conducted on future prospects of FBR commercialization. Targets of further improving safety and economy were set to make commercial power plants that would be superior to future LWRs. Promising innovative technologies studied domestically and overseas were extracted by evaluating prospects for commercialization, effect, and plant applicability. Several commercial plants were conceptualized by introducing such technology to large-scale and oxide-fuel reactors. Estimates of construction cost, etc., proved that the targets could be achieved. A concept of long-term technological development was synthesized. (author)

  3. Creep fatigue design of FBR components

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chellapandi, P.

    1997-01-01

    This paper deals with the characteristic features of Fast Breeder Reactor (FBR) with reference to creep fatigue, current creep fatigue design approach in compliance with RCCMR (1987) design code, material data, effects of weldments and neutron irradiation, material constitutive models employed, structural analysis and further R and D required for achieving maturity in creep fatigue design of FBR components. For the analysis reported in this paper, material constitutive models developed based on ORNIb (Oak Ridge National Laboratory) and Chaboche viscoplastic theories are employed to demonstrate the potential of FBR components for higher plant temperatures and/or longer life. The results are presented for the studies carried out towards life prediction of Prototype Fast Breeder Reactor (PFBR) components. (author). 24 refs, 8 figs, 5 tabs

  4. Analytical Dimensional Reduction of a Fuel Optimal Powered Descent Subproblem

    Science.gov (United States)

    Rea, Jeremy R.; Bishop, Robert H.

    2010-01-01

    Current renewed interest in exploration of the moon, Mars, and other planetary objects is driving technology development in many fields of space system design. In particular, there is a desire to land both robotic and human missions on the moon and elsewhere. The landing guidance system must be able to deliver the vehicle to a desired soft landing while meeting several constraints necessary for the safety of the vehicle. Due to performance limitations of current launch vehicles, it is desired to minimize the amount of fuel used. In addition, the landing site may change in real-time in order to avoid previously undetected hazards which become apparent during the landing maneuver. This complicated maneuver can be broken into simpler subproblems that bound the full problem. One such subproblem is to find a minimum-fuel landing solution that meets constraints on the initial state, final state, and bounded thrust acceleration magnitude. With the assumptions of constant gravity and negligible atmosphere, the form of the optimal steering law is known, and the equations of motion can be integrated analytically, resulting in a system of five equations in five unknowns. It is shown that this system of equations can be reduced analytically to two equations in two unknowns. With an additional assumption of constant thrust acceleration magnitude, this system can be reduced further to one equation in one unknown. It is shown that these unknowns can be bounded analytically. An algorithm is developed to quickly and reliably solve the resulting one-dimensional bounded search, and it is used as a real-time guidance applied to a lunar landing test case.

  5. Characterization of alternative FBR development strategies

    International Nuclear Information System (INIS)

    Boegel, A.J.; Clausen, M.J.

    1981-08-01

    Near-term decisions regarding the nature and place of the FBR development program must be made. This study is part of a larger program designed to provide the Department of Energy (DOE) with imformation that can be used to make strategic programmatic decisions. The focus of this report is the description of alternative approaches for developing the FBR and the quantification of the duration and cost of each alternative. The time frames of the alternative approaches are investigated in companion reports (White 1981 and Fraley 1981). The results of these analyses will be described in a summary report

  6. Status of feasibility study for various technical options of FBR systems

    International Nuclear Information System (INIS)

    Kani, Yoshio

    2000-01-01

    JNC (Japan Nuclear Cycle Development Institute) has started a new research project of feasibility studies (F/S) for a wide variety option of fast breeder reactor (FBR) and related fuel cycle in order to develop an economically competitive FBR cycle system fro commercialization. JNC and the electric untilities in Japan have established a new organization in JNC to perform the F/S since July 1, 1999. The organization has undertaken feasibility studies (F/S) in order to determine promising FBR cycle concepts and define necessary RandD tasks. The long-term targets of commercialized FBR cycle system are set as ensuring safety, economic competitiveness relative to future LWRs, efficient utilization of resources, reduction in environmental burden, and enhancement of nuclear non-proliferation. This paper describes the progress of design studies for a wide variety of technical options of FBR plants in the framework of the F/S. We make efforts towards considering all key issues so as not to fail to notice the best concept in a commercialized stage. In the study of technical options, the identified coolant types are sodium, heavy metal (lead and lead-bismuth), gas (carbon dioxide and helium ) and water (boiling water, pressurized water and supercritical water). The classified types of fuel are mixed oxide, nitride and metal. Design studies of small size modular plant concepts are also performed. We study many reactor concepts in combination with a coolant type and a fuel type, understand characteristics of each reactor concept based on our experience and an extensive survey of literature, and make a draft design of each reactor concept for rough estimation of construction costs. We also check how far the concept accomplishes each index (safety, economy, resource utilization, etc.) of design requirements, and will select several promising reactor concepts. (author)

  7. Fuel Cycle Externalities: Analytical Methods and Issues, Report 2

    International Nuclear Information System (INIS)

    Barnthouse, L.W.; Cada, G.F.; Cheng, M.-D.; Easterly, C.E.; Kroodsma, R.L.; Lee, R.; Shriner, D.S.; Tolbert, V.R.; Turner, R.S.

    1994-01-01

    that also have not been fully addressed. This document contains two types of papers that seek to fill part of this void. Some of the papers describe analytical methods that can be applied to one of the five steps of the damage function approach. The other papers discuss some of the complex issues that arise in trying to estimate externalities. This report, the second in a series of eight reports, is part of a joint study by the U.S. Department of Energy (DOE) and the Commission of the European Communities (EC)* on the externalities of fuel cycles. Most of the papers in this report were originally written as working papers during the initial phases of this study. The papers provide descriptions of the (non-radiological) atmospheric dispersion modeling that the study uses; reviews much of the relevant literature on ecological and health effects, and on the economic valuation of those impacts; contains several papers on some of the more complex and contentious issues in estimating externalities; and describes a method for depicting the quality of scientific information that a study uses. The analytical methods and issues that this report discusses generally pertain to more than one of the fuel cycles, though not necessarily to all of them. The report is divided into six parts, each one focusing on a different subject area

  8. Estimating Fuel Cycle Externalities: Analytical Methods and Issues, Report 2

    Energy Technology Data Exchange (ETDEWEB)

    Barnthouse, L.W.; Cada, G.F.; Cheng, M.-D.; Easterly, C.E.; Kroodsma, R.L.; Lee, R.; Shriner, D.S.; Tolbert, V.R.; Turner, R.S.

    1994-07-01

    of complex issues that also have not been fully addressed. This document contains two types of papers that seek to fill part of this void. Some of the papers describe analytical methods that can be applied to one of the five steps of the damage function approach. The other papers discuss some of the complex issues that arise in trying to estimate externalities. This report, the second in a series of eight reports, is part of a joint study by the U.S. Department of Energy (DOE) and the Commission of the European Communities (EC)* on the externalities of fuel cycles. Most of the papers in this report were originally written as working papers during the initial phases of this study. The papers provide descriptions of the (non-radiological) atmospheric dispersion modeling that the study uses; reviews much of the relevant literature on ecological and health effects, and on the economic valuation of those impacts; contains several papers on some of the more complex and contentious issues in estimating externalities; and describes a method for depicting the quality of scientific information that a study uses. The analytical methods and issues that this report discusses generally pertain to more than one of the fuel cycles, though not necessarily to all of them. The report is divided into six parts, each one focusing on a different subject area.

  9. Laser-based analytical monitoring in nuclear-fuel processing plants

    Energy Technology Data Exchange (ETDEWEB)

    Hohimer, J.P.

    1978-09-01

    The use of laser-based analytical methods in nuclear-fuel processing plants is considered. The species and locations for accountability, process control, and effluent control measurements in the Coprocessing, Thorex, and reference Purex fuel processing operations are identified and the conventional analytical methods used for these measurements are summarized. The laser analytical methods based upon Raman, absorption, fluorescence, and nonlinear spectroscopy are reviewed and evaluated for their use in fuel processing plants. After a comparison of the capabilities of the laser-based and conventional analytical methods, the promising areas of application of the laser-based methods in fuel processing plants are identified.

  10. Laser-based analytical monitoring in nuclear-fuel processing plants

    International Nuclear Information System (INIS)

    Hohimer, J.P.

    1978-09-01

    The use of laser-based analytical methods in nuclear-fuel processing plants is considered. The species and locations for accountability, process control, and effluent control measurements in the Coprocessing, Thorex, and reference Purex fuel processing operations are identified and the conventional analytical methods used for these measurements are summarized. The laser analytical methods based upon Raman, absorption, fluorescence, and nonlinear spectroscopy are reviewed and evaluated for their use in fuel processing plants. After a comparison of the capabilities of the laser-based and conventional analytical methods, the promising areas of application of the laser-based methods in fuel processing plants are identified

  11. Survey report on trends of technical development on FBR cycle in Russia. Result report of business entrusted by the Japan Nuclear Cycle Development Institute

    International Nuclear Information System (INIS)

    Takeda, Hiroshi; Kamata, Kuniko

    2001-02-01

    This survey was carried out for aims to smoothly promote the FBR cycle cooperation carried out between the Japan Nuclear Cycle Development Institute and Nuclear Energy Agency in Russia Republic and to contribute their future cooperative planning, on a survey of technical developmental trend for items shown as follows: 1) recent trend on Russian FBR cycle technology, 2) Russian laws related on Russian FBR cycle cooperation, and 3) trends on separation and reprocessing technologies in Russia. Here was described on results on the survey, shown in the following items: 1) performing method, 2) Russian FBR fuel cycle; recent technical development, 3) basic laws on nuclear energy application in Russia, and 4) trends on separation and reprocessing technologies in Russia. (G.K.)

  12. Analytical chemistry challenges at the back end of fuel cycle

    International Nuclear Information System (INIS)

    Panja, S.; Dhami, P.S.; Gandhi, P.M.

    2015-01-01

    Among the various nuclear fuel cycle activities, spent fuel reprocessing and nuclear waste management play key role for adaptation of closed fuel cycle option and success of three stage Indian nuclear power programme. Reprocessing mainly aims to recover fissile and fertile component from spent fuel using well known PUREX/THOREX processes. Waste management deals with all the activities which are essential for safe management of radioactive wastes that get generated during entire nuclear fuel cycle operation

  13. System concept for FBR cycle data base

    International Nuclear Information System (INIS)

    Kofuji, Hirohide; Saigusa, Toshiie; Hirao, Kazunori

    2000-03-01

    Accompanying with the progress of the 'Feasibility Study on FBR cycle system; FS', various kinds of technical information, facility design parameters, and related data will be obtained and they should be stored in data bases and be used as fundamental data for the FS. So the several data bases are going to be set up at each section and controlled by the management system through a local area network. Among above data bases, a prototype of FBR cycle data base that will record data for FBR scenario study and synthetic assessment is to be completed in Phase I by fiscal year 2000, so the data base system concept has been examined in the current fiscal year, 1999. As the results of the system concept examination, two types of prototypes have been selected, one is to be set up as the data table containing digital data that are extracted from technical papers, another is as image data of papers with index information. Referring to examples of data bases in other companies, it was kept in mind to use a package software for general purpose and to utilize data existing now. (author)

  14. Success tree analysis on the technologies development for FBR commercialization

    International Nuclear Information System (INIS)

    An, Shigehiro; Taniyama, Hiroshi; Nagai, Hiroshi.

    1991-01-01

    In order to obtain a secure energy supply in future, it is important to establish a system for plutonium utilization via the FBR which is superior to the uranium utilization system with respect to both safety and good economics. In spite of this obvious need, the commercialization of the FBR is facing delays. Although several factors, for example, improvement of LWR technologies, stable supply of low cost uranium, opposition to nuclear power, etc. are contributors, the primary reason for the delay is the unfavorable economics of the FBR itself. In this paper the key technologies leading to reduced FBR costs are identified and their development strategies are discussed. (author)

  15. Joyo, the irradiation and demonstration test facility of FBR development

    International Nuclear Information System (INIS)

    Aoyama, T.; Sekine, T.; Nakai, S.; Suzuki, S.

    2006-01-01

    life fuel, has been irradiated. Then, leading very high burn up irradiation tests with ODS cladding will be started sequentially, and some rigs will open the door of fuel breach to investigate fuel life limit design. Targeted to realize economical fuel, short-process MOX fuel pellets, vipac MOX fuel or metal fuel will be tested instead of conventional MOX fuel. In the transmutation area, which is aimed at reducing the environmental burden of long-lived radionuclides, MOX fuel with 5% americium added was fabricated in the Alpha-Gamma Facility at Oarai, and has just started irradiation in 2006. The test results will be contributed to the Global Nuclear Energy Partnership (GNEP). Then we will proceed to test for incineration of 99Tc and 129I, starting around 2008. In the reactor engineering area, system reliability demonstration of a self actuated shut down system was performed, starting in 2004. After the year 2010, fuel slow transient safety testing, anticipated-transient testing without scram, and in-service inspection and repair demonstrations are being considered. The test results will be utilized in the strategic feasibility study of FBR fuel cycle systems in Japan, which is similar to the Generation-IV study. In light of the shutdown of several fast reactors around the world, the ability to make such major contributions to reactor development takes on even greater significance. Irradiation tests, steady-state and safety related operations of JOYO are also expected to promote the development of JAEA's prototype FBR, Monju

  16. Role and future needs of analytical chemistry in nuclear fuel reprocessing and waste management

    International Nuclear Information System (INIS)

    Misra, S.D.

    2007-01-01

    The Indian nuclear programme encompasses the entire nuclear fuel cycle and covers a wide range of activities including mining and milling of uranium, fuel fabrication, reactor operation, spent fuel reprocessing and waste management. In the present review on the role of analytical chemistry in fuel reprocessing and waste management plants, an attempt has been made to summarize the contribution of this important chemical discipline in compositional characterization of raw materials, nuclear waste, process streams and plant products as well as in process control and troubleshooting. Some areas requiring developmental efforts by analytical chemists are also highlighted

  17. Application of the PSA method to decay heat removal systems in a large scale FBR design

    International Nuclear Information System (INIS)

    Kotake, S.; Satoh, K.; Matsumoto, H.; Sugawara, M.; Sakata, K.; Okabe, A.

    1993-01-01

    The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10 -7 /d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

  18. Fuel assembly bow: analytical modeling and resulting design improvements

    International Nuclear Information System (INIS)

    Stabel, J.; Huebsch, H.P.

    1995-01-01

    The bowing of fuel assemblies may result in a contact between neighbouring fuel assemblies and in connection with a vibration to a resulting wear or even perforation at the corners of the spacer grids of neighbouring assemblies. Such events allowed reinsertion of a few fuel assemblies in Germany only after spacer repair. In order to identify the most sensitive parameters causing the observed bowing of fuel assemblies a new computer model was develop which takes into a account the highly nonlinear behaviour of the interaction between fuel rods and spacers. As a result of the studies performed with this model, design improvements such as a more rigid connection between guide thimbles and spacer grids, could be defined. First experiences with this improved design show significantly better fuel behaviour. (author). 5 figs., 1 tabs

  19. Analytical fuel property effects: Small combustors, phase 2

    Science.gov (United States)

    Hill, T. G.; Monty, J. D.; Morton, H. L.

    1985-01-01

    The effects of non-standard aviation fuels on a typical small gas turbine combustor were studied and the effectiveness of design changes intended to counter the effects of these fuels was evaluated. The T700/CT7 turboprop engine family was chosen as being representative of the class of aircraft power plants desired for this study. Fuel properties, as specified by NASA, are characterized by low hydrogen content and high aromatics levels. No. 2 diesel fuel was also evaluated in this program. Results demonstrated the anticipated higher than normal smoke output and flame radiation intensity with resulting increased metal temperatures on the baseline T700 combustor. Three new designs were evaluated using the non standard fuels. The three designs incorporated enhanced cooling features and smoke reduction features. All three designs, when burning the broad specification fuels, exhibited metal temperatures at or below the baseline combustor temperatures on JP-5. Smoke levels were acceptable but higher than predicted.

  20. Review of HEDL fuel pin transient analyses analytical programs

    International Nuclear Information System (INIS)

    Scott, J.H.; Baars, R.E.

    1975-05-01

    Methods for analysis of transient fuel pin performance are described, as represented by the steady-state SIEX code and the PECT series of codes used for steady-state and transient mechanical analyses. The empirical fuel failure correlation currently in use for analysis of transient overpower accidents is described. (U.S.)

  1. Development of 3-D FBR heterogeneous core calculation method based on characteristics method

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Maruyama, Manabu; Hamada, Yuzuru; Nishi, Hiroshi; Ishibashi, Junichi; Kitano, Akihiro

    2002-01-01

    A new 3-D transport calculation method taking into account the heterogeneity of fuel assemblies has been developed by combining the characteristics method and the nodal transport method. In the axial direction the nodal transport method is applied, and the characteristics method is applied to take into account the radial heterogeneity of fuel assemblies. The numerical calculations have been performed to verify 2-D radial calculations of FBR assemblies and partial core calculations. Results are compared with the reference Monte-Carlo calculations. A good agreement has been achieved. It is shown that the present method has an advantage in calculating reaction rates in a small region

  2. Study on design method for seismically isolated FBR plants

    International Nuclear Information System (INIS)

    Hirata, Kazuta; Yabana, Shuichi; Ohtori, Yasuki; Ishida, Katsuhiko; Sawada, Yoshihiro; Shiojiri; Hiroo; Mazda, Taiji

    1998-01-01

    CRIEPI conducted 'Demonstration test on FBR seismic isolation system' from 1987 to 1996 under contract with Ministry of International Trade and Industry, Japan. In the demonstration test, base isolation technologies are prepared and demonstrated to apply to FBR and the design guidelines are proposed. In this report overall contents of the design guidelines entitled Design guidelines for seismically base isolated FBR plants' are included. The design guidelines, as a rule, are limited to apply to FBR plants where entire reactor building is isolated in the horizontal direction using laminated rubber bearings as isolators. The design guidelines and its concepts, however, will be useful for the development of similar guidelines for other isolation systems using different type of isolation methods and other nuclear facilities. The design guidelines consist of three parts and appendices. The first part is 'Policy for Safety Design of Base Isolated FBR Plants' specifying the principles and the requirements in the planning and the design for the safety of base isolated FBR plants. The second part is Policy for Seismic Design of Base Isolated FBR' describing the principles and the requirements in the seismic design and the evaluation of safety for base isolated FBR plants. The third part is 'Design Methods for Seismic Isolated FBR Plants' detailing the methods, procedures and parameters to be used in the design and the evaluation of safety fro base isolated FBR plants. In appendices examples of design procedures for base isolated reactor building and laminated rubber bearings as well as various test data on laminated rubber bearings, etc. are shown. (author)

  3. Nondestructive analysis of irradiated fuels

    International Nuclear Information System (INIS)

    Dudey, N.D.; Frick, D.C.

    1977-01-01

    The principal nondestructive examination techniques presently used to assess the physical integrity of reactor fuels and cladding materials include gamma-scanning, profilometry, eddy current, visual inspection, rod-to-rod spacing, and neutron radiography. LWR fuels are generally examined during annual refueling outages, and are conducted underwater in the spent fuel pool. FBR fuels are primarily examined in hot cells after fuel discharge. Although the NDE techniques are identical, LWR fuel examinations emphasize tests to demonstrate adherence to technical specification and reliable fuel performance; whereas, FBR fuel examinations emphasize aspects more related to the relative performance of different types of fuel and cladding materials subjected to variable irradiation conditions

  4. Temperature fluctuation reducing device for FBR type reactor

    International Nuclear Information System (INIS)

    Ootsuka, Fumio; Shiratori, Fumihiro.

    1991-01-01

    In existent FBR type reactors, since temperature fluctuation in the reactor upper portion has been inevitable, thermal fatigue may be caused possibly in reactor core upper mechanisms. Then, a valve is disposed to a control rod lower guide tube contained in a reactor container for automatically controlling the amount of passing coolants in accordance with the temperature of the passing coolants, to mix and control coolants passing through a fuel assembly in adjacent with the guide tube and coolants passing through the guide tube. Further, a rectification cylinder is disposed, in which a portion of coolants passing through the fuel assembly is caused to flow. An orifice is disposed to the cylinder with an exit being disposed to the upstream thereof such that the coolants not flown into the rectification cylinder and the coolants passing through the guide tube are mixed to moderate the temperature fluctuation. That is, a portion of the coolants flown into the rectification cylinder can not pass through the orifice, but flow backwardly to the upstream and is discharged out of the rectification cylinder from the coolants exit and mixed sufficiently with coolants passing through the guide tube. In this way, temperature fluctuation can be moderated. (N.H.)

  5. Analytical fuel property effects, small combustors, phase 1

    Science.gov (United States)

    Cohen, J. D.

    1983-01-01

    The effects of nonstandard aviation fuels on a typical small gas turbine combustor was analyzed. The T700/CT7 engine family was chosen as being representative of the class of aircraft power plants desired. Fuel properties, as specified by NASA, are characterized by low hydrogen content and high aromatics levels. Higher than normal smoke output and flame radiation intensity for the current T700 combustor which serves as a baseline were anticipated. It is, therefore, predicted that out of specification smoke visibility and higher than normal shell temperatures will exist when using NASA ERBS fuels with a consequence of severe reduction in cyclic life. Three new designs are proposed to compensate for the deficiencies expected with the existing design. They have emerged as the best of the eight originally proposed redesigns or combinations thereof. After the five choices that were originally made by NASA on the basis of competing performance factors, General Electric narrowed the field to the three proposed.

  6. Bio Diesel An Alternative Vehicles Fuel; Analytical View

    International Nuclear Information System (INIS)

    El Banna, S.; El Deen, O.N.

    2004-01-01

    Transesterification of a vegetable oil was conducted as early as 1853, by scientists E. Duffy and J. Patrick, many years before the first diesel engine became functional(1). Rudolf Diesel's prime model, a single 10 ft (3 m) iron cylinder with a flywheel at its base, ran on its own power for the first time in Augsburg, Germany on August 10, 1893(2). Diesel later demonstrated his engine at the World Fair in Paris, France in 1898. This engine stood as an example of Diesel's vision because it was powered by peanut oil-a bio fuel. He believed that the utilization of a biomass fuel was the real future of his engine. In a 1912 speech, Rudolf Diesel said, (I) t he use of vegetable oils for engine fuels may seem insignificant today, but such oils may become, in the course of time, as important as petroleum and the coal-tar products of the present time. Rudolf Diesel was not the only inventor to believe that biomass fuels would be the mainstay of the transportation industry. Henry Ford designed his automobiles, beginning with the 1908 Model T(1), to use ethanol. Ford was so convinced that renewable resources were the key to the success of his automobiles that he built a plant to make ethanol in the Midwest and formed a partnership with Standard Oil to sell it in their distributing stations

  7. Design criteria for FBR core components

    International Nuclear Information System (INIS)

    Desprez, D.; Ravenet, A.; Bernard, A.

    1987-01-01

    This paper outlines the general approach adopted in France to take into account the specific behavior of irradiated steel for functional and structural verification of fast breeder reactor core components. Functional verification deals with the distortions which appear in structures as a result of void swelling and irradiation creep. Specific criteria must be defined to limit these distortions to acceptable values: these criteria are highly dependent on the subassembly and core design. Structural verification deals with modifications of the mechanical properties of steel submitted to FBR flux. Conventional standards and rules are not applicable, and a new methodology must be defined to take into account the new characteristics of irradiated steel. The general R and D program set up to investigate these areas is presented here as it is implemented in France but with emphasis on integration in a joint European program

  8. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    Energy Technology Data Exchange (ETDEWEB)

    Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

    1999-07-01

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  9. Accurate reactivity void coefficient calculation for the fast spectrum reactor FBR-IME

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Fabiano P.C.; Vellozo, Sergio de O.; Velozo, Marta J., E-mail: fabianopetruceli@outlook.com, E-mail: vellozo@cbpf.br, E-mail: martajann@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Militar

    2017-07-01

    This paper aims to present an accurate calculation of the void reactivity coefficient for the FBR-IME, a fast spectrum reactor in development at the Engineering Military Institute (IME). The main design peculiarity lies in using mixed oxide [MOX - PuO{sub 2} + U(natural uranium)O{sub 2}] as fuel core. For this task, SCALE system was used to calculate the reactivity for several voids distributions generated by bubbles in the sodium beyond its boiling point. The results show that although the void reactivity coefficient is positive and location dependent, they are offset by other feedback effects, resulting in a negative overall coefficient. (author)

  10. Development of an analytical model to assess fuel property effects on combustor performance

    Science.gov (United States)

    Sutton, R. D.; Troth, D. L.; Miles, G. A.; Riddlebaugh, S. M.

    1987-01-01

    A generalized first-order computer model has been developed in order to analytically evaluate the potential effect of alternative fuels' effects on gas turbine combustors. The model assesses the size, configuration, combustion reliability, and durability of the combustors required to meet performance and emission standards while operating on a broad range of fuels. Predictions predicated on combustor flow-field determinations by the model indicate that fuel chemistry, as defined by hydrogen content, exerts a significant influence on flame retardation, liner wall temperature, and smoke emission.

  11. Evaluation of alternative fuels for the Greek road transport sector using the analytic hierarchy process

    International Nuclear Information System (INIS)

    Tsita, Katerina G.; Pilavachi, Petros A.

    2012-01-01

    This paper evaluates alternative fuels for the Greek road transport sector, using the Analytic Hierarchy Process. Seven different alternatives of fuel mode are considered in this paper: internal combustion engine (ICE) and its combination with petroleum and 1st and 2nd generation biofuels blends, fuel cells, hybrid vehicles, plug-in hybrids and electric vehicles. The evaluation of alternative fuel modes is performed according to cost and policy aspects. In order to evaluate each alternative fuel, one base scenario and ten alternative scenarios with different weight factors selection per criterion are presented. After deciding the alternative fuels’ scoring against each criterion and the criteria weights, their synthesis gives the overall score and ranking for all ten alternative scenarios. It is concluded that ICE blended with 1st and 2nd generation biofuels are the most suitable alternative fuels for the Greek road transport sector. - Highlights: ► Alternative fuels for the Greek road transport sector have been evaluated. ► The method of the AHP was used. ► Seven different alternatives of fuel mode are considered. ► The evaluation is performed according to cost and policy aspects. ► The ICE with 1st and 2nd generation biofuels are the most suitable fuels.

  12. Analytical model describes ion conduction in fuel cell membranes

    Science.gov (United States)

    Herbst, Daniel; Tse, Steve; Witten, Thomas

    2014-03-01

    Many fuel cell designs employ polyelectrolyte membranes, but little is known about how to tune the parameters (water level, morphology, etc.) to maximize ion conductivity. We came up with a simple model based on a random, discrete water distribution and ion confinement due to neighboring polymer. The results quantitatively agree with molecular dynamics (MD) simulations and explain experimental observations. We find that when the ratio of water volume to polymer volume, Vw /Vp , is small, the predicted ion self-diffusion coefficient scales roughly as Dw T√{Vw /Vp } exp(- ⋯Vp /Vw) , where Dw T is the limiting value in pure water at temperature T . At high water levels the model also agrees with MD simulation, plateauing to Dw T . The model predicts a maximum conductivity at a water level higher than is typically used, and that it would be beneficial to increase water retention even at the expense of lower ion concentration. Also, membranes would conduct better if they phase-separated into water-rich and polymer-rich regions. US ARMY MURI #W911NF-10-1-0520.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, E.A.

    1978-10-01

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

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

    International Nuclear Information System (INIS)

    Hakkila, E.A.

    1978-10-01

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

  15. Steam generator of FBR type reactor

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1992-01-01

    Liquid metal (for example, mercury) which is scarcely reactive with metal sodium is contained and cover gases which are scarcely reactive with the liquid metal are filled in a steam generator of an FBR type reactor and it is closed. The heat of primary sodium is transferred to the liquid metal, which is not reactive with sodium, in a primary thermal conduction portion. Since the temperature of the primary thermal conduction portion is high, the density is extremely low. On the other hand, since a second thermal conduction portion is kept at a single phase and the temperature is lower compared with that of the first thermal conduction portion, the density is kept high. since the density difference and gas jetting speed generate a great circulating force to liquid metal passing the opening of a partition plate, heat can be conducted on the side of water without disposing pumps. The steam concentration in the liquid metal is low being in a single phase of steams, corrosion caused from the outside of pipes of the primary thermal conduction pipe is scarcely promoted. Even if sodium leaks should be caused, since the sodium concentration in the liquid metal is extremely low and the reactivity is low, the temperature of the liquid metal is not elevated. (N.H.)

  16. Vibration-proof FBR type reactor

    International Nuclear Information System (INIS)

    Kawamura, Yutaka.

    1992-01-01

    In a reactor container in an FBR type reactor, an outer building and upper and lower portions of a reactor container are connected by a load transmission device made of a laminated material of rubber and steel plates. Each of the reactor container and the outer building is disposed on a lower raft disposed on a rock by way of a vibration-proof device made of a laminated material of rubber and steel plates. Vibration-proof elements for providing vertical eigen frequency of the vibration-proof system comprising the reactor building and the vibration-proof device within a range of 3Hz to 5Hz are used. That is, the peak of designed acceleration for response spectrum in the horizontal direction of the reactor structural portions is shifted to side of shorter period from the main frequency region of the reactor structure. Alternatively, rigidity of the vibration-proof elements is decreased to shift the peak to the side of long period from the main frequency region. Designed seismic force can be greatly reduced both horizontally and vertically, to reduce the wall thickness of the structural members, improve the plant economy and to ensure the safety against earthquakes. (N.H.)

  17. Reactor scram device for FBR type reactor

    International Nuclear Information System (INIS)

    Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

    1991-01-01

    In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

  18. Steam generator for FBR type reactor

    International Nuclear Information System (INIS)

    Watabe, Ichiro

    1998-01-01

    In a steam generator for an FBR type reactor, a heating fluid is introduced from outer circumferential one end of a cylindrical main body by way of a heating fluid distribution structure main body, put to heat exchange with helical coils disposed inside of the main body and then discharged from the other end of the main body, the distribution structure comprises a heating fluid entrance connected to and passed through the outer one circumferential end of the main body and an inner distribution duct comprising a cylindrical body having a large number of small distribution apertures on the circumference and having ring-like flanges protruded outwardly on both ends, with the end of one flange being secured to the inner surface of the main body so that an entrance for heating fluid is opened between both of the flanges. Since the other flange is left free, there is no restriction between the main body and the inner distribution duct, and the heating fluid distribution structure can be free from thermal stresses. (N.H.)

  19. Steam generator for FBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Watabe, Ichiro

    1998-11-24

    In a steam generator for an FBR type reactor, a heating fluid is introduced from outer circumferential one end of a cylindrical main body by way of a heating fluid distribution structure main body, put to heat exchange with helical coils disposed inside of the main body and then discharged from the other end of the main body, the distribution structure comprises a heating fluid entrance connected to and passed through the outer one circumferential end of the main body and an inner distribution duct comprising a cylindrical body having a large number of small distribution apertures on the circumference and having ring-like flanges protruded outwardly on both ends, with the end of one flange being secured to the inner surface of the main body so that an entrance for heating fluid is opened between both of the flanges. Since the other flange is left free, there is no restriction between the main body and the inner distribution duct, and the heating fluid distribution structure can be free from thermal stresses. (N.H.)

  20. Development of FBR piping bellows joint

    International Nuclear Information System (INIS)

    Tsukimori, Kazuyuki; Iwata, Koji

    1991-01-01

    Reduction of construction cost is one of the most important problems to realize a FBR (Fast Breeder Reactor) Plant. Significant reduction of the construction cost of a reactor building, related equipments and facilities can be expected by shortening the length of its long cooling pipes. Since the bellows has a great capacity for absorbing thermal expansion displacement, application of bellows expansion joints is considered as the most influential measure for reduction of the piping length. To confirm technological possibilities of application and practical use of bellows joints in the main piping systems, extensive R and D's, development of various methods for evaluating the strength of bellows, establishment of inspection and maintenance techniques, studies on safety logic, etc., were carried out by PNC from 1983 to 1988. Through these studies, technological possibilities of bellows joints were confirmed and the results were summarized in the 'Structural Design Guide for Class 1 Piping Bellows Expansion Joints of Fast Breeder Reactor for Elevated Temperature Service' and the 'Inspection and Maintenance Standards of Piping bellows expansion Joints'. (author)

  1. Analytical Evaluation to Determine Selected PAHs by HPLC in a Type 2 Fuel

    International Nuclear Information System (INIS)

    Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Escolano Segovia, O.; Garcia Frutos, F. J.

    2009-01-01

    An evaluation of analytical parameters to determine selected PAHs in a fuel oil type II by HPLC coupled to fluorescence and diode detectors is presented. The study was focused on four conventional treatments of these kinds of oil samples and the main objective was giving a measure of confidence level of PAH results in the fuel oil. This study was performed in the frame of the project Assessment of natural attenuation of PAHs in agricultural soil contaminated with fuel from an accidental spill (Spanish National Plain I+D+I, CTM2007-64537). This paper is presented as follows: Analysis of reference material 1582 (NIST) by using the four kinds of sample treatments of interest. Application of variance analysis to compare results obtained from type II fuel by using each sample treatment and chromatographic detector. Finally, a statistic calculation was performed to measure uncertainty components in chromatographic analysis. (Author)

  2. Development of FBR cycle data base system (II)

    International Nuclear Information System (INIS)

    Kubota, Sadae; Ohtaki, Akira; Hirao, Kazuhiro

    2003-05-01

    In the 'Feasibility Study on Commercialized FBR Cycle Systems (F/S)', scenario evaluations, cost-benefit evaluations and system characteristic evaluations to show the significance of the FBR cycle system introduction concretely are performed while design studies for FBR plants, reprocessing systems and fabrication systems are conducted. In these evaluations, future society of various conditions and situation is assumed, and investigation and analysis about needs and social effects of FBR cycle are carried out. In this study, promising FBR cycle concepts are suggested by taking information such as domestic and foreign policies and bills, an economic prediction, a supply and demand prediction of resources, a project of technology development into consideration in addition to system design information. The development of the FBR Cycle Database which this report introduced started in 1999 fiscal year to enable managed unitarity and searched reference information to use for the above scenario evaluations, cost-benefit evaluations and system characteristic evaluations. In 2000 fiscal year, its prototype was made and used tentatively, and we extracted the problems in operation and functions from that, and, in 2001 fiscal year, the entry system and the search system using the Web page were made in order to solve problems of the prototype, and started use in our group. Moreover, in 2002 fiscal year, we expanded and improved the search system and promoted the efficiency of management work, and use in JNC through intranet of the database was started. In addition, as a result of having made the entry of about 350 data in 2002 fiscal year, the collected number of the database reaches about 7,250 by the end of March, 2003. We are to continue the entry of related information of various evaluations in F/S phase 2 from now on. In addition, we are to examine improvement of convenience of the search system and cooperation with the economy database. (author)

  3. Analytical Evaluation to Determine Selected PAHs in a Contaminated Soil With Type II Fuel

    International Nuclear Information System (INIS)

    Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Garcia Frutos, F. J.

    2010-01-01

    A study on the optimization of an ultrasonic extraction method for selected PAHs determination in soil contaminated by type II fuel and by using HPLC with fluorescence detector is presented. The main objective was optimize the analytical procedure, minimizing the volume of solvent and analysis time and avoiding possible loss by evaporation. This work was carried out as part of a project that investigated a remediation process of agricultural land affected by an accidental spillage of fuel (Plan Nacional I + D + i, CTM2007-64 537). The paper is structured as: Optimization of wavelengths in the chromatographic conditions to improve resolution in the analysis of fuel samples. Optimization of the main parameters affecting in the extraction process by sonication. Comparison of results with those obtained by accelerated solvent extraction. (Author) 3 refs.

  4. Experience with, and programme of, FBR and HWR development in Japan

    International Nuclear Information System (INIS)

    Iida, M.; Sawai, S.; Nomoto, S.

    1983-01-01

    Nuclear power generation in Japan is moving forward on the long-term development programme of nuclear power from the LWR to the FBR, essentially in the same way as in other advanced nuclear countries. In this development programme the unique HWR is also included; it can use plutonium produced in LWRs together with depleted uranium before the introduction of commercial FBRs. This report describes the status of the FBR and HWR development project being carried out by the Power Reactor and Nuclear Fuel Development Corporation (PNC) based upon the Long-Term Programme on Research, Development and Utilization of Nuclear Energy in Japan. Operational experience and technical results are shown for the experimental fast reactor JOYO (100 MW(th)), which reached initial criticality in 1977. The status of the 280 MW(e) prototype reactor MONJU, under construction as of 1982, is described. The conceptual design of the subsequent 1000 MW(e) demonstration plant is outlined, as is additional future planning. Research and development results, mainly carried out at Oarai Engineering Center of PNC, are shown. The 165 MW(e) prototype FUGEN is a heavy-water-moderated, boiling-light-water-cooled, pressure-tube-type reactor which uses plutonium mixed-oxide fuel. This report describes the relationship of the fuel cycle to the HWR in Japan and also discusses the operational experience of the prototype FUGEN, which has operated since 1979. Also described is the design of the 600 MW(e) demonstration plant and the programme of related research and development. (author)

  5. Nuclear Materials Characterization in the Materials and Fuels Complex Analytical Hot Cells

    International Nuclear Information System (INIS)

    Rodriquez, Michael

    2009-01-01

    As energy prices skyrocket and interest in alternative, clean energy sources builds, interest in nuclear energy has increased. This increased interest in nuclear energy has been termed the 'Nuclear Renaissance'. The performance of nuclear fuels, fuels and reactor materials and waste products are becoming a more important issue as the potential for designing new nuclear reactors is more immediate. The Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) Analytical Laboratory Hot Cells (ALHC) are rising to the challenge of characterizing new reactor materials, byproducts and performance. The ALHC is a facility located near Idaho Falls, Idaho at the INL Site. It was built in 1958 as part of the former Argonne National Laboratory West Complex to support the operation of the second Experimental Breeder Reactor (EBR-II). It is part of a larger analytical laboratory structure that includes wet chemistry, instrumentation and radiochemistry laboratories. The purpose of the ALHC is to perform analytical chemistry work on highly radioactive materials. The primary work in the ALHC has traditionally been dissolution of nuclear materials so that less radioactive subsamples (aliquots) could be transferred to other sections of the laboratory for analysis. Over the last 50 years though, the capabilities within the ALHC have also become independent of other laboratory sections in a number of ways. While dissolution, digestion and subdividing samples are still a vitally important role, the ALHC has stand alone capabilities in the area of immersion density, gamma scanning and combustion gas analysis. Recent use of the ALHC for immersion density shows that extremely fine and delicate operations can be performed with the master-slave manipulators by qualified operators. Twenty milligram samples were tested for immersion density to determine the expansion of uranium dioxide after irradiation in a nuclear reactor. The data collected confirmed modeling analysis with very tight

  6. Chemical interaction at the FBR cladding fuel interfaces

    International Nuclear Information System (INIS)

    Delbrassine, A.; Retels, J.; Dirven, P.

    1978-01-01

    Pins containing UO 2 -30 wt.%PuO 2 and/or Caesium and/or Telluriom as doping elements have been irradiated for about 40 days in the BR2 reactor. The effects of two Cs/Te ratios, namely 1.3 and 4 and a wide range of O/M ratios on the inner corrosion of the clad have been investigated. The influence of Tellurium on the attack of the cladding has been pointed out. It may be responsible for the Chromium NS Nickel depletion in the grain boundaries of the steel. It is necessary to measure the effective Ts/Te ratio associated with the local corrosion layers. This local Cs/Te ratio should be more useful than the initial mean Cs/Te ratio in a pin for understanding the corrosion phenomene. (author)

  7. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    International Nuclear Information System (INIS)

    Horhoianu, Grigore; Ionescu, Drags; Pauna, Eduard

    2012-01-01

    When nuclear power reactors are operated in a load following (LF) mode, the nuclear fuel may be subjected to step changes in power on weekly, daily, or even hourly basis, depending on the grid's needs. Two load following tests performed in TRIGA Research Reactor of Institute for Nuclear Research (INR) Pitesti were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets in the corrosive environment. The 3D finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath at ridge region. This paper summarizes the results of the analytical assessment for SCF and their relation to CANDU fuel performance in LF tests conditions. (orig.)

  8. Semi-analytical calculation of fuel parameters for shock ignition fusion

    Directory of Open Access Journals (Sweden)

    S A Ghasemi

    2017-02-01

    Full Text Available In this paper, semi-analytical relations of total energy, fuel gain and hot-spot radius in a non-isobaric model have been derived and compared with Schmitt (2010 numerical calculations for shock ignition scenario. in nuclear fusion. Results indicate that the approximations used by Rosen (1983 and Schmitt (2010 for the calculation of burn up fraction have not enough accuracy compared with numerical simulation. Meanwhile, it is shown that the obtained formulas of non-isobaric model cannot determine the model parameters of total energy, fuel gain and hot-spot radius uniquely. Therefore, employing more appropriate approximations, an improved semianalytical relations for non-isobaric model has been presented, which  are in a better agreement with numerical calculations of shock ignition by Schmitt (2010.

  9. The effect of fuel pyrolysis on the coal particle combustion: An analytical investigation

    Directory of Open Access Journals (Sweden)

    Baghsheikhi Mostafa

    2016-01-01

    Full Text Available The aim of this work is to analytically investigate the symmetrical combustion of an isolated coal particle with the fuel pyrolysis effect. The modelling concept of coal particles is similar to that of the liquid droplet combustion but in the case of coal devolatilization, the particles do not shrink like droplet does due to evaporation of liquid fuel. The rate of devolatilization of volatiles can be calculated using the equation that is similar to Arrhenius equation. This model is based on an assumption of combined quasi-steady and transient behaviour of the process and especially focuses on predicting the variations of temperature profile, radius of pyrolysis and transfer number. It is revealed that the entrance of pyrolysis effect into the governing equations leads to the reduction in the film radius and consequently a reduction in the stand-off ratio and transfer number.

  10. MA Doping Analysis on Breeding Capability and Protected Plutonium Production of Large FBR

    Science.gov (United States)

    Permana, Sidik; Suzuki, Mitsutoshi; Kuno, Yusuke

    2010-06-01

    Spent fuel from LWR can be seen as long-live waste if it is not recycled or as a "new fuel" resource if it is recycled into the reactors. Uranium and plutonium have been used for "new fuel" resources from LWR spent fuel as MOX fuel type which is loaded into thermal reactor or fast reactor types. Other actinides from the spent fuel such as neptunium, americium and curium as minor actinide (MA) are considered to be loaded into the reactors for specific purposes, recently. Those purposes such as for increasing protected plutonium production and breeding capability for protected plutonium as well as in the same time those amount of MA can be reduced to a small quantity as a burner or transmutation purpose. Some investigations and scientific approaches are performed in order to increase a material "barrier" in plutonium isotope composition by increasing the even mass number of plutonium isotope such as Pu-238, Pu-240 and Pu-242 as plutonium protected composition. Higher material barrier which related to intrinsic properties of plutonium isotopes with even mass number (Pu-238, Pu-240 and Pu-242), are recognized because of their intense decay heat (DH) and high spontaneous fission neutron (SFN) rates. Those even number mass of plutonium isotope contribute to some criteria of plutonium characterization which will be adopted for present study such as IAEA, Pellaud and Kessler criteria (IAEA, 1972; Pellaud, 2002; and Kessler, 2007). The present paper intends to evaluate the breeding capability as a fuel sustainability index of the reactors and to analyze the composition of protected plutonium production of large power reactor based on the FaCT FBR as reference (Ohki, et al., 2008). Three dimensional FBR core configuration has been adopted which is based on the core optimization calculation of SRAC-CITATION code as reactor core analysis and JENDL-3.3 is adopted for nuclear data library. Some MA doping materials are loaded into the blanket regions which can be considered as

  11. Analytical applications of microbial fuel cells. Part I: Biochemical oxygen demand.

    Science.gov (United States)

    Abrevaya, Ximena C; Sacco, Natalia J; Bonetto, Maria C; Hilding-Ohlsson, Astrid; Cortón, Eduardo

    2015-01-15

    Microbial fuel cells (MFCs) are bio-electrochemical devices, where usually the anode (but sometimes the cathode, or both) contains microorganisms able to generate and sustain an electrochemical gradient which is used typically to generate electrical power. In the more studied set-up, the anode contains heterotrophic bacteria in anaerobic conditions, capable to oxidize organic molecules releasing protons and electrons, as well as other by-products. Released protons could reach the cathode (through a membrane or not) whereas electrons travel across an external circuit originating an easily measurable direct current flow. MFCs have been proposed fundamentally as electric power producing devices or more recently as hydrogen producing devices. Here we will review the still incipient development of analytical uses of MFCs or related devices or set-ups, in the light of a non-restrictive MFC definition, as promising tools to asset water quality or other measurable parameters. An introduction to biological based analytical methods, including bioassays and biosensors, as well as MFCs design and operating principles, will also be included. Besides, the use of MFCs as biochemical oxygen demand sensors (perhaps the main analytical application of MFCs) is discussed. In a companion review (Part 2), other new analytical applications are reviewed used for toxicity sensors, metabolic sensors, life detectors, and other proposed applications. Copyright © 2014 Elsevier B.V. All rights reserved.

  12. Core monitoring system for FBR type reactor

    International Nuclear Information System (INIS)

    Azekura, Kazuo.

    1981-01-01

    Purpose: To determine power distribution ON-line after the change of the insertion degree of control rods by the provision of means for calculating power change coefficient at each of the points due to the change in the insertion degree from the specific change of insertion degree and multiplying the same with the newest power distribution determined periodically by the diffusion calculation. Constitution: The monitoring system additionally comprises a calculation device for power change coefficient that calculates the power change coefficient in a fuel assembly adjacent to a control rod based on the data concerning the operation of the control rod, and a provisional power distribution calculation device that executes multiplication between the power distribution calculated in a periodical power distribution calculation device based on the calculation instruction and stored in the core and the power change coefficient from the power change coefficient calculation device and forecasts the provisional power distribution. Then, based on the result of the foregoing calculations, 2-dimensional power distribution, maximum temperature for the cladding tube of the specified fuel assembly, maximum temperature of pellets in the specified fuel assembly, maximum power density and the like are calculated in various display value calculation devices and displayed on a display device. (Horiuchi, T.)

  13. JAEA FBR Plant Engineering Center annual report 2011

    International Nuclear Information System (INIS)

    2012-11-01

    The FBR Plant Engineering Center was established on April 1, 2009 located in a research building, of which care is taken by the International Nuclear Information Training Center, Tsuruga Head Office, at Shiraki in Tsuruga. The mission of the center is to perform R and D (research and development) works both for analysis of operational experiences at the prototype fast breeder reactor “Monju” and for technology development concerning design and operation of “Monju”. Moreover it is also required to apply the results to next generation fast breeder reactors, which is an important role of Advanced Nuclear System Research and Development Directorate. And in these R and D activities, it is expected to conduct the works in cooperation with domestic or foreign research organizations or universities by a joint-study or a collaborative-work manner. The R and D activities have been carried out specifically on the “demonstration of the reliability as a power generation plant” and “establishment of sodium handling technology”, which are originally intended missions of “Monju”. And the other R and Ds have been promoted both for the plant engineering, such as plant maintenance, to effectively use an existing reactor in order to apply the R and D results to a future demonstration reactor, and for the irradiation test study, such as advanced fuel irradiation, to use “Monju” as an irradiation test bed. In order to perform these R and D activities, five R and D groups have been set up in the center. They are operation-and-maintenance engineering, sodium engineering, reactor-core-and-fuel engineering, plant engineering, and safety engineering groups. However, the Japanese atomic energy policy is being reviewed after the accident of the Fukushima Daiichi nuclear power station caused by a tsunami generated by the Tohoku-district-off-the-Pacific-Ocean Earthquake on March 11, 2011, and all the R and D activities using “Monju” have been suspended since late 2011

  14. Current status of FBR development in Japan

    International Nuclear Information System (INIS)

    Ichimiya, Masakazu

    2008-01-01

    'Fast Reactor Cycle Technology Development (FaCT)' project has been conducted since 2006. In this project, design study and research and development (R and D) on innovative technologies for fast reactor (FR) cycle system are implemented in order to present the conceptual designs of commercial and demonstration facilities by 2015 and start operating demonstration fast reactor in 2025. The R and Ds has been stepped forward into the development stage to establish the realization of innovative technologies which bring excellent performance to fast reactor cycle system. The purpose of R and D by 2010 is to decide whether innovative technologies shall be adopted. In the FaCT project, R and D stage for the realization of innovative technologies, it is important to take full advantage of JAEA's R and D facilities toward demonstration and commercialization stages. Particularly, it is indispensible for the realization of innovative technologies to develop the fuel and material by irradiation tests using an experimental reactor 'Joyo', to verify the reliability for power-generating plant through the experience of operation and maintenance, and to establish technologies of operation, maintenance and repair for the plant with a prototype reactor 'Monju'. Several possible R and D have been effectively carried out within the frameworks of international cooperation, such as Global Nuclear Energy Partnership (GNEP), Generation IV International Forum (GIF), and International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). (author)

  15. Multigroup, spatial kinetics for MOX-fueled LWRs based on harmonic analytical nodal method

    Science.gov (United States)

    Jiang, Guobing

    2000-10-01

    There has been substantial evidence during the last several years that the core neutronics methods that have been developed for uranium fueled LWRs do not perform satisfactorily when applied to the same cores fueled with mixed oxide, or more generally to heterogeneous cores with very different neutron spectra. A two-dimensional, 97 group MOX benchmark problem was developed and applied to analyze deficiencies of the current generation of LWR analysis methods. The errors in the current two group, coarse mesh nodal diffusion methods were described in terms of four primary effects: (1) a homogenization effect, (2) a spatial discretization effect, (3) a group collapsing effect, and (4) a transport effect. The specific objective of the research here was to address the first three of these effects with the development of a four energy group advanced nodal method. Several methods have been proposed over the last several years for extending the current class of nodal methods to four energy groups. A Taylor series analysis was performed of the order of error in the various analytic nodal methods proposed. The analysis showed that the harmonic part of the error dominated in the Taylor expansion and it was therefore prudent to retain the harmonic solution in all four energy groups. A new nodal kernel referred to as the Harmonic Analytic Nodal Method (HANM) was developed and implemented within the framework of the nonlinear nodal method. HANM was applied to a MOX benchmark problem and results were compared to a 97 group reference solution. The errors in the two group solution were reduced by about 50% through the application of a four group HANM with minimal increase in the computational burden.

  16. Bio-analytical applications of microbial fuel cell-based biosensors for onsite water quality monitoring.

    Science.gov (United States)

    ElMekawy, A; Hegab, H M; Pant, D; Saint, C P

    2018-01-01

    Globally, sustainable provision of high-quality safe water is a major challenge of the 21st century. Various chemical and biological monitoring analytics are presently utilized to guarantee the availability of high-quality water. However, these techniques still face some challenges including high costs, complex design and onsite and online limitations. The recent technology of using microbial fuel cell (MFC)-based biosensors holds outstanding potential for the rapid and real-time monitoring of water source quality. MFCs have the advantages of simplicity in design and efficiency for onsite sensing. Even though some sensing applications of MFCs were previously studied, e.g. biochemical oxygen demand sensor, recently numerous research groups around the world have presented new practical applications of this technique, which combine multidisciplinary scientific knowledge in materials science, microbiology and electrochemistry fields. This review presents the most updated research on the utilization of MFCs as potential biosensors for monitoring water quality and considers the range of potentially toxic analytes that have so far been detected using this methodology. The advantages of MFCs over established technology are also considered as well as future work required to establish their routine use. © 2017 The Society for Applied Microbiology.

  17. Experimental and analytical analysis of polarization and water transport behaviors of hydrogen alkaline membrane fuel cell

    Science.gov (United States)

    Huo, Sen; Zhou, Jiaxun; Wang, Tianyou; Chen, Rui; Jiao, Kui

    2018-04-01

    Experimental test and analytical modeling are conducted to investigate the operating behavior of an alkaline electrolyte membrane (AEM) fuel cell fed by H2/air (or O2) and explore the effect of various operating pressures on the water transfer mechanism. According to the experimental test, the cell performance is greatly improved through increasing the operating pressure gradient from anode to cathode which leads to significant liquid water permeation through the membrane. The high frequency resistance of the A901 alkaline membrane is observed to be relatively stable as the operating pressure varies based on the electrochemical impedance spectroscopy (EIS) method. Correspondingly, based on the modeling prediction, the averaged water content in the membrane electrode assembly (MEA) does not change too much which leads to the weak variation of membrane ohmic resistance. This reveals that the performance enhancement should give the credit to better electro-chemical reaction kinetics for both the anode and cathode, also prone by the EIS results. The reversion of water back diffusion direction across the membrane is also observed through analytical solution.

  18. Scenarios and analytical methods for UF6 releases at NRC-licensed fuel cycle facilities

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Dykstra, J.; Holt, D.D.; Huxtable, W.P.; Just, R.A.; Williams, W.R.

    1984-06-01

    This report identifies and discusses potential scenarios for the accidental release of UF 6 at NRC-licensed UF 6 production and fuel fabrication facilities based on a literature review, site visits, and DOE enrichment plant experience. Analytical tools needed for evaluating source terms for such releases are discussed, and the applicability of existing methods is reviewed. Accident scenarios are discussed under the broad headings of cylinder failures, UF 6 process system failures, nuclear criticality events, and operator errors and are categorized by location, release source, phase of UF 6 prior to release, release flow characteristics, release causes, initiating events, and UF 6 inventory at risk. At least three types of releases are identified for further examination: (1) a release from a liquid-filled cylinder outdoors, (2) a release from a pigtail or cylinder in a steam chest, (3) an indoor release from either (a) a pigtail or liquid-filled cylinder or (b) other indoor source depending on facility design and operating procedures. Indoor release phenomena may be analyzed to determine input terms for a ventilation model by using a time-dependent homogeneous compartment model or a more complex hydrodynamic model if time-dependent, spatial variations in concentrations, temperature, and pressure are important. Analytical tools for modeling directed jets and explosive releases are discussed as well as some of the complex phenomena to be considered in analyzing UF 6 releases both indoors and outdoors

  19. Fast breeder fuel cycle

    International Nuclear Information System (INIS)

    1978-07-01

    This contribution is prepared for the answer to the questionnaire of working group 5, subgroup B. B.1. is the short review of the fast breeder fuel cycles based on the reference large commercial Japanese LMFBR. The LMFBRs are devided into two types. FBR-A is the reactor to be used before 2000, and its burnup and breeding ratio are relatively low. The reference fuel cycle requirement is calculated based on the FBR-A. FBR-B is the one to be used after 2000, and its burnup and breeding ratio are relatively high. B.2. is basic FBR fuel reprocessing scheme emphasizing the differences with LWR reprocessing. This scheme is based on the conceptual design and research and development work on the small scale LMFBR reprocessing facility of Japan. The facility adopts a conventional PUREX process except head end portions. The report also describes the effects of technical modifications of conventional reprocessing flow sheets, and the problems to be solved before the adoption of these alternatives

  20. TN-68 Spent Fuel Transport Cask Analytical Evaluation for Drop Events

    International Nuclear Information System (INIS)

    Shah, M.J.; Klymyshyn, Nicholas A.; Adkins, Harold E.; Koeppel, Brian J.

    2007-01-01

    The U.S. Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (10 CFR), Title 10, Part 71 (1). Both the International Atomic Energy Agency regulations for transporting radioactive materials (2, paragraph 727), and 10 CFR 71.73 require casks to be evaluated for hypothetical accident conditions, which includes a 9-meter (m) (30-ft) drop-impact event onto a flat, essentially unyielding, horizontal surface, in the most damaging orientation. This paper examines the behavior of one of the NRC certified transportation casks, the TN-68 (3), for drop-impact events. The specific area examined is the behavior of the bolted connections in the cask body and the closure lid, which are significantly loaded during the hypothetical drop-impact event. Analytical work to evaluate the NRC-certified TN-68 spent fuel transport cask (3) for a 9-m (30-ft) drop-impact event on a flat, unyielding, horizontal surface, was performed using the ANSYS (4) and LS DYNA (5) finite-element analysis codes. The models were sufficiently detailed, in the areas of bolt closure interfaces and containment boundaries, to evaluate the structural integrity of the bolted connections under 9-m (30-ft) free-drop hypothetical accident conditions, as specified in 10 CFR 71.73. Evaluation of the cask for puncture, caused by a free drop through a distance of 1-m (40-in.) onto a mild steel bar mounted on a flat, essentially unyielding, horizontal surface, required by 10 CFR 71.73, was not included in the current work, and will have to be addressed in the future. Based on the analyses performed to date, it is concluded that, even though brief separation of the flange and the lid surfaces may occur under some conditions, the seals would close at the end of the drop events, because the materials remain elastic during the duration of the event

  1. Design and analytic evaluation of a rim effect reduction type LWR fuel for extending burnup

    International Nuclear Information System (INIS)

    Matsumura, Tetsuo; Kameyama, Takanori; Kinoshita, Motoyasu

    1991-01-01

    We have designed a new concept fuel design 'Rim effect reduction type fuel' which has thin natural UO 2 layer on surface of a UO2 pellet. Our neutronic analyses with ANRB code show this fuel design can reduce rim effect (burnup at plelet rim) by about 30 GWd/t comparing a normal fuel. It is known that a high burnup fuel has different microstructure from as-fabricated one at fuel rim (which is called as rim region) due to rim effect. Therefore this fuel design can expect smaller rim region than a normal fuel. Our fuel performance analyses with EIMUS code show this fuel design can reduce fuel center temperature at high burnup if thermal conductivity of fuel pellet decreases with burnup in inverse proportion. However, this fuel design increases fuel center temperature at low and middle burnup than a normal fuel due to increase of thermal power density at pellet center. Additionally Irradiation experiment of this fuel design can be considered to offer important data which make clear the relation between rim effect and fuel performance. (author)

  2. Analytical, 1-Dimensional Impedance Model of a Composite Solid Oxide Fuel Cell Cathode

    DEFF Research Database (Denmark)

    Mortensen, Jakob Egeberg; Søgaard, Martin; Jacobsen, Torben

    2014-01-01

    An analytical, 1-dimensional impedance model for a composite solid oxide fuel cell cathode is derived. It includes geometrical parameters of the cathode, e.g., the internal surface area and the electrode thickness, and also material parameters, e.g., the surface reaction rate and the vacancy...... diffusion coefficient. The model is successfully applied to a total of 42 impedance spectra, obtained in the temperature range 555°C–852°C and in the oxygen partial pressure range 0.028 atm–1.00 atm for a cathode consisting of a 50/50 wt% mixture of (La0.6Sr0.4)0.99CoO3 − δ and Ce0.9Gd0.1O1.95 − δ....... The surface exchange coefficient in oxygen for T = 802°C and [Formula] is found to be kEx = 1.42 × 10− 4 m s− 1 and with an activation energy of Ea = 107 kJ mol− 1, in fair agreement with literature. A parameter variation and a steady state analysis is performed, verifying the soundness of the model...

  3. An Innovative Hybrid 3D Analytic-Numerical Approach for System Level Modelling of PEM Fuel Cells

    Directory of Open Access Journals (Sweden)

    Gregor Tavčar

    2013-10-01

    Full Text Available The PEM fuel cell model presented in this paper is based on modelling species transport and coupling electrochemical reactions to species transport in an innovative way. Species transport is modelled by obtaining a 2D analytic solution for species concentration distribution in the plane perpendicular to the gas-flow and coupling consecutive 2D solutions by means of a 1D numerical gas-flow model. The 2D solution is devised on a jigsaw puzzle of multiple coupled domains which enables the modelling of parallel straight channel fuel cells with realistic geometries. Electrochemical and other nonlinear phenomena are coupled to the species transport by a routine that uses derivative approximation with prediction-iteration. A hybrid 3D analytic-numerical fuel cell model of a laboratory test fuel cell is presented and evaluated against a professional 3D computational fluid dynamic (CFD simulation tool. This comparative evaluation shows very good agreement between results of the presented model and those of the CFD simulation. Furthermore, high accuracy results are achieved at computational times short enough to be suitable for system level simulations. This computational efficiency is owed to the semi-analytic nature of its species transport modelling and to the efficient computational coupling of electrochemical kinetics and species transport.

  4. Long-term logistic analysis of FBR introduction strategy: avoiding both uranium and plutonium shortage

    International Nuclear Information System (INIS)

    Suzuki, T.

    1995-01-01

    Despite comfortable predictions on short to mid-term uranium resources, there is still a concern about long-term availability of competitive uranium resources. In order to achieve substantial uranium saving, early introduction of Fast Breeder Reactor (FBR) is desirable. But it is also known that rapid introduction of FBR could result in plutonium storage. Will there be enough plutonium on a global scale to sustain fast FBR growth? is there any other way to save uranium resource? This paper concludes that multi-option strategies to achieve flexible long-term strategy to avoid both uranium and plutonium storage are desirable. (authors)

  5. Analytical calculation of the fuel temperature reactivity coefficient for pebble bed and prismatic high temperature reactors for plutonium and uranium-thorium fuels

    International Nuclear Information System (INIS)

    Talamo, Alberto

    2007-01-01

    We analytically evaluated the fuel coefficient of temperature both for pebble bed and prismatic high temperature reactors when they utilize as fuel plutonium and minor actinides from light water reactors spent fuel or a mixture of 50% uranium, enriched 20% in 235 U, and 50% thorium. In both cores the calculation involves the evaluation of the resonances integrals of the high absorbers fuel nuclides 240 Pu, 238 U and 232 Th and it requires the esteem of the Dancoff-Ginsburg factor for a pebble bed or prismatic core. The Dancoff-Ginsburg factor represents the only discriminating parameter in the results for the two different reactors types; in fact, both the pebble bed and the prismatic reactors share the same the pseudo-cross-section describing an infinite medium made of graphite filled by TRISO particles. We considered only the resolved resonances with a statistical spin factor equal to one and we took into account 267, 72, 212 resonances in the range 1.057-5692, 6.674-14485, 21.78-3472 eV for 240 Pu, 238 U and 232 Th, respectively, for investigating the influence on the fuel temperature reactivity coefficient of the variation of the TRISO kernel radius and TRISO particles packing fraction from 100, 200 to 300 μm and from 10% to 50%, respectively. Finally, in the pebble bed core, we varied the radius of the pebble for setting a fuel temperature reactivity coefficient similar to the one of a prismatic core

  6. The design of seismic isolated demonstration FBR plant

    International Nuclear Information System (INIS)

    Inagaki, Tatsutoshi; Watanabe, Yukio; Ueta, Masahiro; Tarutani, Kohei; Shibata, Yoji; Okada, Keizo; Hayashi, Yuji

    1996-01-01

    The demonstration fast breeder reactor (DFBR) has been under development as an essential step toward the commercialization of the fast breeder reactor (FBR) around the year 2030. The Japan Atomic Power Company (JAPC) has completed the conceptual design of the DFBR, based on which the Japanese utilities have decided that the DFBR will be a top-entry loop-type reactor with electricity output of 660 MWe and a horizontal seismic isolated plant, which will reduce the horizontal seismic load on the building and components and thus reduce the amount of plant materials. The design study on optimization of the DFBR plant was started in 1994. The purpose of this study is to reduce the plant construction cost, to use additional measures to increase core safety, to assess the robustness required for the containment facility, and to gauge the potential for licensing the horizontal seismic isolation plant. This paper outlines the design concept of the seismic isolation DFBR

  7. Leak detector for a steam generator in FBR type reactors

    International Nuclear Information System (INIS)

    Miyaji, Nobuyoshi.

    1979-01-01

    Purpose: To facilitate maintenance for liquid leak detectors such as exchange of nickel membrane sensors during operation in a sodium-cooled fbr type reactor. Constitution: A pipeway capable of supplying a cover gas such as argon into the cylinder of a hydrogen detector containing a nickel membrane sensor is provided in a liquid leak detector constituting a part of a by-pass loop. The pipeway is also adapted to be evacuated. A pipeway and a small sodium tank for drain use are provided on the side of the by-pass loop near valves. Then, after closing the inlet and outlet valves to disconnect the by-pass loop from the sodium main pipeway, the cover gas is supplied to drive liquid sodium to the drain tank. After the drain of the liquid sodium, the sensor can be replaced. (Ikeda, J.)

  8. Development of thermal hydraulic analysis code for IHX of FBR

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Naohara, Nobuyuki

    1991-01-01

    In order to obtain flow resistance correlations for thermal-hydrauric analysis code concerned with an intermediate heat exchanger (IHX) of FBR, the hydraulic experiment by air was carried out through a bundle of tubes arranged in an in-line and staggard fashion. The main results are summarized as follows. (1) On pressure loss per unit length of a tube bundle, which is densely a regular triangle arrangement, the in-line fashion is almost the same as the staggard one. (2) In case of 30deg sector model for IHX tube bundle, pressure loss is 1/3 in comparison with the in-line or staggard arrangement. (3) By this experimental data, flow resistance correlations for thermalhydrauric analysis code are obtained. (author)

  9. After-heat removing system in FBR type reactor

    International Nuclear Information System (INIS)

    Ohashi, Yukio.

    1990-01-01

    The after-heat removing system of the present invention removes the after heat generated in a reactor core without using dynamic equipments such as pumps or blowers. There are disposed a first heat exchanger for heating a heat medium by the heat in a reactor container and a second heat exchanger situated above the first heat exchanger for spontaneously air-cooling the heat medium. Recycling pipeways connect the first and the second heat exchangers to form a recycling path for the heat medium. Then, since the second heat exchanger for spontaneously air-cooling the heat medium is disposed above the first heat exchanger and they are connected by the recycling pipeways, the heat medium can be circulated spontaneously. Accordingly, dynamic equipments such as pumps or blowers are no more necessary. As a result, the after-heat removing system of the FBR type reactor of excellent safety and reliability can be obtained. (I.S.)

  10. BWR noise spectra and application of noise analysis to FBR

    International Nuclear Information System (INIS)

    Nomura, T.

    1975-01-01

    Work related to noise analysis, in Tokyo Shibaura Electric Co. Ltd. (Toshiba) and Nippon Atomic Industry Group Co. Ltd. (NAIG) for the past several years is reviewed. After considering the Japan-United States Seminar on Reactor Noise Analysis in 1968, other subjects discussed were boiling water reactor noise analysis and work in relation to FBR. Parts of these are related to each other. For example, boiling detection and temperature fluctuations are problems pertinent to both fields. As the main problems in zero-power-reactor noise are now basically understood, only a brief description of the experiments involving the advanced two detector method is made. Focus is rather placed on the area of power plant noise. (author)

  11. Analytic Methods for Benchmarking Hydrogen and Fuel Cell Technologies; NREL (National Renewable Energy Laboratory)

    Energy Technology Data Exchange (ETDEWEB)

    Melaina, Marc; Saur, Genevieve; Ramsden, Todd; Eichman, Joshua

    2015-05-28

    This presentation summarizes NREL's hydrogen and fuel cell analysis work in three areas: resource potential, greenhouse gas emissions and cost of delivered energy, and influence of auxiliary revenue streams. NREL's hydrogen and fuel cell analysis projects focus on low-­carbon and economic transportation and stationary fuel cell applications. Analysis tools developed by the lab provide insight into the degree to which bridging markets can strengthen the business case for fuel cell applications.

  12. Continuous Ethanol Production Using Immobilized-Cell/Enzyme Biocatalysts in Fluidized-Bed Bioreactor (FBR)

    Energy Technology Data Exchange (ETDEWEB)

    Nghiem, NP

    2003-11-16

    The immobilized-cell fluidized-bed bioreactor (FBR) was developed at Oak Ridge National Laboratory (ORNL). Previous studies at ORNL using immobilized Zymomonas mobilis in FBR at both laboratory and demonstration scale (4-in-ID by 20-ft-tall) have shown that the system was more than 50 times as productive as industrial benchmarks (batch and fed-batch free cell fermentations for ethanol production from glucose). Economic analysis showed that a continuous process employing the FBR technology to produce ethanol from corn-derived glucose would offer savings of three to six cents per gallon of ethanol compared to a typical batch process. The application of the FBR technology for ethanol production was extended to investigate more complex feedstocks, which included starch and lignocellulosic-derived mixed sugars. Economic analysis and mathematical modeling of the reactor were included in the investigation. This report summarizes the results of these extensive studies.

  13. Improved analysis on multiple recycling of fuel in prototype fast ...

    Indian Academy of Sciences (India)

    2015-11-27

    Nov 27, 2015 ... An FBR closed fuel cycle involves recycling of the discharge fuel, after reprocessing and refabrication, to utilize the unburnt fuel remains and the freshly bred fissile material. Our previous study in this regard for the PFBR indicated a comfortable feasibility of multiple recycling with selfsufficiency. In the ...

  14. Standardization of dosimetry and damage analysis work for U.S. LWR, FBR, and MFR development program

    International Nuclear Information System (INIS)

    McElroy, W.N.; Doran, D.G.; Gold, R.; Morgan, W.C.; Grundl, J.A.; McGarry, E.D.; Kam, F.B.K.; Swank, J.H.; Odette, G.R.

    1978-01-01

    The accuracy requirements for various measured/calculated exposure and correlation parameters associated with current dosimetry and damage analysis procedures and practices depend on the accuracy needs of reactor development efforts in testing, design, safety, operations, and surveillance programs. Present state-of-the-art accuracies are estimated to be in the range of +-2 to 30 percent (1 sigma), depending on the particular parameter. There now appears to be international agreement, at least for the long term, that most reactor fuels and materials programs will not be able to accept an uncertainty greater than about +5 percent (1 sigma). The current status of dosimetry and damage analysis standardization work within the U.S. for LWR, FBR and MFR is reviewed in this paper

  15. The manual of a computer software 'FBR Plant Planning Design Prototype System'

    International Nuclear Information System (INIS)

    2003-10-01

    This is a manual of a computer software 'FBR Plant Planning Design Prototype System', which enables users to conduct case studies of deviated FBR design concepts based on 'MONJU'. The calculations simply proceed as the user clicks displayed buttons, therefore step-by-step explanation is supposed not be necessary. The following pages introduce only particular features of this software, i.e, each interactive screens, functions of buttons and consequences after clicks, and the quitting procedure. (author)

  16. Study on reprocessing plant for the transition period from LWR to FBR. Concept for the next reprocessing plant from the viewpoint of plutonium supply and demand

    International Nuclear Information System (INIS)

    Shimada, Takashi; Matsui, Minefumi; Nishimura, Masashi; Ishida, Yasuhiro; Mori, Yukihide; Kuroda, Kazuhiko

    2011-01-01

    This paper discusses the reprocessing plant concept suitable for the transition period from the Light Water Reactors (LWRs) to the Fast Breeder Reactors (FBRs). This transition requires the reprocessing of spent fuels in order to supply an adequate volume of fissile plutonium (Pu-fissile) for the FBRs. The transition period would continue for more than 60 years, and the reprocessing plant should match with the change in the power generation plan during the transition period. The ability to supply Pu-fissile has been evaluated for two plant concepts. One is the independent-type concept, which contains two processes for reprocessing either LWR or FBR fuels. The other is the modularized-type concept, which contains only one process for reprocessing both the LWR and FBR fuels. The result showed the superiority of the modularized-type concept over the independent-type concept, because the former can enhance the ability to supply Pu-fissile with less reprocessing capacity. Therefore, the reprocessing plant suitable for the transition period is that based on the modularized-type concept. (author)

  17. Analytical solution and experimental validation of the energy management problem for fuel cell hybrid vehicles

    NARCIS (Netherlands)

    P.P.J. van den Bosch; Edwin Tazelaar; M. Grimminck; Stijn Hoppenbrouwers; Bram Veenhuizen

    2011-01-01

    The objective of an energy management strategy for fuel cell hybrid propulsion systems is to minimize the fuel needed to provide the required power demand. This minimization is defined as an optimization problem. Methods such as dynamic programming numerically solve this optimization problem.

  18. Analytical solution of the energy management for fuel cell hybrid propulsion systems

    NARCIS (Netherlands)

    P.P.J. van den Bosch; E. Tazelaar; Bram Veenhuizen

    2012-01-01

    The objective of an energy management strategy for fuel cell hybrid propulsion systems is to minimize the fuel needed to provide the required power demand. This minimization is defined as an optimization problem. Methods such as dynamic programming numerically solve this optimization problem.

  19. Current status of feasibility studies on commercialized fuel cycle system for Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Ojima, Hisao; Nagaoki, Yoshihiro

    2000-01-01

    A 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' is underway at the Japan Nuclear Cycle Development Institute (JNC). The study will select the promising concepts with their R and D tasks in order to commercialize the fast breeder reactor (FBR) cycle system. The feasibility studies (F/S) have to present surveyed and screened various relevant technologies, and defined the design requirement of the commercialized fuel cycle system for FBR. The promising technical options are being evaluated and conceptual designs are being examined. At the end of JFY2000, several candidate concepts of the commercialized FBR cycle system will be proposed. (author)

  20. Quality assurance program for surveillance of fast reactor mixed oxide fuel analytical chemistry

    International Nuclear Information System (INIS)

    Rein, J.E.; Zeigler, R.K.; Waterbury, G.R.; McClung, W.E.; Praetorius, P.R.; Delvin, W.L.

    1976-01-01

    An effective quality assurance program for the chemical analysis of nuclear fuel is essential to assure that the fuel will meet the strict chemical specifications required for optimum reactor performance. Such a program has been in operation since 1972 for the fuels manufactured for the Fast Flux Test Facility. This program, through the use of common quality control and calibration standards, has consistently provided high levels of agreement among laboratories in all areas of analysis. The paper presented gives a summary of the chemical specifications for the fuel and source material, an outline of the requirements for laboratory qualifications and the preparation of calibration and quality control materials, general administration details of the plan, and examples where the program has been useful in solving laboratory problems

  1. The Analytical Repository Source-Term (AREST) model: Analysis of spent fuel as a nuclear waste form

    International Nuclear Information System (INIS)

    Apted, M.J.; Liebetrau, A.M.; Engel, D.W.

    1989-02-01

    The purpose of this report is to assess the performance of spent fuel as a final waste form. The release of radionuclides from spent nuclear fuel has been simulated for the three repository sites that were nominated for site characterization in accordance with the Nuclear Waste Policy Act of 1982. The simulation is based on waste package designs that were presented in the environmental assessments prepared for each site. Five distinct distributions for containment failure have been considered, and the release for nuclides from the UO 2 matrix, gap (including grain boundary), crud/surface layer, and cladding has been calculated with the Analytic Repository Source-Term (AREST) code. Separate scenarios involving incongruent and congruent release from the UO 2 matrix have also been examined using the AREST code. Congruent release is defined here as the condition in which the relative mass release rates of a given nuclide and uranium from the UO 2 matrix are equal to their mass ratios in the matrix. Incongruent release refers to release of a given nuclide from the UO 2 matrix controlled by its own solubility-limiting solid phase. Release of nuclides from other sources within the spent fuel (e.g., cladding, fuel/cladding gap) is evaluated separately from either incongruent or congruent matrix release. 51 refs., 200 figs., 9 tabs

  2. Study on the FBR cycle introduction scenario. 2. A study on the role of nuclear energy under the diversity of energy supply-and-demand

    International Nuclear Information System (INIS)

    Ohtaki, Akira; Ono, Kiyoshi; Hirao, Kazunori

    2002-03-01

    This report concerns it self with the results of an investigation about the possibility of future nuclear utilization in the part of FBR Cycle Introduction Scenario Study in the JNC's 'Feasibility Study on Commercialized Fast Reactor Cycle System (the F/S)'. We have investigated about the problems that confront energy industries and electric power companies, the capacities of distributed generation, the coexistence method of a distributed generation and large-scale power supply generation, and the development status of a small-scale nuclear reactor from a wide viewpoint. Especially the spread of distributed generation causes the decrease of the electricity demand which the electric power companies supplies. Since introduction scale of a distributed power supply is also expected to increase in the future, it will give some influences to a future nuclear plan and a power supply plan. The hydrogen utilization with out greenhouse gas mission is expected to spread with distributed generation, such as a fuel cell and a micro-gas turbine. Therefore, we proposed the new business model that the hydrogen produced by using nuclear surplus electricity is consumed distributed generation, such as a fuel cell and a micro-gas turbine. We plan to evaluate quantitatively the best power supply composition based on this load stability business model, FBR introduction capacities, the load factor, and the amount of CO 2 reduction. (author)

  3. Development of advanced methodology for defect assessment in FBR power plants

    International Nuclear Information System (INIS)

    Meshii, Toshiyuki; Asayama, Tai

    2001-03-01

    As a preparation for developing a code for FBR post construction code, (a) JSME Code NA1-2000 was reviewed on the standpoint of applying it to FBR power plants and the necessary methodologies for defect assessment for FBR plants were pointed out (b) large capacity-high speed fatigue crack propagation (FCP) testing system was developed and some data were acquired to evaluate the FCP characteristics under thermal stresses. Results showed that the extended research on the following items are necessary for developing FBR post construction code. (1) Development of assessment for multiple defects due to creep damage. Multiple defects due to creep damage are not considered in the existing code, which is established for nuclear power plants in service under negligible-creep temperature. Therefore method to assess the integrity of these multiple defects due to creep damage is necessary. (2) FCP resistance for small load. Since components of FBR power plants are designed to minimize thermal stresses, the accuracy of FCP resistance for small load is important to estimate the crack propagation under thermal stresses accurately. However, there is not a sufficient necessary FCP data for small loads, maybe because the data is time consuming. Therefore we developed a large capacity-high speed FCP testing system, made a guideline for accelerated test and acquired some data to meet the needs. Continuous efforts to accumulate small load FCP data for various materials are necessary. (author)

  4. Modelling of solid polymer and direct methanol fuel cells: Phenomenological equations and analytical solutions

    Science.gov (United States)

    Kauranen, P. S.

    1993-04-01

    In the solid state concept of a direct methanol fuel cell (DMFC), methanol is directly oxidized at the anode of a solid polymer electrolyte fuel cell (SPEFC). Mathematical modelling of the transport and reaction phenomena within the electrodes and the electrolyte membrane is needed in order to get a closer insight into the operation of the fuel cell. In the work, macro-homogenous porous electrode and dilute solution theories are used to derive the phenomenological equations describing the transport and reaction mechanisms in a SPEFC single cell. The equations are first derived for a conventional H2/air SPEFC, and then extended for a DMFC. The basic model is derived in a one dimensional form in which it is assumed that species transport take place only in the direction crossing the cell sandwich. In addition, two dimensional descriptions of the catalyst layer are reviewed.

  5. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    Abstract. In a thermal neutron reactor, multiple recycle of U–Pu fuel is not possible due to degradation of fissile content of Pu in just one recycle. In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving ...

  6. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder ...

  7. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    Fast breeder reactors (FBRs) are destined to play a crucial role inthe Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of ...

  8. Analytical throughput-estimating methods for the Hot Fuel Examination Facility

    International Nuclear Information System (INIS)

    Keyes, R.W.; Phipps, R.D.

    1983-01-01

    The Hot Fuel Examination Facility (HFEF) supports the operation and experimental programs of the major Liquid Metal Fast Breeder Reactor (LMFBR) test facilities; specifically, the Fast Flux Test Facility (FFTF), the Experimental Breeder Reactor II (EBR-II), and the Transient Reactor Test (TREAT) Facility. Successful management of HFEF and of LMFBR safety and fuels and materials programs, therefore, requires reliable information regarding HFEF's capability to handle expected or proposed program work loads. This paper describes the 10-step method that has been developed to consider all variables which significantly affect the HFEF examination throughput and quickly provide the necessary planning information

  9. A structurally based analytic model for estimation of biomass and fuel loads of woodland trees

    Science.gov (United States)

    Robin J. Tausch

    2009-01-01

    Allometric/structural relationships in tree crowns are a consequence of the physical, physiological, and fluid conduction processes of trees, which control the distribution, efficient support, and growth of foliage in the crown. The structural consequences of these processes are used to develop an analytic model based on the concept of branch orders. A set of...

  10. Early-in-life thermal performance of UO2--PuO2 fast reactor fuel

    International Nuclear Information System (INIS)

    Baker, R.B.; Leggett, R.D.

    1979-01-01

    Results from the combined analyses of two thermal performance tests, HEDL P-19 and HEDL P-20 are described. The tests were designed to provide data on the power required to cause incipient fuel melting early in life under conditions prototypic of FFTF driver fuel pins and similar FBR fuel systems

  11. JOYO modification program for demonstration tests of FBR innovative technology development

    International Nuclear Information System (INIS)

    Yoshimi, H.; Hachiya, Y.

    1990-01-01

    A plan is under way at PNC to modify the experimental fast reactor JOYO. The project is called MARK-III (MK-III) program. The purpose of MK-III is to expand the function of JOYO, and to make it possible to receive demonstration tests of new or high level technologies for FBR development. The MK-III program consists of two main modifications: conversion to a highly efficient irradiation facility; and a modification for demonstration testing of new technologies and concepts that have a high potential to reduce FBR plant construction cost, to evaluate plant reliability and to improve plant safety. These modifications are scheduled to start in 1991

  12. Method for selecting FBR development strategies in the presence of uncertainty

    International Nuclear Information System (INIS)

    Fraley, D.W.; Burnham, J.B.

    1981-12-01

    This report describes the methods used to probabilistically analyze data related to the uranium supply the FBR's competitive dates, development strategies' time and costs, and economic benefits. It also describes the econometric methods used to calculate the economic risks of mistiming the development. Seven strategies for developing the FBR are analyzed. The various measures of a strategy's performance - timing, costs, benefits, and risks - are combined into several criteria which are used to evaluate the seven strategies. Methods are described for selecting a strategy based on a number of alternative criteria

  13. The environment effect on creep fatigue strength for FBR high temperature structural material, Type 304 steel

    International Nuclear Information System (INIS)

    Matsubara, Masaaki; Nitta, Akito

    1988-01-01

    In order to rationalize FBR high temperature structural design, the creep fatigue strength of Type 304 stainless steel for FBR main vessel was investigated in air and vacuum. The results obtained were as follows: Independent of strain wave forms, creep fatigue lives in vacuum were longer than those in air, and especially strain wave forms of fatigue damage type had noticeably longer lives. Also, a tendency to have too much longer lives at low strain level being important for real plant condition was shown. At last, it was confirmed that cyclic deformation behavior in vacuum was coincident with that in air. (author)

  14. Analytical solutions for the temperature field in a 2D incompressible inviscid flow through a channel with walls of solid fuel

    Directory of Open Access Journals (Sweden)

    Sorin BERBENTE

    2011-12-01

    Full Text Available A gas (oxidizer flows between two parallel walls of solid fuel. A combustion is initiated: the solid fuel is vaporized and a diffusive flame occurs. The hot combustion products are submitted both to thermal diffusion and convection. Analytical solutions can be obtained both for the velocity and temperature distributions by considering an equivalent mean temperature where the density and the thermal conductivity are evaluated. The main effects of heat transfer are due to heat convection at the flame. Because the detailed mechanism of the diffusion flame is not introduced the reference chemical reaction is the combustion of premixed fuel with oxidizer in excess. In exchange the analytical solution is used to define an ideal quasi-uniform combustion that could be realized by an n adequate control. The given analytical closed solutions prove themselves flexible enough to adjust the main data of some existing experiments and to suggest new approaches to the problem.

  15. Part 5. Fuel cycle options

    International Nuclear Information System (INIS)

    Lineberry, M.J.; McFarlane, H.F.; Amundson, P.I.; Goin, R.W.; Webster, D.S.

    1980-01-01

    The results of the FBR fuel cycle study that supported US contributions to the INFCE are presented. Fuel cycle technology is reviewed from both generic and historical standpoints. Technology requirements are developed within the framework of three deployment scenarios: the reference international, the secured area, and the integral cycle. Reprocessing, fabrication, waste handling, transportation, and safeguards are discussed for each deployment scenario. Fuel cycle modifications designed to increase proliferation defenses are described and assessed for effectiveness and technology feasibility. The present status of fuel cycle technology is reviewed and key issues that require resolution are identified

  16. Validation of the analytical methods in the LWR code BOXER for gadolinium-loaded fuel pins

    International Nuclear Information System (INIS)

    Paratte, J.M.; Arkuszewski, J.J.; Kamboj, B.K.; Kallfelz, J.M.; Abdel-Khalik, S.I.

    1990-01-01

    Due to the very high absorption occurring in gadolinium-loaded fuel pins, calculations of lattices with such pins present are a demanding test of the analysis methods in light water reactor (LWR) cell and assembly codes. Considerable effort has, therefore, been devoted to the validation of code methods for gadolinia fuel. The goal of the work reported in this paper is to check the analysis methods in the LWR cell/assembly code BOXER and its associated cross-section processing code ETOBOX, by comparison of BOXER results with those from a very accurate Monte Carlo calculation for a gadolinium benchmark problem. Initial results of such a comparison have been previously reported. However, the Monte Carlo calculations, done with the MCNP code, were performed at Los Alamos National Laboratory using ENDF/B-V data, while the BOXER calculations were performed at the Paul Scherrer Institute using JEF-1 nuclear data. This difference in the basic nuclear data used for the two calculations, caused by the restricted nature of these evaluated data files, led to associated uncertainties in a comparison of the results for methods validation. In the joint investigations at the Georgia Institute of Technology and PSI, such uncertainty in this comparison was eliminated by using ENDF/B-V data for BOXER calculations at Georgia Tech

  17. A strategy analysis of the fast breeder reactor introduction and nuclear fuel cycle systems deployment

    International Nuclear Information System (INIS)

    Wajima, Tsunetaka; Kawashima, Katsuyuki; Yamashita, Takashi

    1996-01-01

    A study is made on a strategy analysis of the long term nuclear fuel cycle systems deployment in accordance with the nuclear power growth projection and fast breeder reactor (FBR) introduction. In the analysis, the reprocessed plutonium (Pu) is charged into the reactor in such a way that the reprocessed Pu is not stored outside the reactor, i.e., there is no excess Pu outside the reactor. The analysis characterized the fuel cycle systems, and showed the usefulness of the present method to determine future directions for the FBR introduction and nuclear fuel cycle systems deployment. Concerning an intermediate-term strategy, the time of introduction and required capacities of a second commercial LWR reprocessing plant, Pu-thermal, and the first FBR reprocessing plant deployment are evaluated. A long term strategy analysis shows that the two or three large plants are run in parallel for each fuel cycle facility and that FBR related facilities deal with a markedly large amount of Pu. It is concluded that the early stage introduction of FBRs of significant capacities seems necessary to materialize a consistent total FBR/fuel cycle system where Pu balance becomes feasible through its flexible operation of, for instance, adjusting breeding ratio, in order to keep the transparency of the Pu utilization. (author)

  18. Urban transportation energy conservation: analytic procedures for estimating changes in travel demand and fuel consumption. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Atherton, T.J.; Suhrbier, J.H.

    1979-10-01

    This series of reports provides metropolitan planning organizations with analytical tools that can be used to evaluate the effectiveness of alternative transportation policies in achieving reductions in overall fuel consumption. To ensure a high measure of accuracy, the analysis goes beyond the first order effects, i.e., the shift from single occupant autos as the mode chosen for the work trip to more fuel efficient means of travel. Questions treated include what will happen with the autos left at home as a result of increased carpooling for work trips. Will certain policies, such as gasoline price increases, directly impact non-work tripmaking. Will a particular transportation policy affect all segments of the population, or will certain groups be impacted significantly more than others. The methodology developed links together several disaggregate travel demand models to predict auto ownership, work trip mode choice, and non-work travel demands. This report introduces the theoretical basis for the travel demand models used, describes these models and their linkages both with each other and with the various submodels, and documents the assumptions made in developing the model system and using it to forecast responses to alternative transportation policies. Emphasis is placed on the conceptual framework of the model system and specification of the individual models and submodels.

  19. Analytical and numerical study of radiation effect up to high burnup in power reactor fuels

    International Nuclear Information System (INIS)

    Lemes, M; Denis, A; Soba, A

    2012-01-01

    In the present work the behavior of fuel pellets for power reactors in the high burnup range (average burnup higher than 50 MWd/kgHM) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup, as long as a new microstructure develops, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behaviour. The evolution of porosity in the high burnup structure (HBS) is assumed to be determinant of the retention capacity of the fission gases released by the matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Starting from several works published in the open literature, a model was developed to describe the behaviour and evolution of porosity at local burnup values ranging from 60 to 300 MWd/KgHM. The model is mathematically expressed by a system of non-linear differential equations that take into account the open and closed porosity, the interactions between pores and the free surface and phenomena like pore's coalescence and migration and gas venting. Interactions of different orders between open and closed pores, growth of pores radius by vacancies trapping, the evolution of the pores number density, the internal pressure and over pressure within the pores, the fission gas retained in the matrix and released to the free volume are analyzed. The results of the simulations performed in the present work are in excellent agreement with experimental data available in the open literature and with results calculated by other authors (author)

  20. An experimental and analytical study of fluid flow and critical heat flux in PWR fuel elements

    International Nuclear Information System (INIS)

    Bowditch, F.H.; Mogford, D.J.

    1987-02-01

    This report describes experiments that have been carried out at the Winfrith Establishment of the United Kingdom Atomic Energy Authority to determine the critical heat flux characteristics of pressurized water reactor fuel elements over an unusually wide range of coolant flow conditions that are relevant to both normal and fault conditions of reactor operation. The experiments were carried out in the TITAN loop using an electrically heated bundle of 25 rods of 9.5 mm diameter on a 12.7 mm pitch fitted with plain grids in order to provide a generic base for code validation. The fully tabulated experimental data for critical heat flux, pressure drop and sub-channel mixing are encompassed by ranges of pressure between 20 and 160 Bar, coolant flow between 150 and 3600 Kg/m 2 s, and coolant inlet temperature between 150 and 320 0 C. The results of the experiments are compared with predicted data based upon several established critical heat flux correlations. It is concluded that the extrapolation of some correlations to conditions beyond their intended range of application can lead to dangerous over estimates of critical heat flux, but the Winfrith WSC-2 and the EPRI NP-2609 correlations perform well over the whole data range and correlate all data with RMS errors of 9% and 6% respectively. (author)

  1. The current status of research and development concerning steam generator acoustic leak detection for the demonstration FBR plant

    International Nuclear Information System (INIS)

    Higuchi, Masahisa

    1990-01-01

    The Japan Atomic Power Co. (JAPC) started the research and development into Acoustic Leak Detection for the Demonstration FBR (D-FBR) plant in 1989. Acoustic Leak Detection is expected as a water leak detection system in the Steam Generator for the first D-FBR plant. JAPC is presently analyzing data on Acoustic Leak Detection in order to form some basic concepts and basic specifications about leak detection. Both low frequency types and high frequency types are selected as candidates for Acoustic Leak Detection. After a review of both types, either one will be selected for the D-FBT plant. A detailed Research and Development plan on Acoustic Leak Detection, which should be carried out prior to starting the construction of the D-FBR plant, is under review. (author). 3 figs, 2 tabs

  2. Some difference of concepts between design guideline for FBR base isolation system and aseismic design guideline of LWR in Japan

    International Nuclear Information System (INIS)

    Shibata, Heki

    1992-01-01

    This paper deals with the concept and the relation of 'the Base Isolation System and FBR' to the Safety Criteria and the Guideline of the Aseismic Design of LWR in Japan. The Central Research Institute of Electric Power Industries have been working for FBR last several years. The author has been contribute to their works, and this is one of the subjects. He described his own idea obtained through the cooperative work with CRIEPI. (author)

  3. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

  4. Method of locating a leaking fuel element in a fast breeder power reactor

    International Nuclear Information System (INIS)

    Honekamp, J.R.; Fryer, R.M.

    1976-01-01

    Failed fuel detection and location in an FBR is accomplished by mass spectrometric analysis of the cover gas for the 134 Xe/ 133 Xe ratio, and correlation with the theoretical ratio in fuel elements of known power level and burnup. (E.C.B.)

  5. Heat resistant/radiation resistant cable and incore structure test device for FBR type reactor

    International Nuclear Information System (INIS)

    Tanimoto, Hajime; Shiono, Takeo; Sato, Yoshimi; Ito, Kazumi; Sudo, Shigeaki; Saito, Shin-ichi; Mitsui, Hisayasu.

    1995-01-01

    A heat resistant/radiation resistant coaxial cable of the present invention comprises an insulation layer, an outer conductor and a protection cover in this order on an inner conductor, in which the insulation layer comprises thermoplastic polyimide. In the same manner, a heat resistant/radiation resistant power cable has an insulation layer comprising thermoplastic polyimide on a conductor, and is provided with a protection cover comprising braid of alamide fibers at the outer circumference of the insulation layer. An incore structure test device for an FBR type reactor comprises the heat resistant/radiation resistant coaxial cable and/or the power cable. The thermoplastic polyimide can be extrusion molded, and has excellent radiation resistant by the extrusion, as well as has high dielectric withstand voltage, good flexibility and electric characteristics at high temperature. The incore structure test device for the FBR type reactor of the present invention comprising such a cable has excellent reliability and durability. (T.M.)

  6. Developing maintenance technologies for FBR's heat exchanger units by advanced laser processing

    International Nuclear Information System (INIS)

    Nishimura, Akihiko; Shimada, Yukihiro

    2011-01-01

    Laser processing technologies were developed for the purpose of maintenance of FBR's heat exchanger units. Ultrashort laser processing fabricated fiber Bragg grating sensor for seismic monitoring. Fiber laser welding with a newly developed robot system repair cracks on inner wall of heat exchanger tubes. Safety operation of the heat exchanger units will be improved by the advanced laser processing technologies. These technologies are expected to be applied to the maintenance for the next generation FBRs. (author)

  7. Formulae for evaluating buckling strength of FBR main vessels under earthquake loading

    International Nuclear Information System (INIS)

    Matsuura, Shinichi; Nakamura, Hideharu; Ogiso, Seitaro; Murakami, Toshiaki; Kawamoto, Youji.

    1991-01-01

    Although plastic shear-bending buckling of cylindrical part of FBR main vessels under horizontal earthquake loading is one of the most critical problems in the structural design, the evaluation method of the buckling strength is not specified in related standards in Japan. Central Research Institute of Electric Power Industry (CRIEPI), commissioned by the Ministry of International Trade and Industry of the Japanese Government, is carrying out verification tests of fast breeder technologies (from 1987 thru 1993 FY). The Demonstration Test and Research Program of Buckling of FBR is made a part of the commissioned research program and was finished the first half after establishing a seismic buckling design guideline (a tentative draft) in 1989 FY. The purpose of this paper is to describe the results of static buckling tests and numerical analyses, and to present formulae for determining the buckling strength of FBR main vessels. According to the understanding that shear-bending interaction effect is not so large, the elastic shear and bending buckling formulae and plasticity reduction by φ-method are presented individually. Regarding to the shear buckling, shape imperfection within plate thickness is considered in the formulae originally, and the correction factor to larger imperfection is provided. (author)

  8. Analytical Evaluation to Determine Selected PAHs by HPLC in a Type 2 Fuel; Evaluacion Analitica de 4 Metodos de Determinacion de PAHs medianteHPLC en un Fuel de Tipo II

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Escolano Segovia, O.; Garcia Frutos, F. J.

    2009-05-21

    An evaluation of analytical parameters to determine selected PAHs in a fuel oil type II by HPLC coupled to fluorescence and diode detectors is presented. The study was focused on four conventional treatments of these kinds of oil samples and the main objective was giving a measure of confidence level of PAH results in the fuel oil. This study was performed in the frame of the project Assessment of natural attenuation of PAHs in agricultural soil contaminated with fuel from an accidental spill (Spanish National Plain I+D+I, CTM2007-64537). This paper is presented as follows: Analysis of reference material 1582 (NIST) by using the four kinds of sample treatments of interest. Application of variance analysis to compare results obtained from type II fuel by using each sample treatment and chromatographic detector. Finally, a statistic calculation was performed to measure uncertainty components in chromatographic analysis. (Author)

  9. Analytical methods for fissionable material determinations in the nuclear fuel cycle. Progress report, October 1, 1976--September 30, 1977

    International Nuclear Information System (INIS)

    Waterbury, G.R.

    1978-01-01

    Development of dissolution techniques for difficult-to-dissolve nuclear materials, development of methods and automated instruments for plutonium and uranium determinations, preparation of plutonium-containing materials for the Safeguards Analytical Laboratory Evaluation (SALE) program, analysis of SALE uranium materials, preparation of certified reference material plutonium metal, measurement of longer plutonium isotope half-lives, and study of ion exchange behavior of elements in various media continued. Gas-solid reaction of carbonyl chloride with uranium-bearing materials at elevated temperature is superior to reaction with chlorine for uranium volatilization and separation. Neither reaction with a variety of nonaqueous solvents nor reaction with molten selenium oxide provides practical dissolution of refractory materials characteristic of nuclear fuel cycle materials. The LASL automated spectrophotometer has been used to determine 0.1-mg amounts without instrumental or procedural changes. A microgram-sensitive spectrophotometric method for uranium has been developed, and the automated spectrophotometer is being modified to its use. A controlled-potential coulometric method has been developed for selective determination of plutonium. An automated analyzer to use this method is being built. Uranium-plutonium mixed oxide powder, for SALE samples, has not remained stable during storage, but high-density pellets have. In a DOE interlaboratory program, the half-life of 239 Pu has been measured, experiments on 241 Pu half-life measurement are in progress, and 240 Pu half-life measurement is planned. Ion exchange distributions for over 50 elements have been measured to determine cation exchange in nitric acid and anion exchange in both hydrobromic and hydriodic acids

  10. Analytical Evaluation of Preliminary Drop Tests Performed to Develop a Robust Design for the Standardized DOE Spent Nuclear Fuel Canister

    International Nuclear Information System (INIS)

    Ware, A.G.; Morton, D.K.; Smith, N.L.; Snow, S.D.; Rahl, T.E.

    1999-01-01

    The Department of Energy (DOE) has developed a design concept for a set of standard canisters for the handling, interim storage, transportation, and disposal in the national repository, of DOE spent nuclear fuel (SNF). The standardized DOE SNF canister has to be capable of handling virtually all of the DOE SNF in a variety of potential storage and transportation systems. It must also be acceptable to the repository, based on current and anticipated future requirements. This expected usage mandates a robust design. The canister design has four unique geometries, with lengths of approximately 10 feet or 15 feet, and an outside nominal diameter of 18 inches or 24 inches. The canister has been developed to withstand a drop from 30 feet onto a rigid (flat) surface, sustaining only minor damage - but no rupture - to the pressure (containment) boundary. The majority of the end drop-induced damage is confined to the skirt and lifting/stiffening ring components, which can be removed if de sired after an accidental drop. A canister, with its skirt and stiffening ring removed after an accidental drop, can continue to be used in service with appropriate operational steps being taken. Features of the design concept have been proven through drop testing and finite element analyses of smaller test specimens. Finite element analyses also validated the canister design for drops onto a rigid (flat) surface for a variety of canister orientations at impact, from vertical to 45 degrees off vertical. Actual 30-foot drop testing has also been performed to verify the final design, though limited to just two full-scale test canister drops. In each case, the analytical models accurately predicted the canister response

  11. Analytical methods for fissionable material determinations in the nuclear fuel cycle. Progress report, October 1, 1976--September 30, 1977

    Energy Technology Data Exchange (ETDEWEB)

    Waterbury, G.R. (comp.)

    1978-01-01

    Development of dissolution techniques for difficult-to-dissolve nuclear materials, development of methods and automated instruments for plutonium and uranium determinations, preparation of plutonium-containing materials for the Safeguards Analytical Laboratory Evaluation (SALE) program, analysis of SALE uranium materials, preparation of certified reference material plutonium metal, measurement of longer plutonium isotope half-lives, and study of ion exchange behavior of elements in various media continued. Gas-solid reaction of carbonyl chloride with uranium-bearing materials at elevated temperature is superior to reaction with chlorine for uranium volatilization and separation. Neither reaction with a variety of nonaqueous solvents nor reaction with molten selenium oxide provides practical dissolution of refractory materials characteristic of nuclear fuel cycle materials. The LASL automated spectrophotometer has been used to determine 0.1-mg amounts without instrumental or procedural changes. A microgram-sensitive spectrophotometric method for uranium has been developed, and the automated spectrophotometer is being modified to its use. A controlled-potential coulometric method has been developed for selective determination of plutonium. An automated analyzer to use this method is being built. Uranium-plutonium mixed oxide powder, for SALE samples, has not remained stable during storage, but high-density pellets have. In a DOE interlaboratory program, the half-life of /sup 239/Pu has been measured, experiments on /sup 241/Pu half-life measurement are in progress, and /sup 240/Pu half-life measurement is planned. Ion exchange distributions for over 50 elements have been measured to determine cation exchange in nitric acid and anion exchange in both hydrobromic and hydriodic acids.

  12. Progress in researches on MOX fuel pellet producing technology in China

    International Nuclear Information System (INIS)

    Hu Xiaodan

    2010-01-01

    Being the key section of nuclear-fuel cycle, the producing technology of MOX(UO 2 -PuO 2 ) fuel had driven to maturity in France, England, Russia, Belgium, etc. MOX fuel had been applied in FBR and LWR successfully in those countries. With the rapidly developing of nuclear-generated power, the MOX fuel for FBR and LWR was active demanded in China. However, the producing technology of MOX fuel developed slowly. During the period of 'the seventh five year's project', MOX fuel pellet was produced by mechanically mixed method and oxalate deposited method, respectively. Parts of cool performance of MOX fuel pellet produced by oxalate deposited method reached the qualification of fuel for FBR. During the period of 'the ninth five year's project' and 'the tenth five year's project', the technical route of producing MOX fuel was determined, and the test line of producing MOX fuel was built preliminarily. In the same time, the producing technology and analyzing technology of MOX fuel pellet by mechanically mixed was studied roundly, and the representative analogue pellet(UO 2 -CeO 2 ) was produced. That settled the supporting technology for the commercial process and research of MOX fuel rod and MOX fuel module. (authors)

  13. In pile testing program of mixed oxide fuels fabricated under the new GSP method

    International Nuclear Information System (INIS)

    Marinucci, G.; Nobili, A.; Lanchi, M.; Caracchini, R.; Dupont, G.; Galtier, J.

    1983-01-01

    ENEA, AGN and CEA have collaborated in developing and effectuating the new fuel fabrication technique GSP. In that framework, three irradiation experiments in the SILOE reactor are projected for studying certain characteristics of GSP fuel in comparison with standard fuels (type FBR). The three experiments facilitate the analysis of the thermal behaviour and the fuel pin deformations up to a high burnup. In this paper, the experimental devices are described (5 designs) and the irradiation conditions are briefly discussed. (G.J.P.)

  14. Determination of Reaction Mechanisms Occurring at Fuel Cell Electrocatalysts Using Electrochemical Methods, Spectroelectrochemical Measurements and Analytical Techniques

    Science.gov (United States)

    Coutanceau, C.; Baranton, S.; Lamy, C.

    There is now a great interest in developing different kinds of fuel cells for several applications (stationary electric power plants, transportation, portable electronic devices). For many applications, hydrogen is the most convenient fuel, but it is not a primary fuel, so that it has to be produced from different sources: water, fossil fuels (natural gas, hydrocarbons, etc.), biomass resources, etc. When produced from fossil fuel and biomass resources, hydrogen gas contains a non negligible amount of CO, which acts as a poisoning species for platinum electrocatalysts. Other fuels, particularly alcohols, which are liquid under ambient temperature and pressure, are more convenient due to the easiness of their handling and distribution and high theoretical energy density (6 to 8 kWh kg-1, for methanol and ethanol, respectively). Direct Methanol Fuel Cells (DMFCs) and Direct Ethanol Fuel Cells (DEFCs) are based on the Proton Exchange Membrane Fuel Cell (PEMFC) system, in which hydrogen is replaced by the alcohol. Moreover, due to the presence of carbon monoxide, the issues for PEMFCs working with reformate gas are close to those met in Direct Alcohol Fuel Cells (DAFCs), where the dissociative adsorption of alcohol leads to the formation of adsorbed CO species.

  15. The use of spectrophotometry in FBR reprocessing analysis

    International Nuclear Information System (INIS)

    Brown, M.L.; Mills, C.L.; Kyffin, T.W.

    1986-09-01

    The spectrophotometric methods of analysis currently in use at DNPDE are described. It considers the ways in which the problems of containment and physical handling of active solutions have been overcome, and summarises performance of the methods during several PFR fuel reprocessing campaigns. The introduction of a new micro-computer controlled fibre-optic spectrophotometer is considered in terms of its advantages over the existing systems, both in safe sample handling and computational abilities. Its performance is compared with existing methods. Finally, a novel system for measurement of plutonium valency, and americium in plutonium using a ''spectral stripping'' technique, is discussed. The results of this method are compared with those obtained using conventional techniques. (author)

  16. Analytical Evaluation to Determine Selected PAHs in a Contaminated Soil With Type II Fuel; Metodo Optimizado de Extraccion por Ultrasonidos para la Determinacion de PAHs Seleccionados en un Suelo Contaminado con Fuel de Tipo II

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Garcia Frutos, F. J.

    2010-10-21

    A study on the optimization of an ultrasonic extraction method for selected PAHs determination in soil contaminated by type II fuel and by using HPLC with fluorescence detector is presented. The main objective was optimize the analytical procedure, minimizing the volume of solvent and analysis time and avoiding possible loss by evaporation. This work was carried out as part of a project that investigated a remediation process of agricultural land affected by an accidental spillage of fuel (Plan Nacional I + D + i, CTM2007-64 537). The paper is structured as: Optimization of wavelengths in the chromatographic conditions to improve resolution in the analysis of fuel samples. Optimization of the main parameters affecting in the extraction process by sonication. Comparison of results with those obtained by accelerated solvent extraction. (Author) 3 refs.

  17. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

    International Nuclear Information System (INIS)

    Reyes F, M. C.; Del Valle G, E.; Filio L, C.

    2013-10-01

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  18. The Value of Renewable Energy as a Hedge Against Fuel Price Risk: Analytic Contributions from Economic and Finance Theory

    Energy Technology Data Exchange (ETDEWEB)

    Bolinger, Mark A; Wiser, Ryan

    2008-09-15

    For better or worse, natural gas has become the fuel of choice for new power plants being built across the United States. According to the Energy Information Administration (EIA), natural gas-fired units account for nearly 90% of the total generating capacity added in the U.S. between 1999 and 2005 (EIA 2006b), bringing the nationwide market share of gas-fired generation to 19%. Looking ahead over the next decade, the EIA expects this trend to continue, increasing the market share of gas-fired generation to 22% by 2015 (EIA 2007a). Though these numbers are specific to the US, natural gas-fired generation is making similar advances in many other countries as well. A large percentage of the total cost of gas-fired generation is attributable to fuel costs--i.e., natural gas prices. For example, at current spot prices of around $7/MMBtu, fuel costs account for more than 75% of the levelized cost of energy from a new combined cycle gas turbine, and more than 90% of its operating costs (EIA 2007a). Furthermore, given that gas-fired plants are often the marginal supply units that set the market-clearing price for all generators in a competitive wholesale market, there is a direct link between natural gas prices and wholesale electricity prices. In this light, the dramatic increase in natural gas prices since the 1990s should be a cause for ratepayer concern. Figure 1 shows the daily price history of the 'first-nearby' (i.e., closest to expiration) NYMEX natural gas futures contract (black line) at Henry Hub, along with the futures strip (i.e., the full series of futures contracts) from August 22, 2007 (red line). First, nearby prices, which closely track spot prices, have recently been trading within a $7-9/MMBtu range in the United States and, as shown by the futures strip, are expected to remain there through 2012. These price levels are $6/MMBtu higher than the $1-3/MMBtu range seen throughout most of the 1990s, demonstrating significant price escalation for

  19. Advanced fuel for fast breeder reactors: Fabrication and properties and their optimization

    International Nuclear Information System (INIS)

    1988-06-01

    The present design for FBR fuel rods includes usually MOX fuel pellets cladded into stainless steel tubes, together with UO 2 axial blanket and stainless steel hexagonal wrappers. Mixed carbide, nitride and metallic fuels have been tested as alternative fuels in test reactors. Among others, the objectives to develop these alternative fuels are to gain a high breeding ratio, short doubling time and high linear ratings. Fuel rod and assembly designers are now concentrating on finding the combination of optimized fuel, cladding and wrapper materials which could result in improvement of fuel operational reliability under high burnups and load-follow mode of operation. The purpose of the meeting was to review the experience of advanced FBR fuel fabrication technology, its properties before, under and after irradiation, peculiarities of the back-end of the nuclear fuel cycle, and to outline future trends. As a result of the panel discussion, the recommendations on future Agency activities in the area of advanced FBR fuels were developed. A separate abstract was prepared for each of the 10 presentations of this meeting. Refs, figs and tabs

  20. Analytical chemistry instrumentation

    International Nuclear Information System (INIS)

    Laing, W.R.

    1986-01-01

    Separate abstracts were prepared for 48 papers in these conference proceedings. The topics covered include: analytical chemistry and the environment; environmental radiochemistry; automated instrumentation; advances in analytical mass spectrometry; Fourier transform spectroscopy; analytical chemistry of plutonium; nuclear analytical chemistry; chemometrics; and nuclear fuel technology

  1. Analytical chemistry instrumentation

    International Nuclear Information System (INIS)

    Laing, W.R.

    1986-01-01

    In nine sections, 48 chapters cover 1) analytical chemistry and the environment 2) environmental radiochemistry 3) automated instrumentation 4) advances in analytical mass spectrometry 5) fourier transform spectroscopy 6) analytical chemistry of plutonium 7) nuclear analytical chemistry 8) chemometrics and 9) nuclear fuel technology

  2. LBB assessment on ferrite piping structure of large-scale FBR

    OpenAIRE

    兪 淵植

    2002-01-01

    These days, this interest on LBB(Leak before Break) design becomes to be rising in the viewpoint of the cost reduction and structural inter-grity for the commercialization of FBR plants, LBB design enables pla-nts to be shut down safely before occuring unstable fracture by dete- cting the leak rates even if a crack initiates and penetrates a wall thickness. It is necessary to assess crack growth and penetration be- havior considering in-service conditions under operation temperature, leak re...

  3. Sodium leakage experience at the prototype FBR Monju

    International Nuclear Information System (INIS)

    Miyakawa, A.; Maeda, H.; Kani, Y.; Ito, K.

    2000-01-01

    Monju is Japan's prototype fast breeder reactor: 280 MWe (714 MWt), fueled with mixed oxides of plutonium and uranium, cooled by liquid sodium. Construction was started in 1985 and initial criticality was attained in April 1994. On 8th December 1995, sodium leakage from a secondary circuit occurred in a piping room of the reactor auxiliary building. The secondary sodium leaked through a temperature sensor, due to the breakaway of the tip of the thermocouple well tube installed near the secondary circuit outlet of the intermediate heat exchanger (IHX). The reactor remained cooled and thus, from the viewpoint of radiological hazards, the safety of the reactor was secured. There was no release of radioactive material. There were no adverse effects for personnel and the surrounding environment. The thermocouple well tube failure resulted from high cycle fatigue due to flow induced vibration. It was found that this flow induced vibration was not caused by well-known Von Karman vortex shedding, but a symmetric vortex shedding. The design of the thermocouple well, which was subject to avoid this phenomenon, was reviewed. A new design guide against the flow-induced vibration was prepared by JNC (Japan Nuclear Cycle Development Institute). This is more comprehensive and definitive than the existing guide 'ASME N-1300' (Flow-induced vibration of tube and tube banks). New thermocouple well designs were proposed consistent with this design guide. To prevent a recurrence of the secondary sodium leakage incident, comprehensive design review activities were started for the purpose of checking the safety and reliability of the plant. As a result, several aspects to be improved were identified and improvements and countermeasures have been studied. The main improvements and countermeasures are as follows: To enable the operators to understand and react to incidents quickly, new sodium leakage detectors (TV monitors, smoke sensors) and a new surveillance system will be installed; To

  4. Spent fuel management in Japan

    International Nuclear Information System (INIS)

    Shirahashi, K.; Maeda, M.; Nakai, T.

    1996-01-01

    Japan has scarce energy resources and depends on foreign resources for 84% of its energy needs. Therefore, Japan has made efforts to utilize nuclear power as a key energy source since mid-1950's. Today, the nuclear energy produced from 49 nuclear power plants is responsible for about 31% of Japan's total electricity supply. The cumulative amount of spent fuel generated as of March 1995 was about 11,600 Mg U. Japan's policy of spent fuel management is to reprocess spent nuclear fuel and recycle recovered plutonium and uranium as nuclear fuel. The Tokai reprocessing plant continues stable operation keeping the annual treatment capacity or around 90 Mg U. A commercial reprocessing plant is under construction at Rokkasho, northern part of Japan. Although FBR is the principal reactor to use plutonium, LWR will be a major power source for some time and recycling of the fuel in LWRs will be prompted. (author). 3 figs

  5. Evolution of FBR surveillance using a noise analysis and an on-line signal processing

    International Nuclear Information System (INIS)

    Trapp, J.P.; Lebrun, A.; Lhuillier, C.; Berger, R.; Martin, L.

    1996-01-01

    In a fast breeder reactor with a liquid sodium cooling as in the French nuclear type, the main goal of the surveillance and protection systems is to warrant the reactor integrity against every kind of incidents or accidents threatening the reactor, personnel and environment safety. In the future, the surveillance systems could be used to prevent all the events which could lead to an incident. In this way, a large number of neutronic detectors (fission chambers, boron counters), acoustic (sensors with wave guide), thermal (sodium-steel thermocouple) and vibratory sensors have been implemented in and out the primary vessel of these reactors. After a brief presentation of the standard surveillance and protection systems is given, with a recall of the main results obtained in France in the FBR noise analysis range is given. We show the results recently obtained at Phenix and Superphenix in the case of neutronic, thermal, acoustic and vibratory noises. In addition, we present the ALPES system, recently implemented, on line, at SPX with the goal to improve the core thermal surveillance; numeric processing of the thermocouples at the subassembly outlet are used. We show the main results and the improvements in the domain of the local accidents surveillance. Finally, we summarize the possibilities of an on-line noise analysis system for the early detection of FBR abnormalities. (authors)

  6. Analysis of fluid-structure interaction mechanism of a Na-FBR core while the evacuation of a gas pocket

    International Nuclear Information System (INIS)

    Sargentini, Lucia

    2014-01-01

    The purpose of this study is to improve the knowledge about the core behavior of a sodium fast breeder reactor (Na-FBR) during vibrations through the fluid-structure interaction analysis. Namely, we investigate the flowering of the Phenix core during the SCRAM for negative reactivity (AURN) and the seismic behavior of the core of Astrid project. Three approaches are followed: experimental campaign, performing of analytical solution and development of numerical model. We create a flow regime map to identify the flow regimes in the fluid gap for very short times scales (as AURN) as well as longer time scales (as seismic oscillations). The most suitable equation system (Navier-Stokes, Euler or linearized Euler) is chosen to model the fluid flow in the numerical code. To our knowledge, for the first time, an analytical solution for free vibration and very narrow gaps is proposed. We designed two experimental apparatus (PISE-1a and PISE-2c) composed respectively by 1 and 19 hexagonal assemblies (two crowns) of Poly-methyl methacrylate (PMMA). Every PMMA assembly is fixed to a stainless steel twin-blades support allowing only orthogonal oscillations with respect to generating line of assembly. The twin-blades supports are designed to give the same range frequency of Phenix assembly in liquid sodium. The experimental equipment PISE-1a is used to determine the dynamic characteristics of PISE-2c assembly, to calibrate instrumentation and for validating our numerical model. Free vibration tests in air are performed to evaluate the dynamic characteristics of the body. Free vibration experiments in water allow to assess the added mass and added damping effect on the frequency. Even though the fluid flow during vibration should be completely bidimensional, the fluid flow is affected by a 3D effect - named 'jambage' - at the top and the basis of the assembly. This effect produces a lower frequency than the theoretical value. Tests are modeled with a bidimensional

  7. Summarizing documentation of the laboratory automation system RADAR for the analytical services of a nuclear fuel reprocessing facility

    International Nuclear Information System (INIS)

    Brandenburg, G.; Brocke, W.; Brodda, B.G.; Buerger, K.; Halling, H.; Heer, H.; Puetz, K.; Schaedlich, W.; Watzlawik, K.H.

    1981-12-01

    The essential tasks of the system are on-line open-loop process control based on in-line measurements and automation of the off-line analytical laboratory. The in-line measurements (at 55 tanks of the chemical process area) provide density-, liquid-, level-, and temperature values. The concentration value of a single component may easily be determined, if the solution consists of no more than two phases. The automation of the off-line analytical laboratory contains laboratory organization including sample management and data organization and computer-aided sample transportation control, data acquisition and data processing at chemical and nuclear analytical devices. The computer system consists of two computer-subsystems: a front end system for sample central registration and in-line process control and a central size system for the off-line analytical tasks. The organization of the application oriented system uses a centralized data base. Similar data processing functions concerning different analytical management tasks are structured into the following subsystem: man machine interface, interrupt- and data acquisition system, data base, protocol service and data processing. The procedures for the laboratory management (organization and experiment sequences) are defined by application data bases. Following the project phases, engineering requirements-, design-, assembly-, start up- and test run phase are described. In addition figures on expenditure and experiences are given and the system concept is discussed. (orig./HP) [de

  8. An investigation on technical feasibilities of fuel cycle for high temperature gas-cooled reactor (Case study)

    International Nuclear Information System (INIS)

    Sumita, Junya; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sawa, Kazuhiro

    2008-03-01

    In accordance with the basic policy of effectively using nuclear fuel resources, the FBR cycle, one of the most possible fuel cycle in the future, will be adapted after plu-thermal program by LWR in Japanese nuclear cycle plan. In this paper, a case study of technical investigation of HTGR fuel cycle based on HTGR fuel cycle proposed to adapt to Japanese nuclear fuel cycle plan were carried out from the viewpoint of effective utilization of uranium, fabrication technologies of MOX fuel, reprocessing technologies, amount of interim storage of HTGR fuel and graphite waste. As a result, the fuel cycle for HTGR is expected to be possible technically. (author)

  9. Dynamic model of oxygen starved proton exchange membrane fuel-cell using hybrid analytical-numerical method

    Science.gov (United States)

    Vijayaraghavan, Krishna; DeVaal, Jake; Narimani, Mohammad

    2015-07-01

    One of the primary life-limiting factors in PEM fuel-cells arises from performance degradation resulting from transfer (crossover) leaks. Transfer leaks result in oxygen starvation and models of fuel cells under oxygen starved conditions would allow for detection of fault inception. This paper develops a unified fuel-cell model for when the fuel-cells can either deliver power (termed driving-mode, and for when the cell absorb power (termed driven-mode) for higher leak rates. The model captures the gradient of the reactants both in the GDL and in the flow channel in addition to capturing the various electro-chemical effects. The response of the model under normal conditions is first validated for normal operation against previously published experiments. The response of the model under oxygen-starved conditions is then validated against simulated leaks in three different cell architectures: a Ballard 9-cell Mk1100 stack where hydrogen is injected into one cell, and a Ballard 10-cell Mk902 stack and 20-cell Mk903 stack where hydrogen is injected into the upstream cathode flow. Finally, the response of the model is also validated against an actual leaky Mk902 cell. The model generally agrees well with the measured cell voltage data for all the above experiments.

  10. DEVELOPMENT OF SAMPLING AND ANALYTICAL METHODS FOR THE MEASUREMENT OF NITROUS OXIDE FROM FOSSIL FUEL COMBUSTION SOURCES

    Science.gov (United States)

    The report documents the technical approach and results achieved while developing a grab sampling method and an automated, on-line gas chromatography method suitable to characterize nitrous oxide (N2O) emissions from fossil fuel combustion sources. The two methods developed have...

  11. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  12. Technologists' challenge for independent development of nuclear fuel cycle

    International Nuclear Information System (INIS)

    2000-01-01

    Japan Nuclear Cycle Development Institute (JNC) was established on Oct. 1998 in cooperation of national and local governments concerned and private companies. This report outlines the activities to develop the technology of nuclear fuel cycle by Power Reactor Nuclear Fuel Development Corporation (PNC), which was reorganized to JNC. JNC is now effectively performing the research and the development of nuclear fuel cycle based on the basic principles defined by the national government concerned. First, 'Joyo', a fast breeder reactor (FBR) was constructed and put in operation in later sixties, leading to the criticality in 1977, whereas for 'Monju' it was in the later eighties. In Dec. 1995, however, an accident of sodium leak occurred in 'Monju'. Then, members selected from various fields discussed the strategy of the development of FBR. It was thus reconfirmed that FBR is the important choice as non-fossil energy in Japan. Study on strategy for practical utilization of FBR has started from 1999. Then, an advanced thermal reactor (ATR) called as 'Fugen' was constructed in Tsuruga City by Hitachi, Ltd. and the reactor reached to the criticality in 1978 for the first and to practical operation in 1979. ATR has contributed to the establishment of plutonium utilization system in Japan. Furthermore, PNC has attempted to make exploration development for uranium resource. After the exploration in Canada, Australia, Africa, etc., the amount of Japanese uranium possession reached 13% of uranium mined in the world for 1985-1994. Meanwhile 10% enriched uranium was successively produced in 1994. Development of plutonium fuel, MOX (mixed-oxide fuel) is now in progress. (M.N.)

  13. Simulation and experimental approach to CVD-FBR aluminide coatings on ferritic steels under steam oxidation

    International Nuclear Information System (INIS)

    Leal, J.; Alcala, G.; Bolivar, F.J.; Sanchez, L.; Hierro, M.P.; Perez, F.J.

    2008-01-01

    The ferritic steels used to produce structural components for steam turbines are susceptible to strong corrosion and creep damage due to the extreme working conditions pushed to increase the process efficiency and to reduce pollutants release. The response of aluminide coatings on the P-92 ferritic steel, deposited by CVD-FBR, during oxidation in a simulated steam environment was studied. The analyses were performed at 650 deg. C in order to simulate the working conditions of a steam turbine, and 800 deg. C in order to produce a critical accelerated oxidation test. The Thermo-Calc software was used to predict the different solid phases that could be generated during the oxidation process, in both, coated and uncoated samples. In order to validate the thermodynamic results, the oxides scales produced during steam tests were characterized by different techniques such as XRD, SEM and EDS. The preliminary results obtained are discussed in the present work

  14. Aluminum and aluminum/silicon coatings on ferritic steels by CVD-FBR technology

    International Nuclear Information System (INIS)

    Perez, F.J.; Hierro, M.P.; Trilleros, J.A.; Carpintero, M.C.; Sanchez, L.; Bolivar, F.J.

    2006-01-01

    The use of chemical vapor deposition by fluidized bed reactors (CVD-FBR) offers some advantages in comparison to other coating techniques such as pack cementation, because it allows coating deposition at lower temperatures than pack cementation and at atmospheric pressure without affecting the mechanical properties of material due to heat treatments of the bulk during coating process. Aluminum and aluminum/silicon coatings have been obtained on two different ferritics steels (P-91 and P-92). The coatings were analyzed using several techniques like SEM/EDX and XRD. The results indicated that both coatings were form by Fe 2 Al 5 intermetallic compound, and in the co-deposition the Si was incorporated to the Fe 2 Al 5 structure in small amounts

  15. Ethanol production from lignocellulosic biomass by recombinant Escherichia coli strain FBR5.

    Science.gov (United States)

    Saha, Badal; Cotta, Michael A

    2012-01-01

    Lignocellulosic biomass, upon pretreatment and enzymatic hydrolysis, generates a mixture of hexose and pentose sugars such as glucose, xylose, arabinose and galactose. While Escherichia coli utilizes all these sugars it lacks the ability to produce ethanol from them. Recombinant ethanologenic E. coli strains have been created with a goal to produce ethanol from both hexose and pentose sugars. Herein, we review the current state of the art on the production of ethanol from lignocellulosic hydrolyzates by an ethanologenic recombinant E. coli strain (FBR5). The bacterium is stable without antibiotics and can tolerate ethanol up to 50 gL(-1). It produces up to 45 g ethanol per L and has the potential to be used for industrial production of ethanol from lignocellulosic hydrolyzates.

  16. Simulation and experimental approach to CVD-FBR aluminide coatings on ferritic steels under steam oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Leal, J. [Universidad Complutense de Madrid, Dep. CC. Materiales e Ingenieria Metalurgica, Avenida Complutense s/n, Facultad de Ciencias Quimicas, 28040 Madrid (Spain); Alcala, G. [Universidad Complutense de Madrid, Dep. CC. Materiales e Ingenieria Metalurgica, Avenida Complutense s/n, Facultad de Ciencias Quimicas, 28040 Madrid (Spain)], E-mail: galcades@yahoo.es; Bolivar, F.J.; Sanchez, L.; Hierro, M.P.; Perez, F.J. [Universidad Complutense de Madrid, Dep. CC. Materiales e Ingenieria Metalurgica, Avenida Complutense s/n, Facultad de Ciencias Quimicas, 28040 Madrid (Spain)

    2008-07-15

    The ferritic steels used to produce structural components for steam turbines are susceptible to strong corrosion and creep damage due to the extreme working conditions pushed to increase the process efficiency and to reduce pollutants release. The response of aluminide coatings on the P-92 ferritic steel, deposited by CVD-FBR, during oxidation in a simulated steam environment was studied. The analyses were performed at 650 deg. C in order to simulate the working conditions of a steam turbine, and 800 deg. C in order to produce a critical accelerated oxidation test. The Thermo-Calc software was used to predict the different solid phases that could be generated during the oxidation process, in both, coated and uncoated samples. In order to validate the thermodynamic results, the oxides scales produced during steam tests were characterized by different techniques such as XRD, SEM and EDS. The preliminary results obtained are discussed in the present work.

  17. Equipment installation structure of roof slab for tank type FBR and method of equipment installation

    International Nuclear Information System (INIS)

    Sakai, Takao; Yamakawa, Masanori; Otsuka, Masaya; Sekine, Katsuhisa

    1986-01-01

    Purpose: To reduce equipment thermal stress and deformation by eliminating uneven temperature distribution caused at the equipment through section of the roof slab for the tank FBR, and at the same time, simplify the structure installation. Method: Multiple number of vertical fin projects are fit on the equipment through-section inside wall for the roof slab and the cylindrical equipment peripheral wall, and with these projected fins, the ring space of the through section is vertically divided into multiple sections in the circumferential direction. The vertical fins on the through-section inside wall and the fins on the equipment peripheral wall are contacted with each other by revolving them in the lateral direction. As a result, the natural convection caused by the difference of temperatures in the vertical direction of the ring space becomes a convection within each sector divided, and never generates circumferential circulation, which reduce uneven temperature distribution caused at the equipment through section. (Kawakami, Y.)

  18. Experimental and analytical investigation on the emission and combustion characteristics of CI engine fueled with tamanu oil methyl esters

    Directory of Open Access Journals (Sweden)

    Perumal Navaneetha Krishnan

    2016-01-01

    Full Text Available The emission and combustion characteristics of a four stroke multi fuel single cylinder variable compression ratio engine fueled with tamanu oil methyl ester and its blends 10%, 20%, 40%, and 60% with diesel (on volume basis are examined and compared with standard diesel. Biodiesel produced from tamanu oil by trans-esterification process has been used in this study. The experiment has been conducted at a constant engine speed of 1500 rpm with 50% load and at compression ratios of 16:1, 17:1, 18:1, 19:1, and 20:1. With different blend and for selected compression ratio the exhaust gas emissions such as CO, HC, NOx, CO2, and the combustion characteristics are measured. The variation of the emission parameters for different compression ratios and for different blends is given, and optimum compression ratio which gives best performance has been identified. The results indicate higher rate of pressure rise and minimum heat release rate at higher compression ratio for tamanu oil methyl ester when compared with standard diesel. The blend B40 for tamanu oil methyl ester is found to give minimum emission at 50% load. The blend when used as fuel results in reduction of polluting gases like HC, CO, and increase in NOx emissions. The previously mentioned emission parameters have been validated with the aid of artificial neural network. A separate model is developed for emission characteristics in which compression ratio, blend percentage and load percentage were used as the input parameter whereas CO, CO2, HC, and NOx were used as the output parameter. This study shows that there is a good correlation between the artificial neural network predicted values and the experimental data for different emission parameters.

  19. Analytical review on the hydrogen multilayer intercalation in carbonaceous nanostructures: relevance for development of super-adsorbents for fuel-cell-powered vehicles.

    Science.gov (United States)

    Nechaev, Yu S; Alexeeva, O K; Ochsner, A

    2009-06-01

    The analytical consideration of some recent experimental and theoretical data on the hydrogen on-board storage problem shows the necessity and economical expediency of carrying out further basic studies and initiating a constructive discussion on the physical key-note aspects ("open questions") of the hydrogen sorption by carbon-based nanomaterials: Especially, on the hydrogen multilayer intercalation in carbonaceous nanostructures, their relevance for the development of super-adsorbents for fuel-cell-powered vehicles, i.e., storage materials, which satisfy most of the U.S. DOE targets. It is consistent with the U. S. National Academies' recent recommendations and manifestations of the critical situation of the hydrogen storage problem.

  20. Comprehensive one-dimensional, semi-analytical, mathematical model for liquid-feed polymer electrolyte membrane direct methanol fuel cells

    Science.gov (United States)

    Kareemulla, D.; Jayanti, S.

    Polymer electrolyte membrane direct methanol fuel cells (PEM-DMFCs) have several advantages over hydrogen-fuelled PEM fuel cells; but sluggish methanol electrochemical oxidation and methanol crossover from the anode to the cathode through the PEM are two major problems with these cells. In the present work, a comprehensive one-dimensional, single phase, isothermal mathematical model is developed for a liquid-feed PEM-DMFC, taking into account all the necessary mass transport and electrochemical phenomena. Diffusion and convective effects are considered for methanol transport on the anode side and in the PEM, whereas only diffusional transport of species is considered on the cathode side. A multi-step reaction mechanism is used to describe the electrochemical oxidation of methanol at the anode. Stefan-Maxwell equations are used to describe multi-component diffusion on the cathode side and Tafel type of kinetics is used to describe the simultaneous methanol oxidation and oxygen reduction reactions at the cathode. The model fully accounts for the mixed potential effect caused by methanol crossover at the cathode. It shows excellent agreement with literature data of the limiting current density for different low methanol feed concentrations at different operating temperatures. At high methanol feed concentrations, oxygen depletion on the cathode side, due to excessive methanol crossover, results in mass-transport limitations. The model can be used to optimize the geometric and physical parameters with a view to extracting the highest current density while still keeping a tolerably low methanol crossover.

  1. MOX fuel use as a back-end option: Trends, main issues and impacts on fuel cycle management

    International Nuclear Information System (INIS)

    Fukuda, K.; Choi, J.-S.; Shani, R.; Durpel, L. van den; Bertel, E.; Sartori, E.

    2000-01-01

    In the past decades while the FBIULWR fuel cycle concept was zealously being developed, MOX-fuel use in thermal reactors was taken as an alternative back-end policy option. However, the plutonium recycling with LWRs has evolved to industrial level, gaining high maturity through the incubative period while FBR deployment was envisaged. Today, MOX-fuel use in LWRs makes integral part of the fuel cycle for those countries relying on the recycling policy. Developments to improve the fuel cycle performance, including the minimisation of remaining wastes, and the reactor engineering aspects owing to MOX-fuel use, are continued. This paper jointly presented by IAEA and OECD/NEA brings an integrated overview on MOX use as a back-end policy, covering MOX fuel utilisation, fuel performance and technology, economics, licensing, MOX fuel trends in the coming decades. (author)

  2. Results of current fabrication technology developments and MOX fuel fabrication for ''JOYO'' MK-III initial load fuel

    International Nuclear Information System (INIS)

    Kayano, Masashi

    2003-01-01

    In the Plutonium Fuel Production Facility (PFPF), MOX fuel fabrication technologies have been developed and demonstrated through MOX fuels fabrication for Experimental Fast Reactor ''JOYO'' and Prototype FBR MONJU since 1988. From 1995 to 2000, replacement, modification and repair works for process equipment were conducted to improve performance of the MOX pellet fabrication process in PFPF as scheduled shut-down maintenance. Because the MOX fuel fabrication for ''JOYO'' MK-III initial load fuels was the first fuel fabrication after these major maintenance works, the performance of the MOX pellet fabrication process in PFPF was evaluated though this fuel fabrication experience. This MOX fuel fabrication was completed within the scheduled period and showed higher yield of product MOX pellets than before. Therefore, the performance of the MOX pellet fabrication process in PFPF was improved by this maintenance work. Furthermore, the quality assurance system for MOX fuel fabrication was strengthened by acquisition of ISO9001 certificate in 2002. (author)

  3. Analytical methods for fissionable materials in the nuclear fuel cycle. Progress report, July 1, 1975--September 30, 1976

    International Nuclear Information System (INIS)

    Waterbury, G.R.

    1976-12-01

    Progress continued on development of dissolution techniques for difficult-to-dissolve nuclear materials, development of methods and automated instruments for determinations of plutonium and uranium, preparation of plutonium-containing materials for the Safeguards Analytical Laboratory Evaluation (SALE) program, analysis of SALE uranium materials, and measurement of plutonium isotope half-lives. Gas-solid reactions at elevated temperatures using reactive gases such as chlorine continue to show promise for separating uranium from refractory materials. An extensive study of nonaqueous solvents for the dissolution of refractory materials is in progress. An extraction-separation procedure, highly specific for microgram amounts of uranium, has been developed, and its adaptation to the Los Alamos Scientific Laboratory (LASL) automated spectrophotometer is being evaluated. Development of an electrometric analysis method for plutonium is nearing completion, and design of an automated instrument using the method has been started. Batches of plutonium oxide and mixed uranium--plutonium, intended for issue as Secondary Reference and Calibration Test Materials, are being recharacterized for assay and isotopic contents. The half-life of 239 Pu has been determined by isotope-dilution mass-spectrometric measurement of 235 U grow-in as a function of time

  4. Review of analytical techniques to determine the chemical forms of vapours and aerosols released from overheated fuel

    International Nuclear Information System (INIS)

    Bowsher, B.R.; Nichols, A.L.

    1989-12-01

    A comprehensive review has been undertaken of appropriate analytical techniques to monitor and measure the chemical effects that occur in large-scale tests designed to study severe reactor accidents. Various methods have been developed to determine the chemical forms of the vapours, aerosols and deposits generated during and after such integral experiments. Other specific techniques have the long-term potential to provide some of the desired data in greater detail, although considerable efforts are still required to apply these techniques to the study of radioactive debris. Such in-situ and post-test methods of analysis have been also assessed in terms of their applicability to the analysis of samples from the Phebus-FP tests. The recommended in-situ methods of analysis are gamma-ray spectroscopy, potentiometry, mass spectrometry, and Raman/UV-visible absorption spectroscopy. Vapour/aerosol and deposition samples should also be obtained at well-defined time intervals during each experiment for subsequent post-test analysis. No single technique can provide all the necessary chemical data from these samples, and the most appropriate method of analysis involves a complementary combination of autoradiography, AES, IR, MRS, SEMS/EDS, SIMS/LMIS, XPS and XRD

  5. Analytical methods for fissionable material determinations in the nuclear fuel cycle. Progress report, October 1, 1977--September 30, 1978

    International Nuclear Information System (INIS)

    Waterbury, G.R.

    1979-01-01

    Work has continued on the development of dissolution techniques for difficult-to-dissolve nuclear materials, development of methods and automated instruments for plutonium and uranium determinations, preparation of plutonium-containing materials for the Safeguards Analytical Laboratory Evaluation (SALE) program, preparation of plutonium materials for distribution by the National Bureau of Standards (NBS) as standard reference materials (SRMs), measurement of longer plutonium isotope half-lives, and analysis of SALE uranium materials. New tasks include the development of methods and automated instruments for the determination of thorium and uranium, and an evaluation of the ion-exchange-bead technique for the mass spectrometric measurement of uranium and plutonium isotope distributions. Completed tasks include the measurements of ion exchange distributions of over 50 elements on cation exchange resins from nitric acid media and anion exchange resins from hydrobromic and hydriodic acid media. Using a newly developed procedure, the LASL automated spectrophotometer was modified to determine microgram levels of uranium and to determine milligram levels of uranium and plutonium. Construction of an automated controlled-potential analyzer for the determination of plutonium is nearing completion. Apparatus and procedures for the separation and complexometric titration of thorium and uranium are being developed

  6. Analytical methods for fissionable material determinations in the nuclear fuel cycle. Progress report, October 1, 1978-September 30, 1979

    International Nuclear Information System (INIS)

    Waterbury, G.R.

    1980-03-01

    Work continues on the development of dissolution techniques for difficult-to-dissolve nuclear materials, the development of methods and automated instruments for plutonium, uranium, and thorium determinations, and the preparation of plutonium materials for the Safeguards Analytical Laboratory Evaluation (SALE) program and distribution by the National Bureau of Standards (NBS) as standard reference materials (SRMs). We are measuring the loner plutonium isotope half-lives, evaluating the isotope correlation techniques and the chemistry involved in the mass-spectrometric ion-bead techniques, and analyzing the SALE uranium materials. Completed subtasks include evaluations of various Teflon materials to recommend those acceptable for the dissolution apparatus developed at LASL, investigations of laser-enhanced dissolution of refractory materials, determinations of diverse ion effects on the microgram-sensitive method for determining uranium, fabrication of the first automated controlled-potential coulometric analyzer for determining plutonium, preparation of a 244 Pu material for distribution by NBS as a SRM, and determination of the half-life of 239 Pu. Work has been started on a spectrophotometric method for determining microgram quantities of plutonium, a microcomplexometric titration method for determining uranium, the use of new reagents for separations of plutonium, the preparation and packaging of a new lot of high-purity plutonium metal for distribution by NBS as a plutonium chemical SRM, and determination of half-lives of other plutonium isotopes

  7. Analytical methods for fissionable materials in the nuclear fuel cycle. Progress report, July 1, 1975--September 30, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Waterbury, G.R. (comp.)

    1976-12-01

    Progress continued on development of dissolution techniques for difficult-to-dissolve nuclear materials, development of methods and automated instruments for determinations of plutonium and uranium, preparation of plutonium-containing materials for the Safeguards Analytical Laboratory Evaluation (SALE) program, analysis of SALE uranium materials, and measurement of plutonium isotope half-lives. Gas-solid reactions at elevated temperatures using reactive gases such as chlorine continue to show promise for separating uranium from refractory materials. An extensive study of nonaqueous solvents for the dissolution of refractory materials is in progress. An extraction-separation procedure, highly specific for microgram amounts of uranium, has been developed, and its adaptation to the Los Alamos Scientific Laboratory (LASL) automated spectrophotometer is being evaluated. Development of an electrometric analysis method for plutonium is nearing completion, and design of an automated instrument using the method has been started. Batches of plutonium oxide and mixed uranium--plutonium, intended for issue as Secondary Reference and Calibration Test Materials, are being recharacterized for assay and isotopic contents. The half-life of /sup 239/Pu has been determined by isotope-dilution mass-spectrometric measurement of /sup 235/U grow-in as a function of time.

  8. New type fuel exchange system

    International Nuclear Information System (INIS)

    Meshii, Toshio; Maita, Yasushi; Hirota, Koichi; Kamishima, Yoshio.

    1988-01-01

    When the reduction of the construction cost of FBRs is considered from the standpoint of the machinery and equipment, to make the size small and to heighten the efficiency are the assigned mission. In order to make a reactor vessel small, it is indispensable to decrease the size of the equipment for fuel exchange installed on the upper part of a core. Mitsubishi Heavy Industries Ltd. carried out the research on the development of a new type fuel exchange system. As for the fuel exchange system for FBRs, it is necessary to change the mode of fuel exchange from that of LWRs, such as handling in the presence of chemically active sodium and inert argon atmosphere covering it and handling under heavy shielding against high radiation. The fuel exchange system for FBRs is composed of a fuel exchanger which inserts, pulls out and transfers fuel and rotary plugs. The mechanism adopted for the new type fuel exchange system that Mitsubishi is developing is explained. The feasibility of the mechanism on the upper part of a core was investigated by water flow test, vibration test and buckling test. The design of the mechanism on the upper part of the core of a demonstration FBR was examined, and the new type fuel exchange system was sufficiently applicable. (Kako, I.)

  9. Spent fuel management in Japan

    International Nuclear Information System (INIS)

    Mineo, H.; Nomura, Y.; Sakamoto, K.

    1998-01-01

    In Japan 52 commercial nuclear power units are now operated, and the total power generation capacity is about 45 GWe. The cumulative amount of spent fuel arising is about 13,500 tU as of March 1997. Spent fuel is reprocessed, and recovered nuclear materials are to be recycled in LWRs and FBRs. In February 1997 short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, backend measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away from reactor sites, considering the increasing amount of spent fuel arising. Research and development on spent fuel storage has been carried out, particularly on dry storage technology. Fundamental studies are also conducted to implement the burnup credit into the criticality safety design of storage and transportation casks. Rokkasho reprocessing plant is being constructed towards its commencement in 2003, and Pu utilization in LWRs will be started in 1999. Research and development of future recycling technology are also continued for the establishment of nuclear fuel cycle based on FBRs and LWRs. (author)

  10. Analytical Chemistry Department annual report, 1975

    International Nuclear Information System (INIS)

    Mosen, A.W.

    1976-01-01

    The analytical methods developed or adopted for use in support of radiochemistry and gamma ray spectroscopy, HTGR fuel reprocessing, HTGR fuel development, TRIGA fuel fabrication, and miscellaneous projects are reported

  11. Impurities determination on nuclear fuel element components for the IEA-R1 research reactor by analytical methods based on ED-XRF and ICP-OES

    International Nuclear Information System (INIS)

    Reis, Edson Luis Tocaia dos; Scapin, Marcos; Cotrim, Marycel Elena Barboza; Salvador, Vera Lucia; Pires, Maria Aparecida Faustino

    2009-01-01

    The production of nuclear fuel used in the research reactor at Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) requires a series of chemical and metallurgical processes. The quality of the end product depends on the control over all the stages of the manufacturing process and over the quality of raw materials employed. In fact, spectrometric methods are increasingly used as quantitative analytical techniques applicable to uranium compounds because of simultaneous determination of several elements with minimum amounts of sample. However, the main obstacle of uranium compounds analysis by spectrometric techniques such as optical emission spectrometry with inductively coupled plasma (ICP-OES) is the complex emission spectrum of uranium. The ICP-OES is not appropriately capable of determining the major elements of interest without initial chemical separation of uranium. In this sense, the use of X-ray fluorescence spectrometry (XRF) has been considered for quantitative determination of main elements with the advantage of not being destructive and not requiring a prior preparation of samples for analysis. Due to the simplicity of this technique, its applicability includes research and quality control in universities, research institutions, petrochemical industries, metallurgy, mining, etc. In this work, some components considered impurities in nuclear fuel element samples used in the IEA-R1 research reactor of IPEN/CNEN-SP were chemically characterized by ICP-OES analysis after chromatography extraction separation by using TBP/XAD-14 system and compared to results obtained by energy dispersive X-ray fluorescence spectrometry (EDXRF) and wavelength dispersive X-ray fluorescence (WDXRF). (author)

  12. Resilience and Livelihoods in Supply Chains (RELISC: An Analytical Framework for the Development and Resilience of the UK Wood Fuel Sector

    Directory of Open Access Journals (Sweden)

    Damiete Emmanuel-Yusuf

    2017-04-01

    Full Text Available Bioenergy is an important renewable energy source in the UK, but the bioenergy industry and in particular the wood fuel sub sector, is relatively under-developed. Socioeconomic factors have been identified as critical for facilitating deployment levels and sustainable development. However, previous studies have mostly assessed these factors using quantitative methods and models, which are limited in assessing pertinent contextual factors such as institutional/regulatory governance, supply chain structure and governance, capital resource availability as well as actor decisions. As a step further, this research engages with these under-explored aspects of the system by developing a new analytical framework: the Resilience and Livelihoods in Supply Chains (RELISC framework, which was designed by linking Value Chain Analysis, the Sustainable Livelihoods Approach and a supply chain resilience framework. Its application to a UK wood fuel supply chain produced useful insights. For example, the structure of the chain revealed a high level of dependency on a particular end user and contractor. Key institutional governance was critical in sustaining natural resources and providing access to finance. Internal supply chain governance was limited in ensuring the sustainability of resources and lack of actor awareness and interest were also limiting factors. In addition, five capital analyses revealed gaps in skills, networking and physical infrastructure. Finally, the design of the novel RELISC framework enables it to engage with diverse aspects of the system holistically and its application generated practical recommendations and strategies for supply chain resilience and sector growth, which are useful and applicable to other emerging sectors.

  13. Simulation of Fungal-Mediated Cell Death by Fumonisin B1 and Selection of Fumonisin B1–Resistant (fbr) Arabidopsis Mutants

    Science.gov (United States)

    Stone, Julie M.; Heard, Jacqueline E.; Asai, Tsuneaki; Ausubel, Frederick M.

    2000-01-01

    Fumonisin B1 (FB1), a programmed cell death–eliciting toxin produced by the necrotrophic fungal plant pathogen Fusarium moniliforme, was used to simulate pathogen infection in Arabidopsis. Plants infiltrated with 10 μM FB1 and seedlings transferred to agar media containing 1 μM FB1 develop lesions reminiscent of the hypersensitive response, including generation of reactive oxygen intermediates, deposition of phenolic compounds and callose, accumulation of phytoalexin, and expression of pathogenesis-related (PR) genes. Arabidopsis FB1-resistant (fbr) mutants were selected directly by sowing seeds on agar containing 1 μM FB1, on which wild-type seedlings fail to develop. Two mutants chosen for further analyses, fbr1 and fbr2, had altered PR gene expression in response to FB1. fbr1 and fbr2 do not exhibit differential resistance to the avirulent bacterial pathogen Pseudomonas syringae pv maculicola (ES4326) expressing the avirulence gene avrRpt2 but do display enhanced resistance to a virulent isogenic strain that lacks the avirulence gene. Our results demonstrate the utility of FB1 for high-throughput isolation of Arabidopsis defense-related mutants and suggest that pathogen-elicited programmed cell death of host cells may be an important feature of compatible plant–pathogen interactions. PMID:11041878

  14. Start-up and long-term operation of the Anammox process in a fixed bed reactor (FBR) filled with novel non-woven ring carriers.

    Science.gov (United States)

    Wang, Tao; Zhang, Hanmin; Yang, Fenglin; Li, Yifei; Zhang, Guangyi

    2013-04-01

    A novel kind of non-woven ring carriers was used to improve a fixed bed reactor (FBR) as Anammox reactor. The improved FBR was operated for about 1 year. The Anammox activity occurred on day 39. On day 367, the maximum total nitrogen removal rate reached 9.2 kg Nm(-3)d(-1). FISH analysis showed that Anammox bacteria predominated in the mature sludge and accounted for 78% of the total bacteria. Phylogenetic analysis further showed that Candidatus Kuenenia stuttgartiensis occupied 70% of Anammox bacteria, which benefited keeping the stability of Anammox reactor. The FBR was proved to be a suitable reactor for start-up and long-term operation of Anammox process. Crown Copyright © 2013. Published by Elsevier Ltd. All rights reserved.

  15. The development and application of overheating failure model of FBR steam generator tubes. 2

    International Nuclear Information System (INIS)

    Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi

    2001-11-01

    The JNC technical report 'The Development and Application of Overheating Failure Model of FBR Steam Generator Tubes' summarized the assessment method and its application for the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). This report describes the following items studied after the publication of the above technical report. 1. On the basis of the SWAT-3 experimental data, realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed. New correlations are cosine-shaped temperature profiles with 1,170 C maximum for the 100% and 40% Monju operating conditions, and those with 1,110 C maximum for the 10% condition. 2. For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the assessment. The revised assessment adopts the Katto's correlation for CHF, and the Condie-Bengston IV correlation for post-CHF. 3. Other additional examination for the assessment includes treatments of the whole heating effect (other than the local reaction zone) due to the sodium-water reaction, and the temperature-dependent thermal properties of the heat transfer tube material (2.25Cr-1Mo steel). The revised overheating tube failure assessment method has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% operating conditions when an initial leak is detected by the cover gas pressure detection system. The assessment for the SG system improved for the detection and blowdown systems shows even better safety margins against the overheating tube failure. (author)

  16. Development and study of a control and reactor shutdown device for FBR-type reactors with a modified open core

    International Nuclear Information System (INIS)

    Goswami, S.

    1983-01-01

    The doctoral thesis at hand presents a newly designed control and shutdown device to be used for output control and fast shutdown of modified open core FBR-type reactors. The task was the design of a new control and shutdown device having economic and operation advantages, using reactor components time-tested under reactor conditions. This control and shutdown device was adapted to the specific needs concerning dimensions and design. The actuation is based on the magnetic-jack principle, which has been upgraded for the purpose. The principle is now combined with pneumatic acceleration. The improvements mainly concern a smaller number of piece parts and system simplification. (orig./RW) [de

  17. 3-D numerical simulation on the vibration of liquid sodium's free surface in sodium pool of FBR

    International Nuclear Information System (INIS)

    Han Biao; Yao Zhaohui; Ye Hongkai; Wang Xuefang

    1997-01-01

    This paper succeeds in simulating three-dimensional incompressible flows with free surface, complicated in-flow and out-flow boundary conditions and internal obstacles, and also can treat these fluid flows in arbitrary shape vessel using a partial cell. According to all kinds of the element influencing the free surface's vibration in sodium pool it may give the various wave's form, the highest and lowest position, and the amount of the vibration. This paper introduces the brief principle of VOF numerical method, develops the computational program based on NASA-VOF3D, provides some results about the free surface's vibration in sodium pool of FBR

  18. MONJU experimental data analysis and its feasibility evaluation to build up the standard data base for large FBR nuclear core design

    International Nuclear Information System (INIS)

    Sugino, K.; Iwai, T.

    2006-01-01

    MONJU experimental data analysis was performed by using the detailed calculation scheme for fast reactor cores developed in Japan. Subsequently, feasibility of the MONJU integral data was evaluated by the cross-section adjustment technique for the use of FBR nuclear core design. It is concluded that the MONJU integral data is quite valuable for building up the standard data base for large FBR nuclear core design. In addition, it is found that the application of the updated data base has a possibility to considerably improve the prediction accuracy of neutronic parameters for MONJU. (authors)

  19. Nuclear fuels; Les combustibles nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    neutrons radiation-induced damages in structural materials, fuels and targets for the transmutation in FBR; 4 - the fuel of gas-cooled reactors: fuel particulates, behaviour under irradiation, mechanical modeling, the fuel for very high temperature reactors, the fuel for gas-cooled fast reactors; 5 - the fuel for research reactors; 6 - the Jules Horowitz reactor: a tool for the future fuel studies. (J.S.)

  20. Status of fuel transmutation programmes in Japan and France. Lessons drawn from results

    Energy Technology Data Exchange (ETDEWEB)

    Arai, Y.; Pillon, S

    2004-07-01

    France and Japan are currently developing a comprehensive and complementary programme focusing on the transmutation of minor actinides (MA: Np, Am, Cm) and fission products (FP: Tc, I, Cs) in fast breeder reactors (FBR). A summary of current MA-fuel transmutation programmes in France and Japan is provided in this paper, covering objectives, results and perspectives, with emphasis placed on the complementary effort of the two countries. (authors)

  1. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    , Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO 2 ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO 2 and MOX ceramics (Chromium oxide-doped UO 2 fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The nature of spent nuclear

  2. PENENTUAN FAKTOR YANG BERPENGARUH DALAM FAULTY BEHAVIOR RISK MELALUI PENDEKATAN METODE FUZZY ANALYTIC HIERARCHY PROCESS

    Directory of Open Access Journals (Sweden)

    Rizal Irfan Fuadi

    2015-06-01

    Full Text Available Manajemen keselamatan merupakan pengorganisasian, sumber daya manusia, kebijakan dan prosedur interaktif yang bertujuan untuk mengurangi kemungkinan kerusakan dan kerugian di tempat kerja. Salah satu cara untuk memperbaiki manajemen keselamatan di perusahaan adalah dengan melakukan penelitian mengenai faktor yang berpengaruh dalam risiko kesalahan perilaku. Secara umum, faktor-faktor yang mempengaruhi keselamatan kerja tidak memiliki struktur fisik. Untuk itu, maka masalah pada kondisi nyata dapat direpresentasikan dengan cara yang lebih baik menggunakan angka fuzzy untuk mengevaluasi faktor-faktor ini. Pada penelitian ini, pendekatan Fuzzy AHP bertujuan untuk menentukan tingkat Faulty Behavior Risk (FBR pada sistem kerja.Penentuan faktor yang berpengaruh dalam Faulty Behavior Risk (FBR/risiko dari perilaku yang salah dimulai dengan menentukan responden, penyusunan kuesioner, uji validitas dan reliabilitas, hasil dari kuesioner dijadikan inputan dalam pengolahan data dengan metode FuzzyAnalytic Hierachy Process (FAHP. Kuesioner ini dibuat berdasarkan konsep safety management yang terdiri dari 4 faktor, yaitu faktor organisasi, faktor pribadi, faktor pekerjaan dan faktor lingkungan. Penelitian ini dilakukan pada Bay 2.1 yang memproduksi panel dan bay 7.1 yang memproduksi finning. Berdasarkan perhitungan FBR pada bay 2.1 menunjukkan nilai 0,4793 yang berarti risiko kesalahan perilaku di antara batas bawah (0,25 dan batas atas (0,50. Sedangkan pada bay 7.1 sebesar 0,5317 yang berarti risiko kesalahan perilaku memiliki potensi tinggi karena berada di atas batas atas. Dari hasil penentuan FBR didapatkan nilai  pada bay 2.1 yang memiliki risiko penyebab tertinggi terdapat pada sub faktor kurang persiapan (0,0788 sedangkan pada bay 7.1 dengan nilai FBR sebesar 0,5317 yang memiliki risiko tertinggi terdapat pada sub faktor kelelahan kerja (0,0970. Melalui penelitian ini, faktor penyebab kesalahan perilaku kerja dapat diketahui dan diberikan

  3. In pile programme of first valutation of UO2 + PuO2 fuel produced by a new process (GSP)

    International Nuclear Information System (INIS)

    Caracchin, R.; Lanchi, M.; Marinucci, G.; Nobili, A.; Dupont, G.; Galtier, J.

    1982-01-01

    The main scope of the ENEA-AGN-CEA programme collaboration is a first valutation of fuel elements produced by GSP method. This valuation will be done by in reactor experiment which enable to compare the performance of GSP and 'standard' FBR fuels. The composition is done by means of theree experimental device: P3, Lugel and Digel. The P3 device gives a direct measurement during irradiation of fuel central temperature, power and integral conductivity. The Lugel device measures fuel stack axial variations and Digel device gives the diameter variations of the pin and PCMI

  4. Impact of partitioning and transmutation on high-level waste disposal for the fast breeder reactor fuel cycle

    International Nuclear Information System (INIS)

    Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

    2010-01-01

    The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large

  5. Specialists' meeting on thermodynamics of FBR fuel subassemblies under nominal and non-nominal operating conditions. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-01

    The purpose of the meeting was to provide a forum for exchange of information on thermo- and fluiddynamic investigations of LMFBR-subassembly. Special emphasis was placed on nominal and non-nominal conditions. The technical part of the meeting was divided into four sessions, as follows: status of the thermo- and fluiddynamic activities; physical and mathematical modelling of single phase; rod bundle thermohydraulics; experimental investigations; and future R and D. Separate abstracts are included for each of the papers.

  6. Advanced fuel fabrication

    International Nuclear Information System (INIS)

    Bernard, H.

    1989-01-01

    This paper deals with the fabrication of advanced fuels, such as mixed oxides for Pressurized Water Reactors or mixed nitrides for Fast Breeder Reactors. Although an extensive production experience exists for the mixed oxides used in the FBR, important work is still needed to improve the theoretical and technical knowledge of the production route which will be introduced in the future European facility, named Melox, at Marcoule. Recently, the feasibility of nitride fuel fabrication in existing commercial oxide facilities was demonstrated in France. The process, based on carbothermic reduction of oxides with subsequent comminution of the reaction product, cold pressing and sintering provides (U, Pu)N pellets with characteristics suitable for irradiation testing. Two experiments named NIMPHE 1 and 2 fabricated in collaboration with ITU, Karlsruhe, involve 16 nitride and 2 carbide pins, operating at a linear power of 45 and 73 kW/m with a smear density of 75-80% TD and a high burn-up target of 15 at%. These experiments are currently being irradiated in Phenix, at Marcoule. (orig.)

  7. Multi-frequencies ECT algorithms to remove sodium noise in ISI of ferromagnetic SG tubes of FBR

    International Nuclear Information System (INIS)

    Mihalache, Ovidiu

    2012-01-01

    The paper presents developments and application of multi-frequency eddy current to be used during In-Service Inspection (ISI) of ferromagnetic steam generator (SG) tubes of Fast Breeder Reactors (FBR). Signal enhancement by means of multi-frequency ECT techniques are validated through 3D simulations of both signals and noise due to sodium forms around SG tube or SP. The purpose of such algorithms is to remove from ECT signal the electromagnetic noise resulting from sodium accumulated outside of SG tubes after SG vessel draining. Finite element method (FEM) simulations are used to analyse different sodium build-up scenarios observed experimentally, and to determine optimal multi-frequency ECT algorithms to suppress the most efficiently sodium noise. Also a new 'window multi-frequency' algorithm is applied and validated using 3-dimensional FEM simulations of SP and sodium forms. (author)

  8. Study on thermal electric conversion system for FBR plant. Investigation for effective EVST waste heat recovery system

    International Nuclear Information System (INIS)

    Maekawa, Isamu; Kurata, Chikatoshi

    2004-02-01

    Recently, it has been important to reuse discharged heat energy from present nuclear plant, especially from sodium cooled FBR, which are typical high temperature system, in the view of reduction of environmental burden and improvement of heat efficiency for plant. The thermal electric conversion system can work only the temperature difference and has been applied to the limited fields such as space or military, however, that results show good merits for reliability, maintenance free, and so on. Recently, the development of new thermal electric conversion elements has made remarkable progress. In this study, for the effective utilization of waste heat from Monju', the prototype plant of FBR, we made an investigation of electric power generating system maintaining the cooling faculty by applying the thermal electric conversion system to sodium cooling line of EVST. Using the new type iron based thermal electric conversion elements, which are plentiful, economical and good for environmental harmonization, we have calculated the amount of heat exchange and power generation from sodium cooling line of EVST, and have investigated the module sizing, cost and subject to be settled. The results were , (1)The amount of power generation from sodium cooling line of EVST is smaller about one figure than motive power of sodium cooler fan. However, if Seebeck coefficient and heat conductivity of iron based thermal electric conversion elements shall be improved, power from sodium cooling line shall be able to cover the motive power. (2) The amount of heat released from sodium cooling line after the installation of thermal electric conversion module covers the necessity to maintain the sodium cooling faculty. (3) In case of the installation of module to the sodium cooler, it should be reconstructed because of tube arrangement modification. In case of the installation of module to the sodium connecting line, air ventilation system is needed to suppress the room temperature. (4) As

  9. Analytical methods used at model facility

    International Nuclear Information System (INIS)

    Wing, N.S.

    1984-01-01

    A description of analytical methods used at the model LEU Fuel Fabrication Facility is presented. The methods include gravimetric uranium analysis, isotopic analysis, fluorimetric analysis, and emission spectroscopy

  10. Web Analytics

    Science.gov (United States)

    EPA’s Web Analytics Program collects, analyzes, and provides reports on traffic, quality assurance, and customer satisfaction metrics for EPA’s website. The program uses a variety of analytics tools, including Google Analytics and CrazyEgg.

  11. Purification of uranium products in crystallization system for nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Takeuchi, Masayuki; Yano, Kimihiko; Shibata, Atsuhiro; Sanbonmatsu, Yuji; Nakamura, Kazuhito; Chikazawa, Takahiro; Hirasawa, Izumi

    2016-01-01

    Uranium crystallization system has been developed to establish an advanced aqueous reprocessing for fast breeder reactor (FBR) fuel cycle. In the crystallization system, most part of uranium in dissolved solution of spent FBR-MOX fuels is separated as uranyl nitrate hexahydrate (UNH) crystals by a cooling operation. The targets of U yield and decontamination factor (DF) on the crystallization system are decided from FBR cycle performance and plutonium enrichment management. The DF is lowered by involving liquid and solid impurities on and in the UNH crystals during crystallization. In order to achieve the DF performance (more than 100), we discuss the purification technology of UNH crystals using a Kureha Crystal Purifier (KCP). Results show that more than 90% of uranium in the feed crystals could be recovered as the purified crystals in all test conditions, and the DFs of solid and liquid impurities on the purified UNH crystals are more than 100 under longer residence time of crystals in the column of KCP device. The purification mechanism is mainly due to the repetition of sweating and recrystallization in the column under controlled temperature. (author)

  12. Transport and storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Lung, M.; Lenail, B.

    1987-01-01

    From a safety standpoint, spent fuel is clearly not ideal for permanent disposal and reprocessing is the best method of preparing wastes for long-term storage in a repository. Furthermore, the future may demonstrate that some fission products recovered in reprocessing have economic applications. Many countries have in fact reached the point at which the recycling of plutonium and uranium from spent fuel is economical in LWR's. Even in countries where this is not yet evident, (i.e., the United States), the French example shows that the day will come when spent fuel will be retrieved for reprocessing and recycle. It is highly questionable whether spent fuel will ever be considered and treated as waste in the same sense as fission products and processed as such, i.e., packaged in a waste form for permanent disposal. Even when recycled fuel material can no longer be reused in LWR's because of poor reactivity, it will be usable in FBR's. Based on the considerable experience gained by SGN and Cogema, this paper has provided practical discussion and illustrations of spent fuel transport and storage of a very important step in the nuclear fuel management process. The best of spent fuel storage depends on technical, economic and policy considerations. Each design has a role to play and we hope that the above discussion will help clarify certain issues

  13. A review on manufacturing technology for long-lived radionuclide fuel compounds

    International Nuclear Information System (INIS)

    Hwang, Doo Seong; Park, Jin Ho; Kim, Eung Ho; Chung, Won Myung; Lee, Kui Ill; Woo, Moon Sik; Kim, Yeon Ku; Yoo, Jae Hyung

    1998-03-01

    Thermal neutron reactor (LWR), fast neutron reactor (FBR), accelerator-driven subcritical system have been studied as the potential transmutation devices. The fuel types can be classified according to the concept of each reactor. Oxide fuel is considered in LWR and metal, oxide, and nitride fuels are studied in FBR. In accelerator-driven subcritical system molten salt, metal, and oxide fuels are considered. This review focused on characteristics according to transmutation system, and manufacturing technologies of each fuels. Accelerator-driven system is being proposed as the most reasonable concept in recent, since it has merits in terms of stability and free control of nuclides composition rate in charge of long-lived nuclides. Fluorides molten salt fuel is better chemically stable and corrosion resistant, and lower vapor pressure than chloride molten salt and metal in the fuel type of accelerator-driven system. And then the detail manufacturing technology of fluorides molten salt were reviewed. (author). 62 refs., 23 tabs., 37 figs

  14. Temperature distribution and local boiling behind a central blockage in a simulated FBR subassembly

    International Nuclear Information System (INIS)

    Brook, A.J.; Huber, F.; Peppler, W.

    1976-01-01

    A series of experiments has been carried out to investigate the effects of localised disturbance to the normal coolant flow in a fast reactor fuel element. The tests involved an electrically heated bundle of 169 pins, with a centrally located blockage extending over 49% of the flow area. Test section geometry corresponded to the SNR 300 Mk 1a fuel element. Measured temperature distributions behind the blockage agreed well with those measured in corresponding water experiments. The observed features of local boiling are discussed, and it is shown that a continued capability for cooling the blockage region is preserved, even with intensive local boiling

  15. Development and Validation of a Simple Analytical Model of the Proton Exchange Membrane Fuel Cell (Pemfc) in a Fork-Lift Truck Power System

    DEFF Research Database (Denmark)

    Hosseinzadeh, Elham; Rokni, Masoud

    2013-01-01

    In this study, a general proton exchange membrane fuel cell (PEMFC) model has been developed in order to investigate the balance of plant of a fork-lift truck thermodynamically. The model takes into account the effects of pressure losses, water crossovers, humidity aspects, and voltage...... management, system sensitivity to coolant inlet temperature, air and fuel stoichiometry, anode inlet pressure, stack operating conditions, etc. System efficiency and electrical power at different operating conditions are also discussed. The results show that 12–30% of stack power is allocated...

  16. Development of an analytical framework to assess the role of new technologies for liquid and gaseous fuels. Volume I. Background. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Marshalla, R.A.; Nesbitt, D.M.; Sullivan, S.J.

    1980-03-01

    A central issue in liquid and gaseous fuel R and D is how long conventional oil and gas will last. This volume investigates that issue by exploring the methodologies used to characterize the domestic base of oil and natural gas including detailed reviews of recent resource models and their applicability to the problem of R and D policy analysis in liquid and gaseous fuels. Detailed reviews and comments are made on 9 studies on enhanced oil and gas recovery and its potential, leasing strategies and schedules for the Outer Continental Shelf, and energy models. 38 figures, 19 tables. (DMC)

  17. On recycling of nuclear fuel in Japan

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    In Japan, atomic energy has become to accomplish the important role in energy supply. Recently the interest in the protection of global environment heightened, and the anxiety on oil supply has been felt due to the circumstances in Mideast. Therefore, the importance of atomic energy as an energy source for hereafter increased, and the future plan of nuclear fuel recycling in Japan must be promoted on such viewpoint. At present in Japan, the construction of nuclear fuel cycle facilities is in progress in Rokkasho, Aomori Prefecture. The prototype FBR 'Monju' started the general functional test in May, this year. The transport of the plutonium reprocessed in U.K. and France to Japan will be carried out in near future. This report presents the concrete measures of nuclear fuel recycling in Japan from the long term viewpoint up to 2010. The necessity and meaning of nuclear fuel recycling in Japan, the effort related to nuclear nonproliferation, the plan of nuclear fuel recycling for hereafter in Japan, the organization of MOX fuel fabrication in Japan and abroad, the method of utilizing recovered uranium and the reprocessing of spent MOX fuel are described. (K.I.).

  18. Analytical investigation of high temperature 1 kW solid oxide fuel cell system feasibility in methane hydrate recovery and deep ocean power generation

    International Nuclear Information System (INIS)

    Azizi, Mohammad Ali; Brouwer, Jacob; Dunn-Rankin, Derek

    2016-01-01

    Highlights: • A dynamic Solid Oxide Fuel Cell (SOFC) model was developed. • Hydrate bed methane dissociation model was integrated with the SOFC model. • SOFC operated steadily for 120 days at high pressure deep ocean environment. • Burning some of the dissociated gas for SMR heat leads to more net methane produced. • Higher SOFC fuel utilization produces higher integrated system efficiency. - Abstract: Methane hydrates are potential valuable energy resources. However, finding an efficient method for methane gas recovery from hydrate sediments is still a challenge. New challenges arise from increasing environmental protection. This is due in part to the technical difficulties involved in the efficient dissociation of methane hydrates at high pressures. In this study, a new approach is proposed to produce valuable products of: 1. Net methane gas recovery from the methane hydrate sediment, and 2. Deep ocean power generation. We have taken the first steps toward utilization of a fuel cell system in methane gas recovery from deep ocean hydrate sediments. An integrated high pressure and high temperature solid oxide fuel cell (SOFC) and steam methane reformer (SMR) system is analyzed for this application and the recoverable amount of methane from deep ocean sediments is measured. System analysis is accomplished for two major cases regarding system performance: 1. Energy for SMR is provided by the burning part of the methane gas dissociated from the hydrate sediment. 2. Energy for SMR is provided through heat exchange with fuel cell effluent gases. We found that the total production of methane gas is higher in the first case compared to the second case. The net power generated by the fuel cell system is estimated for all cases. The primary goal of this study is to evaluate the feasibility of integrated electrochemical devices to accomplish energy efficient dissociation of methane hydrate gases in deep ocean sediments. Concepts for use of electrochemical devices

  19. The estimation of heavy metal concentration in FBR reprocessing solvent streams by density measurement

    International Nuclear Information System (INIS)

    Brown, M.L.; Savage, D.J.

    1986-04-01

    The application of density measurement to heavy metal monitoring in the solvent phase is described, including practical experience gained during three fast reactor fuel reprocessing campaigns. An experimental algorithm relating heavy metal concentration and sample density was generated from laboratory-measured density data, for uranyl nitrate dissolved in nitric acid loaded tri-butyl phosphate in odourless kerosene. Differences in odourless kerosene batch densities are mathematically interpolated, and the algorithm can be used to estimate heavy metal concentrations from the density to within +1.5 g/l. An Anton Paar calculating digital densimeter with remote cell operation was used for all density measurements, but the algorithm will give similar accuracy with any density measuring device capable of a precision of better than 0.0005 g/cm 3 . For plant control purposes, the algorithm was simplified using a density referencing system, whereby the density of solvent not yet loaded with heavy metal is subtracted from the sample density. This simplified algorithm compares very favourably with empirical algorithms, derived from numerical analysis of density data and chemically measured uranium and plutonium data obtained during fuel reprocessing campaigns, particularly when differences in the acidity of the solvent are considered before and after loading with heavy metal. This simplified algorithm had been successfully used for plant control of heavy metal loaded solvent during four fast reactor fuel reprocessing campaigns. (author)

  20. Current status of vibro-packed fuel fabrication process development

    International Nuclear Information System (INIS)

    Kihara, Yoshiyuki

    2004-01-01

    In the feasibility study of FBR cycle system, the concept of future commercialized fuel is 'low-decontaminated MA(minor actinide) fuel', which contains high radioactive rare earth elements and MAs. This fuel will optimize a positive feature of the fast reactor, which allows high impurities for the fuel. Therefore, this fuel should be fabricated not in a conventional glove-box but in a cell with remote handling. A vibro-packed fuel, like sphere-pac fuel or vipac fuel, is well-suited, and the development of fabrication procedure is in progress as a candidate for the commercialized fuel fabrication process. Results of the three topics are reported: 1) results of uranium dioxide particle fabrication with external gelation process containing samarium (Sm) as a representative element of rare earth impurities. 2) results of vibro-packing experiments using several simulated materials for the prediction of filling behavior during sphere-pac fuel fabrication and 3) results of evaluation on filling behavior of vipac fuel particles with getter (uranium metal) particles. (author)

  1. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  2. Fuel flexible fuel injector

    Science.gov (United States)

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  3. On the evaluation method of creep and fatigue strength for FBR welded joints

    International Nuclear Information System (INIS)

    Endo Tadayoshi; Sakon Toshio

    In fast breeder reactors, more detailed design at elevated temperature is required than in conventional fossile fuel power plants. One of the most important experiences in conventional power plants is creep and low cycle fatigue cracking of welded joints of high temperature use. Therefore, more careful requirements will be necessary for welded joint properties for fast breeder reactors. It is necessary to develop more reliable welded joints considering the detailed requirements for breeder reactor use. The authors report some problems experienced in conventional plants, and then discuss the mechanism of fracture of welded joint specimens in laboratory testing. Elevated temperature properties required for more reliable welded joints is also discussed. (author)

  4. Integrated fuel processor development

    International Nuclear Information System (INIS)

    Ahmed, S.; Pereira, C.; Lee, S. H. D.; Krumpelt, M.

    2001-01-01

    The Department of Energy's Office of Advanced Automotive Technologies has been supporting the development of fuel-flexible fuel processors at Argonne National Laboratory. These fuel processors will enable fuel cell vehicles to operate on fuels available through the existing infrastructure. The constraints of on-board space and weight require that these fuel processors be designed to be compact and lightweight, while meeting the performance targets for efficiency and gas quality needed for the fuel cell. This paper discusses the performance of a prototype fuel processor that has been designed and fabricated to operate with liquid fuels, such as gasoline, ethanol, methanol, etc. Rated for a capacity of 10 kWe (one-fifth of that needed for a car), the prototype fuel processor integrates the unit operations (vaporization, heat exchange, etc.) and processes (reforming, water-gas shift, preferential oxidation reactions, etc.) necessary to produce the hydrogen-rich gas (reformate) that will fuel the polymer electrolyte fuel cell stacks. The fuel processor work is being complemented by analytical and fundamental research. With the ultimate objective of meeting on-board fuel processor goals, these studies include: modeling fuel cell systems to identify design and operating features; evaluating alternative fuel processing options; and developing appropriate catalysts and materials. Issues and outstanding challenges that need to be overcome in order to develop practical, on-board devices are discussed

  5. Structure and short time degradation studies of sodium zirconium phosphate ceramics loaded with simulated fast breeder (FBR) waste

    Science.gov (United States)

    Ananthanarayanan, A.; Ambashta, R. D.; Sudarsan, V.; Ajithkumar, T.; Sen, D.; Mazumder, S.; Wattal, P. K.

    2017-04-01

    Sodium zirconium phosphate (NZP) ceramics have been prepared using conventional sintering and hot isostatic pressing (HIP) routes. The structure of NZP ceramics, prepared using the HIP route, has been compared with conventionally sintered NZP using a combination of X-ray diffraction (XRD) and (31P and 23Na) nuclear magnetic resonance (NMR) spectroscopy techniques. It is observed that NZP with no waste loading is aggressive toward the steel HIP-can during hot isostatic compaction and significant fraction of cations from the steel enter the ceramic material. Waste loaded NZP samples (10 wt% simulated FBR waste) show significantly low can-interaction and primary NZP phase is evident in this material. Upon exposure of can-interacted and waste loaded NZP to boiling water and steam, 31P NMR does not detect any major modifications in the network structure. However, the 23Na NMR spectra indicate migration of Na+ ions from the surface and possible re-crystallization. This is corroborated by Small-Angle Neutron Scattering (SANS) data and Scanning Electron Microscopy (SEM) measurements carried out on these samples.

  6. Preparation and optimization of ceramic neutron image plates based on BaFBr : Eu2+ and GdF3

    Science.gov (United States)

    Kolb, R.; Zimmermann, J.; Schlapp, M.; Hesse, S.; von Seggern, H.

    2005-09-01

    Commercially available neutron image plates (NIPs) consist of a mixture of a powdered x-ray storage phosphor and a neutron converter, both embedded in an organic binder supported on a polymer sheet. The initiation of the storage mechanism in the phosphor is caused by conversion electrons generated in the neutron converter due to neutron absorption and activation and its subsequent decay. The organic binder phase just provides mechanical stability to the NIP but reduces its efficiency through two effects: first by the absorption of low energy electrons and second by introducing an inactive volume fraction to the layer. Avoiding the organic fraction, for example by preparing a ceramic NIP without binder, could increase the efficiency and spatial resolution without a loss in mechanical stability. In the following, two processes for preparation of ceramic NIPs are reported, both delivering ceramic NIPs consisting solely of GdF3, as the neutron converter and BaFBr : Eu2+, as the storage phosphor. The correlation between the sintering parameters and volume fraction of the neutron converter is investigated with respect to high efficiency and high spatial resolution. The generally observed antidromic behaviour between these two quantities was observed in this study also.

  7. Development of operation support system for MOX fuel production line

    International Nuclear Information System (INIS)

    Gunji, Yasutoshi; Fujiwara, Shigeo; Iso, Hidetoshi; Suzuki, Yoshihiro; Shishido, Toshio

    1989-01-01

    Plutonium Fuel Production Facility (PFPF) FBR line started production of MOX fuels for 'JOYO' MK-II in October 1988. The production campaign for 'Monju' initial core fuel followed that for 'Joyo'. Control of this line is mainly computerized to allow remote operation. According to our test run experiences the automated plant, requires the higher judgement ability of operator when a problem arises. We need the plant which can be operated by even unskilled operator, on the technical level equal to that of skilled operator. This requirement will be satisfied by introduction of new support system into applying Artificial Intelligence technology based on operating experience. Now we have developed some operation support systems taking main aim at high efficiency of production for example, the optimum operation control system the failure diagnosis system and the production planning support system. (author)

  8. Biodiesel Analytical Methods: August 2002--January 2004

    Energy Technology Data Exchange (ETDEWEB)

    Van Gerpen, J.; Shanks, B.; Pruszko, R.; Clements, D.; Knothe, G.

    2004-07-01

    Biodiesel is an alternative fuel for diesel engines that is receiving great attention worldwide. The material contained in this book is intended to provide the reader with information about biodiesel engines and fuels, analytical methods used to measure fuel properties, and specifications for biodiesel quality control.

  9. Fuel cycle studies

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    Programs are being conducted in the following areas: advanced solvent extraction techniques, accident consequences, fuel cycles for nonproliferation, pyrochemical and dry processes, waste encapsulation, radionuclide transport in geologic media, hull treatment, and analytical support for LWBR

  10. Estimating fuel cycle externalities: Analytical methods and issues. Report number 2 on the external costs and benefits of fuel cycles: A study by the U.S. Department of Energy and the Commission of the European Communities

    International Nuclear Information System (INIS)

    1994-07-01

    This report, the second in a series of eight reports, is part of a joint study by the U.S. Department of Energy (DOE) and the Commission of the European Communities (EC) 'on the externalities of fuel cycles.' Part I illustrates the use of the atmospheric dispersion and transformation modeling that this study recommends for airborne pollutants in the coal, biomass, oil, and natural gas fuel cycles. Part II of this volume contains a paper which reviews the scientific literature on ecological impacts associated with power plant discharges. Part III contains papers summarizing the relevant health effects literature. Part IV contains papers on methods of economic evaluation. Part V contains four papers on various issues related to the estimation of externalities and their use in public policy. The final part is Part VI, and it contains a paper which describes a system for summarizing analysts' assessments of the quality of the information that an analysis uses to estimate externalities. This system allows analysts to provide information, not only on their best estimates, but also on a range of estimates, on uncertainty, on the quality of the data, and on other factors that better reflect the full dimension of making estimates under uncertainty. The system has broad applicability beyond fuel cycle externalities, as well

  11. Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

    International Nuclear Information System (INIS)

    Nakano, Yoshihiro; Okubo, Tsutomu

    2011-01-01

    Highlights: → Pu isotopic composition of fuel affects FBR core nuclear characteristics very much. → Spent fuel compositions of next generation LWRs with burnup of 70 GWd/t were obtained. → Pu isotopic composition and amount in the spent fuel with 70 GWd/t were evaluated. → Spectral shift rods of high burnup BWR increases the fissile Pu fraction of spent fuel. → Wide fuel rod pitch of high burnup PWR lowers the fissile Pu fraction of spent fuel. - Abstract: The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs. The HB-BWR employs spectral shift rods and the neutron spectrum is shifted through the operation cycle. The weight fraction of fissile plutonium (Puf) isotopes to the total plutonium in HB-BWR spent fuel after 5 years cooling is 62%, which is larger than that of conventional BWRs with average burnup of 45 GWd/t, because of the spectral shift operation. The amount of Pu produced in the HB-BWR is also larger than that produced in a conventional BWR. The HB-PWR uses a wider pitch 17 x 17 fuel rod assembly to optimize neutron slowing down. The Puf fraction of HB-PWR spent fuel after 5 years cooling is 56%, which is smaller than that of conventional PWRs with average burnup of 49 GWd/t, mainly because of the wider pitch. The amount of Pu produced in the HB-PWR is also smaller than that in conventional PWRs.

  12. Micro-analytical uranium isotope and chemical investigations of zircon crystals from the Chernobyl “lava” and their nuclear fuel inclusions

    Energy Technology Data Exchange (ETDEWEB)

    Pöml, P., E-mail: Philipp.POEML@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Burakov, B. [Laboratory of Applied Mineralogy and Radiogeochemistry, V.G. Khlopin Radium Institute, 28, 2-nd Murinskiy Ave., St. Petersburg 194021 (Russian Federation); Geisler, T. [Steinmann Institut für Geologie, Mineralogie und Paläontologie, University of Bonn, Poppelsdorfer Schloss, 53115 Bonn (Germany); Walker, C.T. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Grange, M.L.; Nemchin, A.A. [Department of Applied Geology, Western Australian School of Mines, Curtin University, GPO Box U1987, Western Australia 6845 (Australia); Berndt, J. [Institut für Mineralogie, Westfälische Wilhelms-Universität, Corrensstraße 24, 48149 Münster (Germany); Fonseca, R.O.C. [Steinmann Institut für Geologie, Mineralogie und Paläontologie, University of Bonn, Poppelsdorfer Schloss, 53115 Bonn (Germany); Bottomley, P.D.W.; Hasnaoui, R. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany)

    2013-08-15

    U isotope data measured on real fragments of the Chernobyl nuclear fuel included in zircon crystals crystallised from the Chernobyl “lava” are presented for the first time. The U isotope data show no anomalies and lie within the expected burnup values for the Chernobyl nuclear fuel. However, the U concentration, the U isotopic composition, and the Ti concentration in the host zircon vary significantly within single crystals as well as between single crystals. Our results indicate that during the time of melt activity temperature and melt composition likely varied considerably. New melt was formed progressively (and solidified) during the accident that reacted and mixed with pre-existing melt that never fully equilibrated. In such an environment zircon crystals crystallised at temperatures below 1250 °C, as estimated from thermodynamic considerations along with the observation that the centre of the investigated zircon crystal contains monoclinic ZrO{sub 2} inclusions. Since the zircon crystals crystallised before the silicate melt spread out into the reactor block basement, the flow of the melt into the basement must also have occurred at temperatures below 1250 °C.

  13. Reaction rate distribution measurement and the core performance evaluation in the prototype FBR Monju

    Energy Technology Data Exchange (ETDEWEB)

    Usami, S.; Suzuoki, Z.; Deshimaru, T. [Monju Construction Office, Japan Nuclear Cycle Development Institute, Fukui-ken (Japan); Nakashima, F. [Tsuruga head Office, Japan Nuclear Cycle Development Institute, Fukui-ken (Japan)

    2001-07-01

    Monju is a prototype fast breeder reactor designed to have an output of 280 MW (714 MWt), fueled with mixed oxides of plutonium and uranium and cooled by liquid sodium. The principal data on plant design and performance are shown in Table 1. Monju attained initial criticality in April 1994 and the reactor physics tests were carried out from May through November 1994. The reaction rate distribution measurement by the foil activation method was one of these tests and was carried out in order to verify the core performance and to contribute to the development of the core design methods. On the basis of the reaction rate measurement data, the Monju initial core breeding ratio and the power distribution were evaluated. (author)

  14. Reaction rate distribution measurement and the core performance evaluation in the prototype FBR Monju

    International Nuclear Information System (INIS)

    Usami, S.; Suzuoki, Z.; Deshimaru, T.; Nakashima, F.

    2001-01-01

    Monju is a prototype fast breeder reactor designed to have an output of 280 MW (714 MWt), fueled with mixed oxides of plutonium and uranium and cooled by liquid sodium. The principal data on plant design and performance are shown in Table 1. Monju attained initial criticality in April 1994 and the reactor physics tests were carried out from May through November 1994. The reaction rate distribution measurement by the foil activation method was one of these tests and was carried out in order to verify the core performance and to contribute to the development of the core design methods. On the basis of the reaction rate measurement data, the Monju initial core breeding ratio and the power distribution were evaluated. (author)

  15. Fuels characterization studies. [jet fuels

    Science.gov (United States)

    Seng, G. T.; Antoine, A. C.; Flores, F. J.

    1980-01-01

    Current analytical techniques used in the characterization of broadened properties fuels are briefly described. Included are liquid chromatography, gas chromatography, and nuclear magnetic resonance spectroscopy. High performance liquid chromatographic ground-type methods development is being approached from several directions, including aromatic fraction standards development and the elimination of standards through removal or partial removal of the alkene and aromatic fractions or through the use of whole fuel refractive index values. More sensitive methods for alkene determinations using an ultraviolet-visible detector are also being pursued. Some of the more successful gas chromatographic physical property determinations for petroleum derived fuels are the distillation curve (simulated distillation), heat of combustion, hydrogen content, API gravity, viscosity, flash point, and (to a lesser extent) freezing point.

  16. Analytical chemistry

    Czech Academy of Sciences Publication Activity Database

    Křivánková, Ludmila

    -, č. 22 (2011), s. 718-719 ISSN 1472-3395 Institutional research plan: CEZ:AV0Z40310501 Keywords : analytical chemistry * analytical methods * nanotechnologies Subject RIV: CB - Analytical Chemistry, Separation http://edition.pagesuite-professional.co.uk/launch.aspx?referral=other&pnum=&refresh=M0j83N1cQa91&EID=82bccec1-b05f-46f9-b085-701afc238b42&skip=

  17. Analytic trigonometry

    CERN Document Server

    Bruce, William J; Maxwell, E A; Sneddon, I N

    1963-01-01

    Analytic Trigonometry details the fundamental concepts and underlying principle of analytic geometry. The title aims to address the shortcomings in the instruction of trigonometry by considering basic theories of learning and pedagogy. The text first covers the essential elements from elementary algebra, plane geometry, and analytic geometry. Next, the selection tackles the trigonometric functions of angles in general, basic identities, and solutions of equations. The text also deals with the trigonometric functions of real numbers. The fifth chapter details the inverse trigonometric functions

  18. Quality assurance in vegetable oil fuels. Analytical methods for determining particle size distributions in vegetable oils; Qualitaetssicherung von Pflanzenoelkraftstoffen. Analytik zur Bestimmung der Partikelgroessenverteilung in Pflanzenoelen

    Energy Technology Data Exchange (ETDEWEB)

    Remmele, E. [Bayerische Landesanstalt fuer Landtechnik, Freising-Weihenstephan (Germany); Wanninger, K. [Bayerische Landesanstalt fuer Landtechnik, Freising-Weihenstephan (Germany); Widmann, B. [Bayerische Landesanstalt fuer Landtechnik, Freising-Weihenstephan (Germany); Schoen, H. [Bayerische Landesanstalt fuer Landtechnik, Freising-Weihenstephan (Germany)

    1997-08-01

    Particle size distribution and impurity concentrations are important criteria for characterizing the purity of a vegetable oil. Laser diffraction spectroscopy is a fast and reproducible method of analysis. A special procedure was developed for vegetable oil fuels. First results have shown that cold-pressed, purified rapeseed oils have a lower percentage of particles bigger than 5 {mu}m than fully refined rapeseed products. (orig.) [Deutsch] Die Partikelgroessenverteilung stellt neben der Gesamtverschmutzung ein wesentliches Qualitaetskriterium fuer die Charakterisierung der Reinheit eines Pflanzenoels dar. Mit Hilfe der Laserbeugungsspektroskopie lassen sich sehr schnell und mit hoher Reproduzierbarkeit Partikelgroessenanalysen durchfuehren. Fuer die Qualitaetssicherung von Pflanzenoelkraftstoffen wurde hierzu eine spezielle Probenzufuehrung entwickelt. Erste Ergebnisse zeigen, dass kaltgepresste, gereinigte Rapsoele im Vergleich zu Rapsoelraffinaten einen prozentual geringeren Anteil Partikel groesser 5 {mu}m aufweisen. (orig.)

  19. Analytical Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — The Analytical Labspecializes in Oil and Hydraulic Fluid Analysis, Identification of Unknown Materials, Engineering Investigations, Qualification Testing (to support...

  20. MOX fuel fabrication: Technical and industrial developments

    International Nuclear Information System (INIS)

    Lebastard, G.; Bairiot, H.

    1990-01-01

    The plutonium available in the near future is generally estimated rather precisely on the basis of the reprocessing contracts and the performance of the reprocessing plants. A few years ago, decision makers were convinced that a significant share of this fissile material would be used as the feed material for fast breeder reactors (FBRs) or other advanced reactors. The facts today are that large reprocessing plants are coming into commercial operations: UP3 and soon UP2-800 and THORP, but that FBR deployment is delayed worldwide. As a consequence, large quantities of plutonium will be recycled in light water reactors as mixed oxide (MOX) fuels. MOX fuel technology has been properly demonstrated in the past 25 years. All specific problems have been addressed, efficient fabrication processes and engineering background have been implemented to a level of maturity which makes MOX fuel behaving as well as Uranium fuel. The paper concentrates on todays MOX fabrication expertise and presents the technical and industrial developments prepared by the MOX fuel fabrication industry for this last decade of the century

  1. Reprocessing free nuclear fuel production via fusion fission hybrids

    Energy Technology Data Exchange (ETDEWEB)

    Kotschenreuther, Mike, E-mail: mtk@mail.utexas.edu [Intitute for Fusion Studies, University of Texas at Austin (United States); Valanju, Prashant; Mahajan, Swadesh [Intitute for Fusion Studies, University of Texas at Austin (United States)

    2012-05-15

    Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively 'new' cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th{sup 232}-U{sup 233} conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO{sub 2} matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U{sup 235} fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

  2. Reprocessing free nuclear fuel production via fusion fission hybrids

    International Nuclear Information System (INIS)

    Kotschenreuther, Mike; Valanju, Prashant; Mahajan, Swadesh

    2012-01-01

    Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th 232 –U 233 conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO 2 matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U 235 fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

  3. High temperature structural design and R and Ds for heat transport system components of FBR 'Monju'

    International Nuclear Information System (INIS)

    Sumikawa, Masaharu; Nakagawa, Yukio; Fukuda, Yoshio; Sukegawa, Masayuki; Ishizaki, Tairo.

    1980-01-01

    The machines and equipments of cooling system for the fast breeder prototype reactor ''Monju'' are operated in creep temperature region, and the upper limit temperature to apply the domestic structural design standard for nuclear machines and equipment is exceeded, therefore the guideline for high temperature structural design is being drawn up, reflecting the results of recent research and development, by the Power Reactor and Nuclear Fuel Development Corp. and others. In order to obtain the basic data for the purpose, the tests on the high temperature characteristics of main structural members and structural elements were carried out, and eight kinds of the inelastic structural analysis program ''HI-EPIC'' series were developed, thus the fundamental technologies of structural desigh in non-linear region were established. Also in the non-linear region, enormous physical quantities must be evaluated, and in the design method based on real elastic analysis, many design diagrams must be employed, therefore for the purpose of improving the reliability of evaluation, the automatic evaluation program ''HI-TEP'' was developed, and preparation has been made for the design of actual machines. The high temperature structural design in ''Monju'', the development of inelastic structural analysis program and high temperature structural analysis evaluation program, and the development of high temperature structures and materials are described. (Kako, I.)

  4. Development of tube rupture evaluation code for FBR steam generator (II). Modification of heat transfer model in sodium side

    International Nuclear Information System (INIS)

    Hamada, H.; Kurihara, A.

    2003-05-01

    The thermal effect of sodium-water reaction jet on neighboring heat transfer tubes was examined to rationally evaluate the structural integrity of the tube for overheating rupture under a water leak in an FBR steam generator. Then, the development of new heat transfer model and the application analysis were carried out. Main results in this paper are as follows. (1) The evaluation method of heat flux and heat transfer coefficient (HTC) on the tube exposed to reaction jet was developed. By using the method, it was confirmed that the heat flux could be realistically evaluated in comparison with the previous method. (2) The HTC between reaction jet and the tube was theoretically examined in the two-phase flow model, and new heat transfer model considering the effect of fluid temperature and cover gas pressure was developed. By applying the model, a tentative experimental correlation was conservatively obtained by using SWAT-1R test data. (3) The new model was incorporated to the Tube Rupture Evaluation Code (TRUE), and the conservatism of the model was confirmed by using sodium-water reaction data such as the SWAT-3 tests. (4) In the application analysis of the PFR large leak event, there was no significant difference of calculation results between the new model and previous one; the importance of depressurization in the tube was confirmed. (5) In the application analysis of the Monju evaporator, it was confirmed that the calculation result in the previous model would be more conservative than that in the new one and that the maximum cumulative damage of 25% could be reduced in the new model. (author)

  5. Evaluation of very low frequencies of ATWS and PLOHS in a loop-type FBR plant by making use of inherently safe features

    International Nuclear Information System (INIS)

    Sakata, K.; Koyama, K.; Aoi, S.; Simonelli, R.B.; Wallace, I.T.

    1987-01-01

    Frequencies of ATWS (Anticipated Transient Without Scram) and PLOHS (Protected Loss of Heat Sink) for a large loop-type FBR plant were evaluated by applying PSA methodologies. The frequencies were found to be so low that ATWS and PLOHS could be excluded from candidates of the design basis events. Furthermore, the inherently safe features introduced to the system design were verified to be very effective for reduction of the Probability of CCF (Common Cause Failure), which deteriorates reliability of both the reactor shutdown and the decay heat removal systems. (orig.)

  6. Summary remarks and recommended reactions for an international data file for dosimetry applications for LWR, FBR, and MFR reactor research, development and testing programs

    International Nuclear Information System (INIS)

    McElroy, W.N.; Lippincott, E.P.; Grundl, J.A.; Fabry, A.; Dierckx, R.; Farinelli, U.

    1979-01-01

    The need for the use of an internationally accepted data file for dosimetry applications for light water reactor (LWR), fast breeder reactor (FBR), and magnetic fusion reactor (MFR) research, development, and testing programs continues to exist for the Nuclear Industry. The work of this IAEA meeting, therefore, will be another important step in achieving consensus agreement on an internationally recommended file and its purpose, content, structure, selected reactions, and associated uncertainy files. Summary remarks and a listing of recommended reactions for consideration in the formulation of an ''International Data File for Dosimetry Applications'' are presented in subsequent sections of this report

  7. Analytical functions used for description of the plastic deformation process in Zirconium alloys WWER type fuel rod cladding under designed accident conditions

    International Nuclear Information System (INIS)

    Fedotov, A.

    2003-01-01

    The aim of this work was to improve the RAPTA-5 code as applied to the analysis of the thermomechanical behavior of the fuel rod cladding under designed accident conditions. The irreversible process thermodynamics methods were proposed to be used for the description of the plastic deformation process in zirconium alloys under accident conditions. Functions, which describe yielding stress dependence on plastic strain, strain rate and temperature may be successfully used in calculations. On the basis of the experiments made and the existent experimental data the dependence of yielding stress on plastic strain, strain rate, temperature and heating rate for E110 alloy was determined. In future the following research work shall be made: research of dynamic strain ageing in E635 alloy under different strain rates; research of strain rate influence on plastic strain in E635 alloy under test temperature higher than 873 K; research of deformation strengthening of E635 alloy under high temperatures; research of heating rate influence n phase transformation in E110 and E635 alloys

  8. Analytical method development and validation for quantification of uranium in compounds of the nuclear fuel cycle by Fourier Transform Infrared (FTIR) Spectroscopy

    International Nuclear Information System (INIS)

    Pereira, Elaine

    2016-01-01

    This work presents a low cost, simple and new methodology for direct quantification of uranium in compounds of the nuclear fuel cycle, based on Fourier Transform Infrared (FTIR) spectroscopy using KBr pressed discs technique. Uranium in different matrices were used to development and validation: UO 2 (NO 3 )2.2TBP complex (TBP uranyl nitrate complex) in organic phase and uranyl nitrate (UO 2 (NO 3 ) 2 ) in aqueous phase. The parameters used in the validation process were: linearity, selectivity, accuracy, limits of detection (LD) and quantitation (LQ), precision (repeatability and intermediate precision) and robustness. The method for uranium in organic phase (UO 2 (NO 3 )2.2TBP complex in hexane/embedded in KBr) was linear (r = 0.9980) over the range of 0.20% 2.85% U/ KBr disc, LD 0.02% and LQ 0.03%, accurate (recoveries were over 101.0%), robust and precise (RSD < 1.6%). The method for uranium aqueous phase (UO 2 (NO 3 ) 2 /embedded in KBr) was linear (r = 0.9900) over the range of 0.14% 1.29% U/KBr disc, LD 0.01% and LQ 0.02%, accurate (recoveries were over 99.4%), robust and precise (RSD < 1.6%). Some process samples were analyzed in FTIR and compared with gravimetric and X-ray fluorescence (XRF) analyses showing similar results in all three methods. The statistical tests (t-Student and Fischer) showed that the techniques are equivalent. The validated method can be successfully employed for routine quality control analysis for nuclear compounds. (author)

  9. Analytic geometry

    CERN Document Server

    Burdette, A C

    1971-01-01

    Analytic Geometry covers several fundamental aspects of analytic geometry needed for advanced subjects, including calculus.This book is composed of 12 chapters that review the principles, concepts, and analytic proofs of geometric theorems, families of lines, the normal equation of the line, and related matters. Other chapters highlight the application of graphing, foci, directrices, eccentricity, and conic-related topics. The remaining chapters deal with the concept polar and rectangular coordinates, surfaces and curves, and planes.This book will prove useful to undergraduate trigonometric st

  10. Design study and evaluation of fuel fabrication systems for FR fuel cycle

    International Nuclear Information System (INIS)

    Namekawa, Takashi; Tanaka, Kenya; Kawaguchi, Koichi; Koike, Kazuhiro; Shimuta, Hiroshi; Suzuki, Yoshihiro

    2004-01-01

    The plant concept for each FBR fuel fabrication system has been constructed and evaluated, which achieves economical improvement, decrease in the environmental burden, better resource utilization, and proliferation resistance by the various innovative techniques employed. The results are as follows: (1) For oxide fuels, the simplified pelletizing method has a high technical feasibility, and it is possible to apply this method to practical process at early stage, because this method is based on wealth results of a conventional method. (2) For oxide fuels, the sphere packing fuel fabrication system by gelation and vibro-compaction processes has the advantage of lesser dispersion of the fine powder due to the use of solution and granule in the process. However this system shoulders additional cost for the liquid waste treatment process to dispose a large bulk of process liquid waste. (3) For the metal fuel, the casting system is generally expected to have high economical efficiency even for small-scale facilities, although verification for fabrication of the TRU alloy slug is required. (author)

  11. Development of system based code for integrity of FBR. Fundamental probabilistic approach, Part 1: Model calculation of creep-fatigue damage (Research report)

    International Nuclear Information System (INIS)

    Kawasaki, Nobuchika; Asayama, Tai

    2001-09-01

    Both reliability and safety have to be further improved for the successful commercialization of FBRs. At the same time, construction and operation costs need to be reduced to a same level of future LWRs. To realize compatibility among reliability, safety and, cost, the Structural Mechanics Research Group in JNC started the development of System Based Code for Integrity of FBR. This code extends the present structural design standard to include the areas of fabrication, installation, plant system design, safety design, operation and maintenance, and so on. A quantitative index is necessary to connect different partial standards in this code. Failure probability is considered as a candidate index. Therefore we decided to make a model calculation using failure probability and judge its applicability. We first investigated other probabilistic standards like ASME Code Case N-578. A probabilistic approach in the structural integrity evaluation was created based on these results, and also an evaluation flow was proposed. According to this flow, a model calculation of creep-fatigue damage was performed. This trial calculation was for a vessel in a sodium-cooled FBR. As the result of this model calculation, a crack initiation probability and a crack penetration probability were found to be effective indices. Last we discussed merits of this System Based Code, which are presented in this report. Furthermore, this report presents future development tasks. (author)

  12. Creep-fatigue life property of FBR high-temperature structural materials under tension-torsion loading and life evaluation method

    International Nuclear Information System (INIS)

    Ogata, Takashi; Nitta, Akito

    1994-01-01

    Creep-fatigue damage in high temperature structural components in a FBR progress under multiaxial stress condition depending on their operating conditions and configuration. Therefore, multiaxial stress effects on creep-fatigue damage evolution must be clarified to make precise creep-fatigue damage evaluation of these components. In this study, creep-fatigue tests in FBR high temperature materials such as SUS304, 316FR stainless steels and a modified 9Cr steel were conducted under biaxial stress subjecting tension-compression and torsion loading, in order to examine biaxial stress effects on failure mechanism and life property, and to discuss creep-fatigue life evaluation methods under biaxial stress. Main results obtained in this study are summarized as follows: 1. The main cracks under cyclic torsion loading propagated by shear mode in three materials. But intergranular failure was occurred in SUS304 and 316FR, and transgranular failure was observed in Mod.9Cr steel. 2. Nonlinear damage accumulation model proposed based on uniaxial creep-fatigue test results was extended to apply for creep-fatigue damage evaluation under biaxial stress state by considering the biaxial stress effects on fatigue and creep damage evolution. 3. It was confirmed that creep-fatigue life under biaxial stress could be predicted by the extended evaluation method with higher accuracy than existing methods. (author)

  13. Novel analytical technique to study nucleobase influence on DNA strand breaks caused by direct ionizing radiation

    International Nuclear Information System (INIS)

    Watson, R.M.; Bernhard, W.A.

    2009-01-01

    Complete text of publication follows. Analysis of the reactions involved in the direct effect of ionizing radiation on DNA is crucial to assessing the risks related with exposure at low dose. The direct interaction of ionizing radiation with DNA initially results in free radicals situated on bases and the backbone, which eventually lead to stable end products that include strand breaks (sb) and free nucleobase release (fbr). The yields of these two products are thought to be related because ejection of an electron from the DNA backbone produces a radical cation that deprotonates to yield a neutral carbon-centered deoxyribose radical. These neutral radicals react when dissolved to produce one strand break and one free base each. Therefore fbr can be used as an indicator of sb. It is commonly presumed that that sb occur independent of the surrounding base context. However recent studies have indicated that a base may indeed have influence over the probability of sb at its backbone unit. In one such study, films prepared from 10- to 30-mer DNA duplexes were irradiated at RT under air using X-rays generated by a tungsten tube operated at 70 kV. The films were dissolved in nuclease free water and stored at 277 K. Unaltered free base release was measured using HPLC, and the yields determined for each base were not strictly proportionate to their presence in the DNA sequence. In fact, this study indicated that strand breaks may be influenced by a number of factors including position within the oligomer as well as the base and its base context. The current study involves further analysis of these factors; instead of using HPLC to separate and measure fbr, which is time consuming and expensive, a novel analytical technique is being used to determine the amount and ratio of fbr for each of the four bases. This technique involves separation of free bases from bulk DNA using filters followed by decomposition of the UV spectra of mixtures of bases at different pH. Decomposition

  14. Analytical quadrics

    CERN Document Server

    Spain, Barry; Ulam, S; Stark, M

    1960-01-01

    Analytical Quadrics focuses on the analytical geometry of three dimensions. The book first discusses the theory of the plane, sphere, cone, cylinder, straight line, and central quadrics in their standard forms. The idea of the plane at infinity is introduced through the homogenous Cartesian coordinates and applied to the nature of the intersection of three planes and to the circular sections of quadrics. The text also focuses on paraboloid, including polar properties, center of a section, axes of plane section, and generators of hyperbolic paraboloid. The book also touches on homogenous coordi

  15. Development of a standard database for FBR core nuclear design (XI). Analysis of the Experimental Fast Reactor 'JOYO' MK-I start-up test and operation data

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2000-03-01

    As a recent research, Japan Nuclear Cycle Development Institute (JNC) develops a database of integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor 'JOYO' MK-I core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. On the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of 'JOYO' MK-I core in comparison with ZPPR-9 core of JUPITER experiments. (J.P.N.)

  16. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  17. Investigation of small scale sphere-pac fuel fabrication plant with external gelation process

    International Nuclear Information System (INIS)

    Maekawa, Kazuhiko; Yoshimura, Tadahiro; Kikuchi, Toshiaki; Hoshino, Yasushi; Munekata, Hideki; Shimizu, Makoto

    2005-02-01

    In feasibility studies on commercialized FBR cycle system, comprehensive system investigation and properties evaluation for candidate FBR cycle systems have been implemented through view point of safety, economics, environmental burden reduction, non-proliferation resistivity, etc. As part of these studies, an investigation of small scale sphere-pac fuel fabrication plant with external gelation process was conducted. Until last fiscal year, equipment layout in cells and overall layout design of the 200t-HM/y scale fuel fabrication plant were conducted as well as schematical design studies on main equipments in gelation and reagent recovery processes of the plant. System property data concerning economics and environmental burden reduction of fuel fabrication plant was also acquired. In this fiscal year, the processes from vibropacking to fuel assemblies storage were added to the investigation range, and a conceptual design of whole fuel fabrication plant was studied as well as deepening the design study on main equipments. The conceptual design study was mainly conducted for small 50t-HM/y scale plant and a revising investigation was done for 200t-HM/y scale plant. Taking the planed comparative evaluation with pellet fuel fabrication system into account, design of equipments which should be equivalent with pellet system, especially in post-vibropacking processes, were standardized in each system. Based on these design studies, system properties data concerning economics and environmental burden reduction of the plant was also acquired. In comparison with existing design, the cell height was lowered on condition that plug type pneumatic system was adopted and fuel fabrication building was downsized by applying rationalized layout design of pellet system to post-vibropacking processes. Reduction of reagent usage at gelation process and rationalization of sintering and O/M controlling processes etc., are foremost tasks. (author)

  18. Applications of superconductivity to nuclear fuel cycle

    International Nuclear Information System (INIS)

    Sasao, Nobuyuki; Kubota, Jun

    1988-01-01

    As the application of superconductivity in nuclear fuel cycle, the plasma process of uranium enrichment, the magnetic separation techniques for fuel reprocessing, waste treatment and so on, and the application of liquid metal MHD to FBRs are explained. Besides, the investigation of rare earth which is the main elements of oxide superconductive materials in the aspect of resources, and the examination of the possibility of actinide superconductive materials including uranium which is a nuclear fuel material are carried out. Through these studies, it was found that by the adoption of superconductivity, that which receives the economical and technical favors most is nuclear power. Nuclearfuel creates rare earth by nuclear fission reaction when it burns in a reactor, and there is the possibility that it becomes the creation of valuable resources for Japan where natural resources are short. The uranium enrichment by the isotope separation using plasma electromagnetic effect was examined in USA, but stopped. Magnetic separation utilizes the gradient of a magnetic field to separate superfine particles, and many applications are conceivable. In the case of liquid metal MHD, the electric conductivity is very high, accordingly the flow velocity and fluid temperature may be relatively low. The development of a superconductive electromagnetic pump for a FBR is discussed. (Kako, I.)

  19. FBR type reactor

    International Nuclear Information System (INIS)

    Nagai, Fumio.

    1979-01-01

    Purpose: To unify the temperature distribution in a nuclear reactor vessel by the provision of a gas recycle path for pressurizing a cover gas to recycle the cover gas and thus stir the gas in a cover gas chamber. Constitution: A plurality of gas inlet tubes and gas discharge tubes are provided to the wall of a cover gas chamber above the liquid level of coolants in a nuclear reactor vessel and the cover gas is recycled through the tubes. The plurality of gas inlet tubes are each provided at their tops with nozzles opening circumferentially and communicated to the outlet of a compressor. While on the other hand, the plurality of gas discharge tubes are communicated to the inlet of a compressor. Upon operation of the compressor, the pressurized cover gas is jetted out from the nozzles, swirls along the inner circumferential surface of the vessel and interrupts and stirs the vertical thermal convection. The gas, after swirling one-half of the inner circumferential surface of the vessel, automatically flows out of the gas discharging tubes opening behind the nozzles and then flows into the inlet of the compressor. (Seki, T.)

  20. FBR type reactors

    International Nuclear Information System (INIS)

    Maemoto, Junko.

    1985-01-01

    Purpose: To moderate abrupt temperature change near the inner walls of a suspended body thereby prevent thermal shocks and thermal deformations to structural materials. Constitution: High temperature coolants during ordinary operation of the nuclear reactor flow from the reactor core through the flow holes of the suspended body and from the upper plenum into an intermediate heat exchanger. The temperature of the coolants is lowered with heat exchanging effect with secondary coolants in the heat exchange and the coolants are then flow through the lower plenum into the reactor core and heated again. Upon generation of reactor scram, the temperature of the coolants at the exit of the reactor core is reduced abruptly and the flow rate is lowered due to the pump coast down. However, mixing of the coolants in the suspended body is accelerated by the coolants at high temperature flowing out of the flow holes and the coolants at the low temperature flowing from the flow hole group, to reduce the temperature difference and moderate the stratification flow forming an abrupt temperature slope. (Yoshihara, H.)

  1. Reports of the 8th new type nuclear fuel materials studying meeting. Present status of the plutonium mixed oxide fuel application

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-05-01

    This was the reports of the 8th New Type Nuclear Fuel Materials Studying Meeting, as a circle of Yayoi Studying Group meeting held on March 17, 1997. This meeting was added to a subtitle of `Present status and problems of plutonium mixed oxide application`, which had 12 lectures. In this meeting, for the MOX fuels putting the most attention in the field of nuclear fuel development at present, many specialists introduced faithfully on present status and problems of its nuclear features, reactor core design, and application to light water reactor and fast reactor. And, following reports were executed: (A) On feature of plutonium and reactor core design; (1) nuclear feature of plutonium, (2) nuclear design of BWR, (3) nuclear design of PWR, (4) nuclear design of FBR, and (5) and (6) properties of the MOX fuel; (B) On application of plutonium to the light water reactor; (1) preparation of the MOX fuel for light water reactor, (2) radiation behavior and using result of the MOX fuel for BWR, and (3) radiation behavior and using result of the MOX fuel for PWR; and (C) On application of plutonium to the fast reactor; (1) fuel preparation, (2) radiation behavior, and (3) reprocessing of the fast reactor fuel. (G.K.)

  2. Analytical chemistry

    International Nuclear Information System (INIS)

    Choi, Jae Seong

    1993-02-01

    This book is comprised of nineteen chapters, which describes introduction of analytical chemistry, experimental error and statistics, chemistry equilibrium and solubility, gravimetric analysis with mechanism of precipitation, range and calculation of the result, volume analysis on general principle, sedimentation method on types and titration curve, acid base balance, acid base titration curve, complex and firing reaction, introduction of chemical electro analysis, acid-base titration curve, electrode and potentiometry, electrolysis and conductometry, voltammetry and polarographic spectrophotometry, atomic spectrometry, solvent extraction, chromatograph and experiments.

  3. Analytical chemistry

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    The division for Analytical Chemistry continued to try and develope an accurate method for the separation of trace amounts from mixtures which, contain various other elements. Ion exchange chromatography is of special importance in this regard. New separation techniques were tried on certain trace amounts in South African standard rock materials and special ceramics. Methods were also tested for the separation of carrier-free radioisotopes from irradiated cyclotron discs

  4. The secondary stress analyses in the fuel pin cladding due to the swelling gradient across the wall thickness

    International Nuclear Information System (INIS)

    Uwaba, Tomoyuki; Ukai, Shigeharu

    2002-01-01

    Irradiation deformation analyses of FBR fuel cladding were made by using the finite element method. In these analyses the history of the stress occurred in the cladding was evaluated paying attention to the secondary stress induced by the swelling difference across the wall thickness. It was revealed that the difference of the swelling incubation dose in the direction of the thickness and the irradiation creep deformation play an important role in the history of the secondary stress. The effect of the stress-enhanced swelling was also analyzed in this study

  5. Video Analytics

    DEFF Research Database (Denmark)

    Nasrollahi, Kamal; Distante, Cosimo; Hua, Gang

    2017-01-01

    This book collects the papers presented at two workshops during the 23rd International Conference on Pattern Recognition (ICPR): the Third Workshop on Video Analytics for Audience Measurement (VAAM) and the Second International Workshop on Face and Facial Expression Recognition (FFER) from Real...... World Videos. The workshops were run on December 4, 2016, in Cancun in Mexico. The two workshops together received 13 papers. Each paper was then reviewed by at least two expert reviewers in the field. In all, 11 papers were accepted to be presented at the workshops. The topics covered in the papers...

  6. Video Analytics

    DEFF Research Database (Denmark)

    This book collects the papers presented at two workshops during the 23rd International Conference on Pattern Recognition (ICPR): the Third Workshop on Video Analytics for Audience Measurement (VAAM) and the Second International Workshop on Face and Facial Expression Recognition (FFER) from Real...... World Videos. The workshops were run on December 4, 2016, in Cancun in Mexico. The two workshops together received 13 papers. Each paper was then reviewed by at least two expert reviewers in the field. In all, 11 papers were accepted to be presented at the workshops. The topics covered in the papers...

  7. Fossil fuels -- future fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    Fossil fuels -- coal, oil, and natural gas -- built America`s historic economic strength. Today, coal supplies more than 55% of the electricity, oil more than 97% of the transportation needs, and natural gas 24% of the primary energy used in the US. Even taking into account increased use of renewable fuels and vastly improved powerplant efficiencies, 90% of national energy needs will still be met by fossil fuels in 2020. If advanced technologies that boost efficiency and environmental performance can be successfully developed and deployed, the US can continue to depend upon its rich resources of fossil fuels.

  8. Synergistic fuel cycles of the future

    International Nuclear Information System (INIS)

    Meneley, D.A.; Dastur, A.R.

    1997-01-01

    Good neutron economy is the basis of the fuel cycle flexibility in the CANDU reactor. This paper describes the fuel cycle options available to the CANDU owner with special emphasis on resource conservation and waste management. CANDU fuel cycles with low initial fissile content operate with relatively high conversion ratio. The natural uranium cycle provides over 55 % of energy from the plutonium that is created during fuel life. Resource utilization is over 7 MWd/kg NU. This can be improved by slight enrichment (between 0.9 and 1.2 wt % U235) of the fuel. Resource utilization increases to 11 MWd/kg NU with the Slightly Enriched Uranium cycle. Thorium based cycles in CANDU operate at near-breeder efficiency. Obey provide attractive options when used with natural uranium or separated (reactor grade and weapons grade) plutonium as driver fuels. In the latter case, the energy from the U233 plus the initial plutonium content amounts to 3.4 GW(th).d/kg Pu-fissile. The same utilization is expected from the use of FBR plutonium in a CANDU thorium cycle. Extension of natural resource is achieved by the use of spent fuels in CANDU. The LWR/CANDU Tandem cycle leads to an additional 77 % of energy through the use of reprocessed LWR fuel (which has a fissile content of 1.6 wt %) in CANDU. Dry reprocessing of LWR fuel with the OREOX process (a more safeguardable alternative to the PUREX process) provides an additional 50 % energy. Uranium recovered (RU) from separation of plutonium contained in spent LWR fuel provides an additional 15 MWd/kg RU. CANDU's low fissile requirement provides the possibility, through the use of non-fertile targets, of extracting energy from the minor actinides contained in spent fuel. In addition to the resource utilization advantage described above, there is a corresponding reduction in waste arisings with such cycles. This is especially significant when separated plutonium is available as a fissile resource. (author)

  9. Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

    International Nuclear Information System (INIS)

    Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

    2005-10-01

    This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor

  10. Fuel assemblies

    International Nuclear Information System (INIS)

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

  11. Analytical mechanics

    CERN Document Server

    Helrich, Carl S

    2017-01-01

    This advanced undergraduate textbook begins with the Lagrangian formulation of Analytical Mechanics and then passes directly to the Hamiltonian formulation and the canonical equations, with constraints incorporated through Lagrange multipliers. Hamilton's Principle and the canonical equations remain the basis of the remainder of the text. Topics considered for applications include small oscillations, motion in electric and magnetic fields, and rigid body dynamics. The Hamilton-Jacobi approach is developed with special attention to the canonical transformation in order to provide a smooth and logical transition into the study of complex and chaotic systems. Finally the text has a careful treatment of relativistic mechanics and the requirement of Lorentz invariance. The text is enriched with an outline of the history of mechanics, which particularly outlines the importance of the work of Euler, Lagrange, Hamilton and Jacobi. Numerous exercises with solutions support the exceptionally clear and concise treatment...

  12. SAF line analytical chemistry system

    International Nuclear Information System (INIS)

    Gerber, E.W.; Sherrell, D.L.

    1983-10-01

    An analytical chemistry system dedicated to supporting the Secure Automated Fabrication (SAF) line is discussed. Several analyses are required prior to the fuel pellets being loaded into cladding tubes to assure certification requirements will be met. These analyses, which will take less than 15 minutes, are described. The automated sample transport system which will be used to move pellets from the fabriction line to the chemistry area is also described

  13. Alternative fuels for the French fast breeder reactors programme

    International Nuclear Information System (INIS)

    Bailly, H.; Bernard, H.; Mansard, B.

    1988-01-01

    At the present time, due to the very competitive cost per kWh produced in France by the PWRs, it appears clear that, despite the improved use of uranium by FBRs, they will only be developed if the cost of the fuel cycle is sufficiently lower than that of the PWRs to compensate for the additional investment. The current economic programme has fixed the following fuel related objectives: - burn-up as high possible, the value of 150 000 MWd/t being considered as a minimum, and not a final target to be achieved, - extension of the duration of reactor operation cycles, leading to high in-pile times for fuel. Reaching the latter objective depends on obtaining high internal breeding gain performances, so that the total reactivity drop related to fuel impoverishment can be minimized. In this respect, a large diameter oxide fuel and/or an axial heterogeneous core concept can be envisaged. Dense fuels could form another solution. The feasibility of the fabrication of carbide and nitride fuels has been demonstrated in several countries and there is currently convergence towards a single type of process based on a carbothermic reaction. The optimization of fabrication procedures for these fuels must be continued to satisfy economic requirements and to obtain a fabrication cost of the same order or magnitude as that of oxide, although higher. If this target is achieved, fabrication will not be the major criterion for the selection of the FBR fuel, which will then be a function of the cost of reprocessing, performances under irradiation and reactor operating requirements

  14. Fuel safety research 2001

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-11-01

    The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

  15. Nuclear power plant types and the management of plutonium and minor actinides - in search of fuel cycle flexibility

    International Nuclear Information System (INIS)

    Thomas, J.B.

    2002-01-01

    Transuranics management concerns all NPP types, because of the specifications for sustainable development. Multiple recycling is mandatory. Neutronic abundance can be obtained in fast spectrum, or by adding external neutrons or (temporarily) with additional 235 U. The LWRs can control the plutonium inventory and significantly reduce the amount of transuranics transferred to the geological repository, thanks to the use of innovative nuclear fuel in a limited part of the NPP fleet. HTR adapted to transuranics burning can help. In the future, in addition to the liquid metal FBR, a strategy based on a gas cooled technological line and advanced fuel opens a second path towards fast spectra. Strategies for defining the optimal mix of reactor types in the nuclear fleet at a given time and demonstrating the fuel cycle flexibility are under study. (author)

  16. A practical approach in reliability analysis for mechanical systems: application to irradiated fuel reprocessing plant

    International Nuclear Information System (INIS)

    Sejourne, S.; Humbert, J.M.; Greneche, D.

    1986-10-01

    Design and tests of equipments for the dismantling of fuel assemblies for LWR or FBR in fuel reprocessing plants are studied for reliability. Three examples are given: 1. Qualitative analysis with fault trees and FMEA for a full equipment. This study enabled to propose preventive actions in order to reduce the severity or the probability of a failure. 2. Calculation of the probability of a failure for a part of a mechanical device. This study enabled to show that this part has a very high probability at failure and then, that it is necessary to foresee the possibility of repair. 3. Estimation of the availability of on equipment. Calculations were performed from test results on various prototypes allowing comparisons between them and allowing to verify if expected availability was reached [fr

  17. Does rim microstructure formation degrade the fuel rod performance?

    International Nuclear Information System (INIS)

    Baron, D.; Spino, J.

    2002-01-01

    High burnup extension of LWR fuel is progressing to reduce the total process flow and eventually the costs of the nuclear fuel cycle. A particular fuel restructuring at high burnups, commonly observed at the periphery of LWR fuel pellets (rim structure), but also in FBR fuels to some extent and in the Plutonium rich clusters of the MOX Fuels, was considered a priori as a limitation for burnup extension. Since more than ten years this rim effect have been deeply investigated. Its causes and consequences are however not yet totally elucidated. The three steps actually identified of this phenomenon are first a progressive disappearing of the intra-granular Xenon, the outset of numerous 0.5 to 1 m pores and finally a grain subdivision around the pores. Penalty of the porosity increase on the thermal conductivity is obvious. One expect the fission gases to remain trapped in the rim porosity up to a 75 MWd/kgUO 2 local burnup. Above this threshold, 15 to 20 % of the fission gases seem to be quickly released. Microindentation tests conducted at ITU have shown the rim structure to resist fracture extension under punching. It is still open whether this implies certain ductility and viscosity of the material, or if it corresponds to stress relaxation by microcracking. Whatever the case be, it is suggested that the rim material would be able to decrease the interaction stresses and to equalise the cladding strains during a power ramp. Moreover, in the RIA tests, it was concluded so far that the grain de-cohesion caused by gas expansion at the grain boundaries was responsible for the cladding strain and failure. However, not the rim zone was affected by grain de-cohesion but the region adjacent to it. Therefore, in front of the question whether the rim structure degrades the fuel rod behaviour, we continue to argue on its benefit for fuel burnup extension. (author)

  18. Alternative Fuels

    Science.gov (United States)

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  19. Study on Doppler coefficient for metallic fuel fast reactor added hydrogeneous moderator

    Energy Technology Data Exchange (ETDEWEB)

    Hirakawa, Naohiro; Iwasaki, Tomohiko; Tsujimoto, Kazuhumi [Tohoku Univ., Sendai (Japan). Faculty of Engineering; Osugi, Toshitaka; Okajima, Shigeaki; Andoh, Masaki; Nemoto, Tatsuo; Mukaiyama, Takehiko

    1998-01-01

    A series of mock-up experiments for moderator added metallic fast reactor core was carried out at FCA to obtain the experimental verification for improvement of reactivity coefficients. Softened neutron spectrum increases Doppler effect by a factor of 2, and flatter adjoint neutron spectrum decreases Na void effect by a factor of 0.6 when hydrogen to heavy metal atomic number ratio is increased from 0.02 to 0.13. The experimental results are analyzed with SLALOM and CITATION-FBR, which is the standard design code system for a fast reactor at JAERI, and SRAC95 and CITATION-FBR. The present code system gives generally good agreement with the experimental results, especially by the use of the latter, the dependence of the Doppler effect to the hydrogen to fuel element atomic number density ratio is disappeared. Therefore, it looks possible to use the present code system for the conceptual design of a fast reactor system with hydrogeneous materials. (author)

  20. Introduction to the study of the treatment of spent nuclear fuel

    International Nuclear Information System (INIS)

    Floh, B.; Araujo, J.A. de; Matsuda, H.T.

    1975-01-01

    An introduction is made to the study of the treatment of spent nuclear fuels. Main topics discussed are: basic information, volatilization processes, treatment of thorium based fuels (Thorex process), analytical chemistry of spent nuclear fuel and design of industrial facilities

  1. Fuel assembly

    International Nuclear Information System (INIS)

    Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

    1998-01-01

    The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

  2. Fuel cells

    Science.gov (United States)

    Hooie, D. T.; Harrington, B. C., III; Mayfield, M. J.; Parsons, E. L.

    1992-07-01

    The primary objective of DOE's Fossil Energy Fuel Cell program is to fund the development of key fuel cell technologies in a manner that maximizes private sector participation and in a way that will give contractors the opportunity for a competitive posture, early market entry, and long-term market growth. This summary includes an overview of the Fuel Cell program, an elementary explanation of how fuel cells operate, and a synopsis of the three major fuel cell technologies sponsored by the DOE/Fossil Energy Phosphoric Acid Fuel Cell program, the Molten Carbonate Fuel Cell program, and the Solid Oxide Fuel Cell program.

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

    In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

  4. Al-Mn CVD-FBR coating on P92 steel as protection against steam oxidation at 650 °C: TGA-MS study

    Science.gov (United States)

    Castañeda, S. I.; Pérez, F. J.

    2018-02-01

    The initial stages oxidation of the P92 ferritic/martensitic steel with and without Al-Mn coating at 650 °C in Ar+40%H2O for 240 h were investigated by mass spectrometry (MS) and thermogravimetric analysis (TGA). TGA-MS measurements were conducted in a closed steam loop. An Al-Mn coating was deposited on P92 steel at 580 °C for 2 h by chemical vapour deposition in a fluidized bed reactor (CVD-FBR). The coating as-deposited was treated in the same reactor at 700 °C in Ar for 2h, in order to produce aluminide phases that form the protective alumina layer (Al2O3) during oxidation. MS measurements at 650 °C of the Al-Mn/P92 sample for 200 h indicated the presence of (Al-Mn-Cr-Fe-O) volatile species of small intensity. Uncoated P92 steel oxidized under the same steam oxidation conditions emitted greater intensities of volatile species of Cr, Fe and Mo in comparison with intensities from coated steel. TGA measurements verified that the mass gained by the coated sample was up to 300 times lower than for uncoated P92 steel. The morphology, composition and structure of samples by Scanning Electron Microscopy SEM, Backscattered Electron (BSE) detection, X-ray Energy Dispersive Spectrometry (EDAX) and X-ray Diffraction (XRD) are described.

  5. X-ray excited optical luminescence, photoluminescence, photostimulated luminescence and x-ray photoemission spectroscopy studies on BaFBr:Eu

    CERN Document Server

    Subramanian, N; Govinda-Rajan, K; Mohammad-Yousuf; Santanu-Bera; Narasimhan, S V

    1997-01-01

    The results of x-ray excited optical luminescence (XEOL), photoluminescence (PL), photostimulated luminescence (PSL) and x-ray photoemission spectroscopy (XPS) studies on the x-ray storage phosphor BaFBr:Eu are presented in this paper. Analyses of XEOL, PL and PSL spectra reveal features corresponding to the transitions from 4f sup 6 td sup 1 to 4f sup 7 configurations in different site symmetries of Eu sup 2 sup +. Increasing x-ray dose is seen to lead to a red shift in the maximum of the PL excitation spectrum for the 391 nm emission. The XEOL and XPS spectra do not show any signature of Eu sup 3 sup + in the samples studied by us, directly raising doubts about the model of Takahashi et al in which Eu sup 2 sup + is expected to ionize to Eu sup 3 sup + upon x-ray irradiation and remain stable until photostimulation. XEOL and PSL experiments with simultaneous x-ray irradiation and He - Ne laser excitation as well as those with sequential x-ray irradiation and laser stimulation bring out the competition betwe...

  6. Analysis of transient fuel failure mechanisms: selected ANL programs

    International Nuclear Information System (INIS)

    Deitrich, L.W.

    1975-01-01

    Analytical programs at Argonne National Laboratory related to fuel pin failure mechanisms in fast-reactor accident transients are described. The studies include transient fuel pin mechanics, mechanics of unclad fuel, and mechanical effects concerning potential fuel failure propagation. (U.S.).

  7. International symposium on fuel rod simulators: development and application

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W. (comp.)

    1981-05-01

    Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

  8. Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

    International Nuclear Information System (INIS)

    Fukaya, Y.; Okubo, T.; Uchikawa, S.

    2008-01-01

    The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the 241 Pu content in the initial fuel, and the decay heat mainly depends on 238 Pu and 244 Cm. The contribution of 244 Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from

  9. Nuclear fuels

    International Nuclear Information System (INIS)

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

    Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

  10. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

  11. Visual in-pile fuel disruption experiments

    International Nuclear Information System (INIS)

    Cano, G.L.; Ostensen, R.W.; Young, M.F.

    1978-01-01

    In a loss-of-flow (LOF) accident in an LMFBR, the mode of disruption of fuel may determine the probability of a subsequent energetic excursion. To investigate these phenomena, in-pile disruption of fission-heated irradiated fuel pellets was recorded by high speed cinematography. Instead of fuel frothing or dust-cloud breakup (as used in the SAS code) massive and very rapid fuel swelling, not predicted by analytical models, occurred. These tests support massive fuel swelling as the initial mode of fuel disruption in a LOF accident. (author)

  12. Hanford analytical sample projections 1996--2001

    Energy Technology Data Exchange (ETDEWEB)

    Joyce, S.M. [Westinghouse Hanford Co., Richland, WA (United States)

    1996-06-26

    This document summarizes the biannual Hanford sample projections for fiscal years 1996 to 2001. Sample projections are based on inputs submitted to Analytical Services covering Environmental Restoration, Tank Waste Remediation Systems (TWRS), Solid Waste, Liquid Effluents, Spent Nuclear Fuels, Transition Projects, Analytical Services, Site Monitoring, and Industrial Hygiene. This information will be used by Hanford Analytical Services to assure that laboratories and resources are available and effectively utilized to meet these documented needs. Sample projections are categorized by radiation level, protocol, sample matrix and Program. Analyses requirements are also presented.

  13. Analytical chemistry in a new analytical hot cell facility

    International Nuclear Information System (INIS)

    Wade, M.A.; Dykes, F.W.; Goettsche, J.H.

    1985-01-01

    The Remote Analytical Laboratory is a new facility at the Idaho Chemical Processing Plant designed to handle samples from the processing of spent nuclear fuel. It consists of a cold laboratory for analyzing process make-up samples, a warm laboratory for analyzing low-level (<100 mR/h) radioactive samples, and a hot cell for analyzing high-level radioactive samples. The hot cell is built in an L shape and contains six work stations, each equipped with a viewing window and two master/slave manipulators. The cell interfaces with a waste handling cell and maintenance area on one end and a glove box complex that interfaces with the warm laboratory on the other end. This paper discusses the remote analytical techniques and equipment developed for use in this facility

  14. Advanced PWR fuel design concepts

    International Nuclear Information System (INIS)

    Andersor, C.K.; Harris, R.P.; Crump, M.W.; Fuhrman, N.

    1987-01-01

    For nearly 15 years, Combustion Engineering has provided pressurized water reactor fuel with the features most suppliers are now introducing in their advanced fuel designs. Zircaloy grids, removable upper end fittings, large fission gas plenum, high burnup, integral burnable poisons and sophisticated analytical methods are all features of C-E standard fuel which have been well proven by reactor performance. C-E's next generation fuel for pressurized water reactors features 24-month operating cycles, optimal lattice burnable poisons, increased resistance to common industry fuel rod failure mechanisms, and hardware and methodology for operating margin improvements. Application of these various improvements offer continued improvement in fuel cycle economics, plant operation and maintenance. (author)

  15. Achieving High Burnup Targets With Mox Fuels: Techno Economic Implications

    International Nuclear Information System (INIS)

    Clement Ravi Chandar, S.; Sivayya, D.N.; Puthiyavinayagam, P.; Chellapandi, P.

    2013-01-01

    For a typical MOX fuelled SFR of power reactor size, Implications due to higher burnup have been quantified. Advantages: – Improvement in the economy is seen upto 200 GWd/ t; Disadvantages: – Design changes > 150 GWd/ t bu; – Need for 8/ 16 more fuel SA at 150/ 200 GWd/ t bu; – Higher enrichment of B 4 C in CSR/ DSR at higher bu; – Reduction in LHR may be required at higher bu; – Structural material changes beyond 150 GWd/ t bu; – Reprocessing point of view-Sp Activity & Decay heat increase. Need for R & D is a must before increasing burnup. bu- refers burnup. Efforts to increase MOX fuel burnup beyond 200 GWd/ t may not be highly lucrative; • MOX fuelled FBR would be restricted to two or four further reactors; • Imported MOX fuelled FBRs may be considered; • India looks towards launching metal fuel FBRs in the future. – Due to high Breeding Ratio; – High burnup capability

  16. Mid-Term Direction of JAEA Nuclear Fuel Cycle Engineering Laboratories

    International Nuclear Information System (INIS)

    Ojima, H.; Sugiyama, T.; Tanaka, K.; Takeda, S.; Nomura, S.

    2009-01-01

    1. Introduction Nuclear Fuel Cycle Engineering Laboratories (NCL) of Japan Atomic Energy Agency (JAEA) has sufficient experience and ability through its 50 year operation to establish the next generation closed cycle. It strives to become a world-class Center Of Excellence. 2. Current activity in NCL: 1) - Recycling of MOX fuel: The Tokai Reprocessing Plant has reprocessed 29 tons of MOX fuel from the ATR Fugenh as a part of 1140 tons of cumulative spent fuel reprocessed. JAEA has supported the pre-operation of the Rokkasho Reprocessing Plant. An innovative MOX pellet fabrication process has been developed in the Plutonium Fuel Development Center, and a part of products obtained by the development are used as a fuel for core confirmation test for re-startup of the FBR Monjuh. Characterization of MOX containing Am and Np has been studied and a new data such as melting point and thermal conductivity were reported. In the Chemical Processing Facility, a hot lab., an advanced aqueous reprocessing technology has been tested for TRU recovery, economical improvement, etc., using irradiated MOX fuel from the FR Joyoh. 2) - Supporting Activity: JAEA has improved the effectiveness and efficiency of existing safeguards activities. The Integrated Safeguards approach for all facilities in NCL has been implemented since August, 2008, as a pioneer and a good example in the world. To reduce anxiety among local residents, NCL has explained its operation plans and exchanged information and opinions with them concerning potential risks to health and environment. Recently, stake-holder participation in the management of NCL was started from the view point of Corporate Social Responsibility. In April, 2008, the agreement was signed with Idaho National Laboratory for cooperation of personnel training in fuel cycle area. 3. Mid-Term Direction: In Japan, feasibility and direction of the transition period from the LWR era to the FBR era should be discussed for the next several years. Study

  17. Modeling and Analytical Simulation of a Smouldering ...

    African Journals Online (AJOL)

    ADOWIE PERE

    ABSTRACT: Modeling of pyrolysis and combustion in a smouldering fuel bed requires the solution of flow, heat and mass transfer through porous media. ..... eAt v g. RT. E gp φρ ρ σ. −. = Analytical Solution. We solve equations (10) – (14) using parameter- expanding method (where details can be found in. (He, 2006)) and ...

  18. Investigation of flow stabilization in a compact reactor vessel of a FBR. Flow visualization in a reactor vessel

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Igarashi, Minoru; Kimura, Nobuyuki; Kamide, Hideki

    2002-01-01

    In the feasibility studies of Commercialized Fast Breeder Reactor Cycle System, a compact reactor vessel is considered from economical improvement point of a sodium cooled loop type fast reactor. The flow field was visualized by water experiment for a reactor vessel with 'a column type UIS (Upper Internal Structure)', which has a slit for fuel handling mechanism and is useful for a compact fast reactor. In this research, the 1/20 scale test equipment using water was made to understand coolant flow through a slit of a column type UIS' and fundamental behavior of reactor upper plenum flow. In the flow visualization tests, tracer particles were added in the water, and illuminated by the slit-shaped pulse laser. The flow visualization image was taken with a CCD camera. We obtained fluid velocity vectors from the visualization image using the Particle Imaging Velocimetry (PIV). The results are as follows. 1. Most of coolant flow through a slit of 'column type UIS' arrived the dip plate directly. In the opposite side of a slit, most of coolant flowed toward reactor vessel wall before it arrived the dip plate. 2. The PIV was useful to measure the flow field in the reactor vessel. The obtained velocity field was consistent with the flow visualization result. 3. The jet through the UIS slit was dependent on the UIS geometry. There is a possibility to control the jet by the UIS geometry. (author)

  19. An Integrated Multicriteria Decision-Making Approach for Evaluating Nuclear Fuel Cycle Systems for Long-term Sustainability on the Basis of an Equilibrium Model: Technique for Order of Preference by Similarity to Ideal Solution, Preference Ranking Organization Method for Enrichment Evaluation, and Multiattribute Utility Theory Combined with Analytic Hierarchy Process

    Directory of Open Access Journals (Sweden)

    Saerom Yoon

    2017-02-01

    Full Text Available The focus on the issues surrounding spent nuclear fuel and lifetime extension of old nuclear power plants continues to grow nowadays. A transparent decision-making process to identify the best suitable nuclear fuel cycle (NFC is considered to be the key task in the current situation. Through this study, an attempt is made to develop an equilibrium model for the NFC to calculate the material flows based on 1 TWh of electricity production, and to perform integrated multicriteria decision-making method analyses via the analytic hierarchy process technique for order of preference by similarity to ideal solution, preference ranking organization method for enrichment evaluation, and multiattribute utility theory methods. This comparative study is aimed at screening and ranking the three selected NFC options against five aspects: sustainability, environmental friendliness, economics, proliferation resistance, and technical feasibility. The selected fuel cycle options include pressurized water reactor (PWR once-through cycle, PWR mixed oxide cycle, or pyroprocessing sodium-cooled fast reactor cycle. A sensitivity analysis was performed to prove the robustness of the results and explore the influence of criteria on the obtained ranking. As a result of the comparative analysis, the pyroprocessing sodium-cooled fast reactor cycle is determined to be the most competitive option among the NFC scenarios.

  20. An integrated multicriteria decision-making approach for evaluating nuclear fuel cycle systems for long-term sustainability on the basis of an equilibrium model: Technique for order of preference by similarity to ideal solution, preference ranking organization method for enrichment evaluation, and multiattribute utility theory combined with analytic hierarchy process

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Sae Rom [Dept of Quantum Energy Chemical Engineering, Korea University of Science and Technology (KUST), Daejeon (Korea, Republic of); Choi, Sung Yeol [Ulsan National Institute of Science and Technology, Ulju (Korea, Republic of); Ko, Wonil [Nonproliferation System Development Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-02-15

    The focus on the issues surrounding spent nuclear fuel and lifetime extension of old nuclear power plants continues to grow nowadays. A transparent decision-making process to identify the best suitable nuclear fuel cycle (NFC) is considered to be the key task in the current situation. Through this study, an attempt is made to develop an equilibrium model for the NFC to calculate the material flows based on 1 TWh of electricity production, and to perform integrated multicriteria decision-making method analyses via the analytic hierarchy process technique for order of preference by similarity to ideal solution, preference ranking organization method for enrichment evaluation, and multiattribute utility theory methods. This comparative study is aimed at screening and ranking the three selected NFC options against five aspects: sustainability, environmental friendliness, economics, proliferation resistance, and technical feasibility. The selected fuel cycle options include pressurized water reactor (PWR) once-through cycle, PWR mixed oxide cycle, or pyroprocessing sodium-cooled fast reactor cycle. A sensitivity analysis was performed to prove the robustness of the results and explore the influence of criteria on the obtained ranking. As a result of the comparative analysis, the pyroprocessing sodium-cooled fast reactor cycle is determined to be the most competitive option among the NFC scenarios.

  1. Material properties of oxide dispersion strengthened (ODS) ferritic steels for core materials of FBR. Tensile properties of sodium exposed and nickel diffused materials

    International Nuclear Information System (INIS)

    Kato, Shoichi; Yoshida, Eiichi

    2002-12-01

    An oxide dispersion strengthened (ODS) ferritic steel is candidate for a long-life core materials of future FBR, because of good swelling resistance and high creep strength. In this study, tensile tests were carried out the long-term extrapolation of sodium environmental effects on the mechanical properties of ODS steels. The tested heats of materials are M93, M11 and F95. The specimens were pre-exposed to sodium for 1,000 and 3,000 hours under non-stress conditions. The pre-exposure to sodium was conducted using a sodium test loop constituted by austenitic steels. For the conditions of sodium exposure test, the sodium temperature was 650 and 700degC, the oxygen concentration in sodium was about 1 ppm and sodium flow rate on the surface of specimen was less than 1x10 -4 m/seconds (nearly static). Further the specimen with the nickel diffused was prepared, which is simulate to nickel diffusing through sodium from the surface of structural stainless steels. The main results obtained were as follows; (1) The tensile strength and the fracture elongation after sodium exposure (maximum 3,000 hours) were same as that of as-received materials. If was considered that the sodium environmental effect is negligible under the condition of this study. (2) Tensile properties of nickel diffused specimens were slightly lower than that of the as-received specimens, but it remains equal to that of thermal aging specimens. (3) The change in microstructure such as a degraded layer was observed on the surface of nickel diffused specimen. In the region of the degraded layer, phase transformations from the α-phase to the γ-phase were recognized. But, the microscopic oxide particles were observed same as that of α-phase base metal. (author)

  2. Analytics for Education

    Science.gov (United States)

    MacNeill, Sheila; Campbell, Lorna M.; Hawksey, Martin

    2014-01-01

    This article presents an overview of the development and use of analytics in the context of education. Using Buckingham Shum's three levels of analytics, the authors present a critical analysis of current developments in the domain of learning analytics, and contrast the potential value of analytics research and development with real world…

  3. 2nd RCM of the CRP on Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems (ADS) and Technical Meeting on Low Enriched Uranium (LEU) Fuel Utilization in Accelerator Driven Sub-critical Systems. Working Material

    International Nuclear Information System (INIS)

    2010-01-01

    The overall objective of the CRP is contributing to the generic R&D efforts in various fields common to innovative fast neutron system development, i.e., heavy liquid metal thermal hydraulics, dedicated transmutation fuels and associated core designs, theoretical nuclear reaction models, measurement and evaluation of nuclear data for transmutation, and development and validation of calculational methods and codes. Ultimately, the CRP’s overall objective is to make contributions towards the realization of a transmutation demonstration facility

  4. Second Karlsruhe international conference on analytical chemistry in nuclear technology

    International Nuclear Information System (INIS)

    1989-01-01

    Around 180 abstracts of invited lectures and poster presentations of the international analytical conference are presented in this book. They cover analytical applications throughout the fuel cycle and radioanalysis of manifold materials. Most of the abstracts are prepared separately for input in INIS and EDB. (RB)

  5. Modeling and analytical simulation of a smouldering carbonaceous ...

    African Journals Online (AJOL)

    Modeling and analytical simulation of a smouldering carbonaceous rod. A.A. Mohammed, R.O. Olayiwola, M Eseyin, A.A. Wachin. Abstract. Modeling of pyrolysis and combustion in a smouldering fuel bed requires the solution of flow, heat and mass transfer through porous media. This paper presents an analytical method ...

  6. Research on the general analytical method of fossil fuel cycle from a viewpoint of the global environment. 3; Chikyu kankyo kara mita sogoteki kaseki nenryo cycle bunseki hyoka shuho no chosa. 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-01

    The general analysis/assessment method of a fossil fuel cycle was studied. Seven kinds of power generation plants such as LNG cycle and coal cycle ones, and four kinds of transport and treatment systems of recovered CO2 such as ocean and underground systems were studied as case studies on life cycle analysis. As data necessary for life cycle analysis, the database was constructed which stores the facilities and operational energy required for a total energy system from mining of fossil fuel to treatment of recovered CO2, and the quantity of environmental waste such as CO2 emission. As a result, the decrease rate of energy balance defined as ratio of input energy to power plant output was estimated to be 14-43% and 20-60% in LNG cycle and coal cycle, respectively. Even if the recovery rate of CO2 in power plants reached 80-90%, reduction of total CO2 emission was limited to only 20-40% because of CO2 emission during mining, liquefaction and transport of fuel. 168 refs., 48 figs., 102 tabs.

  7. Fuel management

    International Nuclear Information System (INIS)

    Schwarz, E.R.

    1975-01-01

    Description of the operation of power plants and the respective procurement of fuel to fulfil the needs of the grid. The operation of the plants shall be optimised with respect to the fuel cost. (orig./RW) [de

  8. Fuel gases

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    This paper gives a brief presentation of the context, perspectives of production, specificities, and the conditions required for the development of NGV (Natural Gas for Vehicle) and LPG-f (Liquefied Petroleum Gas fuel) alternative fuels. After an historical presentation of 80 years of LPG evolution in vehicle fuels, a first part describes the economical and environmental advantages of gaseous alternative fuels (cleaner combustion, longer engines life, reduced noise pollution, greater natural gas reserves, lower political-economical petroleum dependence..). The second part gives a comparative cost and environmental evaluation between the available alternative fuels: bio-fuels, electric power and fuel gases, taking into account the processes and constraints involved in the production of these fuels. (J.S.)

  9. Fuel pellet

    International Nuclear Information System (INIS)

    Hayashi, K.

    1980-01-01

    Fuel pellet for insertion into a cladding tube in order to form a fuel element or a fuel rod. The fuel pellet has got a belt-like projection around its essentially cylindrical lateral circumferential surface. The upper and lower edges in vertical direction of this belt-like projection are wave-shaped. The projection is made of the same material as the bulk pellet. Both are made in one piece. (orig.) [de

  10. Fossil Fuels.

    Science.gov (United States)

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  11. Fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A new fuel can with a loose bottom and head is described. The fuel bar is attached to the loose bottom and head with two grid poles keeping the distance between bottom and head. A bow-shaped handle is attached to the head so that the fuel bar can be lifted from the can

  12. Fuel and target programs for the transmutation at Phenix and other reactors

    International Nuclear Information System (INIS)

    Gaillard-Groleas, G.

    2002-01-01

    The fuels and targets program for transmutation, performed in the framework of the axis 1 of the December 1991 law about the researches on the management of long-lived radioactive wastes, is in perfect consistency with the transmutation scenario studies carried out in the same framework. These studies put forward the advantage of fast breeder reactors (FBR) in the incineration of minor actinides and long-lived fission products. The program includes exploratory and technological demonstration studies covering the different design options. It aims at enhancing our knowledge of the behaviour of materials under irradiation and at ensuring the mastery of processes. The goals of the different experiments foreseen at Phenix reactor are presented. The main goal is to supply a set of results allowing to precise the conditions of the technical feasibility of minor actinides and long-lived fission products incineration in FBRs. (J.S.)

  13. Accelerator breeder: a viable option for the production of nuclear fuels

    International Nuclear Information System (INIS)

    Grand, P.

    1983-01-01

    Despite the growing pains of the US nuclear power industry, our dependence on nuclear energy for the production of electricity and possibly process heat is likely to increase dramatically over the next few deacades. This statement dismisses fusion as being entirely too speculative to be practical within that time frame. Sometime, between the years 2000 and 2050, fissile material will be in short supply whether it is to fuel existing LWR's or to provide initial fuel inventory for FBR's. The accelerator breeder could produce the fuel shortfall predicted to occur during the first half of the 21st century. The accelerator breeder offers the only practical means today of producing, or breeding, large quantities of fissile fuel from fertile materials, albeit at high cost. Studies performed over the last few years at Chalk River Laboratory and at Brookhaven National Laboratory have demonstrated that the accelerator breeder is practical, technically feasible with state-of-the-art technology, and is economically competitive with any other proposed synthetic means of fissile fuel production. This paper gives the parameters of a nearly optimized accelerator-breeder system, then discusses the development needs, and the economics and institutional problems that this breeding concept faces

  14. Nuclear fuel reliability in NPP KRSKO

    International Nuclear Information System (INIS)

    Antolovic, A.; Kurincic, B.

    2001-01-01

    The importance of achieving and maintaining high fuel integrity comes from negative consequences of operation with failed fuel. Failed fuel has a significant effect on operating cost and performance, and increases the radiological consequences to environment. Fuel failures represent a breach in the first barrier (cladding) preventing the release of fission products. Historically NPP Krsko experienced some degradation of fuel cladding integrity. To resolve this problem and to ensure the safe, reliable and cost effective operation of nuclear fuel, NPP Krsko established 'Fuel Integrity Program'. The key elements of the Program are: continuous monitoring and trending of the fuel behaviour through operating cycle, evaluation of key performance indicators (RCS isotopes, operational parameters) to determine whether the fuel defects exist, implementation of appropriate actions to reduce and mitigate the consequences of fuel defects (four action levels), 100% examination of fuel to remove the defective fuel from operation (Ultrasonic (UT), In Mast Sipping (IMS) and visual inspection), evaluating the worldwide experience and fuel performance and, integrating the experience and knowledge into new fuel design (ZIRLO TM cladding, debris filter bottom nozzle, removable top nozzle). Since start of commercial operation fuel integrity has been evaluated considering certain aspects like operation and fuel handling, fuel rod burnup and cycle length, cladding material properties, etc. As a result of successful Fuel Integrity Program NPP Krsko has achieved high performance level in terms of fuel integrity in past four cycles. Also, NPP Krsko calculations show good matching between analytical prediction of number of failed fuel rods from primary coolant activity analysis and inspection results with the Nondestructive Testing (NDT) methods.(author)

  15. Fuel assemblies

    International Nuclear Information System (INIS)

    Nakamura, Mitsuya; Yamashita, Jun-ichi; Mochida, Takaaki.

    1986-01-01

    Purpose: To improve the fuel economy by increasing the reactivity at the latter burning stage of fuel assemblies and thereby increasing the burn-up degree. Constitution: At the later stage of the burning where the infinite multiplication factor of a fuel assembly is lowered, fuel rods are partially discharged to increase the fuel-moderator volume ratio in the fuel assembly. Then, plutonium is positively burnt by bringing the ratio near to an optimum point where the infinite multiplication factor becomes maximum and the reactivity of the fuel assembly is increased by utilizing the spectral shift effect. The number of the fuel rods to be removed is selected so as to approach the fuel-moderator atom number ratio where the infinite multiplication factor is maximum. Further, the positions where the thermal neutron fluxes are low are most effective for removing the rods and those positions between which no fuel rods are present and which are adjacent with neither the channel box nor the water rods are preferred. The rods should be removed at the time when the burning is proceeded at lest for one cycle. The reactivity is thus increased and the burn-up degree of fuels upon taking-out can be improved. (Kamimura, M.)

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

    International Nuclear Information System (INIS)

    Silvennoinen, P.; Vira, J.; Westerberg, R.

    1982-01-01

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

  17. Remote maintenance system technology development for nuclear fuel cycle plants

    International Nuclear Information System (INIS)

    Kashihara, Hidechiyo

    1984-01-01

    The necessity of establishing the technology of remote maintenance, the kinds of maintenance techniques and the change, the image of a facility adopting remote maintenance canyon process, and the outline of the R and D plan to put remote maintenance canyon process in practical use are described. As the objects of development, there are twin arm type servo manipulator system, rack system, remote tube connectors, solution sampling system, inspection system for in-cell equipment, and large plugs for wall penetration. The outline of those are also reported. The development of new remote maintenance technology has been forwarded in the Tokai Works aiming at the application to a glass solidification pilot plant and a FBR fuel recycling test facility. The lowering of the rate of utilization of cells due to poor accessibility and the increase of radiation exposure of workers must be overcome to realize nuclear fuel cycle technology. The maintenance technology is classified into crane canyon method, direct maintenance cell method, remote maintenance cell method and remote maintenance canyon method, and those are described briefly. The development plan of remote maintenance technology is outlined. (Kako, I.)

  18. Clustering in analytical chemistry.

    Science.gov (United States)

    Drab, Klaudia; Daszykowski, Michal

    2014-01-01

    Data clustering plays an important role in the exploratory analysis of analytical data, and the use of clustering methods has been acknowledged in different fields of science. In this paper, principles of data clustering are presented with a direct focus on clustering of analytical data. The role of the clustering process in the analytical workflow is underlined, and its potential impact on the analytical workflow is emphasized.

  19. Fuel assemblies

    International Nuclear Information System (INIS)

    Sadaoka, Noriyuki.

    1986-01-01

    Purpose: To maintain a satisfactory integrity by preventing the increase of corrosion at the outer surface of a fuel can near the point of contact between the fuel can and the spacer due to the use of fuel pellets incorporated with burnable poisons. Constitution: Since reactor coolants are at high temperature and high pressure, zirconium and water are brought into reaction to proceed oxidation at the outer surface of a fuel can to form uniform oxidation layers. However, abrasion corrosion is additionally formed at the contact portion between the spacer and the fuel can, by which the corrosion is increased by about 25 %. For preventing such nodular corrosion, fuel pellets not incorporated with burnable poisons are charged at a portion of the fuel rod where the spacer is supported and fuel pellets incorporated with burnable poisons are charged at the positions other than about to thereby suppress the amount of the corrosion at the portion where the corrosion of the fuel can is most liable to be increased to thereby improve the fuel integrity. That is, radiolysis of coolants due to gamma-rays produced from gadolinium is lowered to reduce the oxygen concentration near the outer surface thereby preventing the corrosion. (Kawakami, Y.)

  20. Analyticity without Differentiability

    Science.gov (United States)

    Kirillova, Evgenia; Spindler, Karlheinz

    2008-01-01

    In this article we derive all salient properties of analytic functions, including the analytic version of the inverse function theorem, using only the most elementary convergence properties of series. Not even the notion of differentiability is required to do so. Instead, analytical arguments are replaced by combinatorial arguments exhibiting…

  1. Fuel Chemistry Division annual progress report for 1986

    International Nuclear Information System (INIS)

    Marathe, S.G.; Aggarwal, S.K.

    1989-01-01

    The research and development activities of the Fuel Chemistry Division during 1986 are reported in the form of summaries. These activities mainly deal with nuclear fuel development, the chemistry of actinides and solid and solution state, analytical methods for chemical quality control of fuels and other related materials. (M.G.B.)

  2. PROBLEMS OF DETERMINING THE FUEL COST FOR INTERNATIONAL ROAD TRANSPORTATION

    Directory of Open Access Journals (Sweden)

    S. Bondarev

    2016-12-01

    Full Text Available When performing international goods transportation the most expensive consumption is the fuel. For planing reliable fuel costs there was conducted analytical and experimental research. According to the research, the method to determine the volume and cost of fuel according to the criterion of its maximum use with minimum cost within the country follow routes is determined.

  3. Fuel spacer

    International Nuclear Information System (INIS)

    Nishida, Koji; Yokomizo, Osamu; Kanazawa, Toru; Kashiwai, Shin-ichi; Orii, Akihito.

    1992-01-01

    The present invention concerns a fuel spacer for a fuel assembly of a BWR type reactor and a PTR type reactor. Springs each having a vane are disposed on the side surface of a circular cell which supports a fuel rods. A vortex streams having a vertical component are formed by the vanes in the flowing direction of a flowing channel between adjacent cylindrical cells. Liquid droplets carried by streams are deposited on liquid membrane streams flowing along the fuel rod at the downstream of the spacer by the vortex streams. In view of the above, the liquid droplets can be deposited to the fuel rod without increasing the amount of metal of the spacer. Accordingly, the thermal margin of the fuel assembly can be improved without losing neutron economy. (I.N.)

  4. Fuel assembly

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi; Matsuzuka, Ryuji.

    1976-01-01

    Object: To provide a fuel assembly which can decrease pressure loss of coolant to uniform temperature. Structure: A sectional area of a flow passage in the vicinity of an inner peripheral surface of a wrapper tube is limited over the entire length to prevent the temperature of a fuel element in the outermost peripheral portion from being excessively decreased to thereby flatten temperature distribution. To this end, a plurality of pincture-frame-like sheet metals constituting a spacer for supporting a fuel assembly, which has a plurality of fuel elements planted lengthwise and in given spaced relation within the wrapper tube, is disposed in longitudinal grooves and in stacked fashion to form a substantially honeycomb-like space in cross section. The fuel elements are inserted and supported in the space to form a fuel assembly. (Kamimura, M.)

  5. Fuel cycle

    International Nuclear Information System (INIS)

    Bahm, W.

    1989-01-01

    The situation of the nuclear fuel cycle for LWR type reactors in France and in the Federal Republic of Germany was presented in 14 lectures with the aim to compare the state-of-the-art in both countries. In addition to the momentarily changing fuilds of fuel element development and fueling strategies, the situation of reprocessing, made interesting by some recent developmnts, was portrayed and differences in ultimate waste disposal elucidated. (orig.) [de

  6. Nuclear fuel

    International Nuclear Information System (INIS)

    Azevedo, J.B.L. de.

    1980-01-01

    All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.) [pt

  7. Fuel assembly

    International Nuclear Information System (INIS)

    Nomata, Terumitsu.

    1993-01-01

    Among fuel pellets to be loaded to fuel cans of a fuel assembly, fuel pellets having a small thermal power are charged in a region from the end of each of spacers up to about 50mm on the upstream of coolants that flow vertically at the periphery of fuel rods. Coolants at the periphery of fuel rods are heated by the heat generation, to result in voids. However, since cooling effect on the upstream of the spacers is low due to influences of the spacers. Further, since the fuel pellets disposed in the upstream region have small thermal power, a void coefficient is not increased. Even if a thermal power exceeding cooling performance should be generated, there is no worry of causing burnout in the upstream region. Even if burnout should be caused, safety margin and reliability relative to burnout are improved, to increase an allowable thermal power, thereby enabling to improve integrity and reliability of fuel rods and fuel assemblies. (N.H.)

  8. Theory of precipitation effects on dead cylindrical fuels

    Science.gov (United States)

    Michael A. Fosberg

    1972-01-01

    Numerical and analytical solutions of the Fickian diffusion equation were used to determine the effects of precipitation on dead cylindrical forest fuels. The analytical solution provided a physical framework. The numerical solutions were then used to refine the analytical solution through a similarity argument. The theoretical solutions predicted realistic rates of...

  9. The fuel cycle

    International Nuclear Information System (INIS)

    2000-01-01

    In this brochure the fuel cycle is presented. The following fuel cycle steps are described: (1) Front of the fuel cycle (Mining and milling; Treatment; Refining, conversion and enrichment; Fuel fabrication); (2) Use of fuel in nuclear reactors; (3) Back end of the fuel cycle (Interim storage of spent fuel; spent fuel reprocessing; Final disposal of spent fuel)

  10. An overview of analytical activities of control laboratory in NFC

    International Nuclear Information System (INIS)

    Balaji Rao, Y.; Subba Rao, Y.; Saibaba, N.

    2015-01-01

    As per the mandate of Department of Atomic Energy (DAE), Nuclear Fuel Complex (NFC) was established in 1971 for manufacturing Fuel Sub-assemblies for both PHWRs and BWRs operating in India on industrial scale. Control Laboratory (C.Lab) was envisaged as a centralized analytical facility to achieve the objectives of NFC on the similar lines of its predecessor, Analytical Chemistry Division at BARC. With highest ever production of 1200 MT of PHWR Fuel and 16 lakhs PHWR Fuel Tubes achieved during production year of 2014-15 and with increase in demand further for fuel requirements, NFC has got demanding situation in next year and accordingly, C. Lab has also geared up to meet the challenging demands of all the production plant. The average annual analytical load comes around 5 Lakhs estimations and to manage such a massive analytical load a proper synergy between good chemistry, process conditions and analytical methods is a necessity and laboratory is able to meet this important requirement consistently

  11. Fuel Cells

    DEFF Research Database (Denmark)

    Smith, Anders; Pedersen, Allan Schrøder

    2014-01-01

    Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications...

  12. Fuel assembly

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitosi.

    1993-01-01

    Fuel pellets containing burnable poison and fuel pellets not containing burnable poison are used together in burnable poison-incorporated fuel rods which is disposed at the outermost layer of a cluster. Since the burnable poison-incorporated fuel rods are disposed at the outermost layer of the cluster where a neutron flux level is high and, accordingly, the power is high originally, local power peaking can be suppressed and, simultaneously, fuels can be burnt effectively without increasing the fuel concentration in the inner and the intermediate layers than that of the outermost layer. In addition, a problem of lacking a reactor core reactivity at an initial stage is solved by disposing both of the fuel pellets together, even if burnable poisons of high concentration are used. This is because the extent of the lowering of the reactivity due to the burnable poison-incorporated fuels is mainly determined by the surface area thereof and the remaining period of the burnable poison is mainly determined by the concentration thereof. As a result, the burnup degree can be improved without lowering the reactor reactivity so much. (N.H.)

  13. Fuel cells:

    DEFF Research Database (Denmark)

    Sørensen, Bent

    2013-01-01

    A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil...... and nuclear fuel-based energy technologies....

  14. Fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Akiyoshi; Bessho, Yasunori; Aoyama, Motoo; Koyama, Jun-ichi; Hirakawa, Hiromasa; Yamashita, Jun-ichi; Hayashi, Tatsuo

    1998-01-01

    In a fuel assembly of a BWR type reactor in which a water rod of a large diameter is disposed at the central portion, the cross sectional area perpendicular to the axial direction comprises a region a of a fuel rod group facing to a wide gap water region to which a control rod is inserted, a region b of a fuel rod group disposed on the side of the wide gap water region other than the region a, a region d of a fuel rod group facing to a narrow gap water region and a region c of a fuel rod group disposed on the side of the narrow gap water region other than the region d. When comparing an amount of fission products contained in the four regions relative to that in the entire regions and average enrichment degrees of fuel rods for the four regions, the relative amount and the average enrichment degree of the fuel rod group of the region a is minimized, and the relative amount and the average enrichment degree of the fuel rod group in the region b is maximized. Then, reactor shut down margin during cold operation can be improved while flattening the power in the cross section perpendicular to the axial direction. (N.H.)

  15. Nuclear fuel

    International Nuclear Information System (INIS)

    Quinauk, J.P.

    1990-01-01

    Since 1985, Fragema has been marketing and selling the Advanced Fuel Assemby AFA whose main features are its zircaloy grids and removable top and bottom nozzles. It is this product, which exists for several different fuel assembly arrays and heights, that will be employed in the reactors at Daya Bay. Fragema employs gadolinium as the consumable poison to enable highperformance fuel management. More recently, the company has supplied fuel assemblies of the mixed-oxide(MOX) and enriched reprocessed uranium type. The reliability level of the fuel sold by Fragema is one of the highest in the world, thanks in particular to the excellence of the quality assurance and quality control programs that have been implemented at all stages of its design and manufacture

  16. Fuel assemblies

    International Nuclear Information System (INIS)

    Echigoya, Hironori; Nomata, Terumitsu.

    1983-01-01

    Purpose: To render the axial distribution relatively flat. Constitution: First nuclear element comprises a fuel can made of zircalloy i.e., the metal with less neutron absorption, which is filled with a plurality of UO 2 pellets and sealed by using a lower end plug, a plenum spring and an upper end plug by means of welding. Second fuel element is formed by substituting a part of the UO 2 pellets with a water tube which is sealed with water and has a space for allowing the heat expansion. The nuclear fuel assembly is constituted by using the first and second fuel elements together. In such a structure, since water reflects neutrons and decrease their leakage to increase the temperature, reactivity is added at the upper portion of the fuel assembly to thereby flatten the axial power distribution. Accordingly, stable operation is possible only by means of deep control rods while requiring no shallow control rods. (Sekiya, K.)

  17. Technology readiness of partitioning and transmutation toward closed fuel cycle in Japan

    International Nuclear Information System (INIS)

    Ikeda, Kazumi; Kurata, Masaki; Morita, Yasuji; Tsujimoto, Kazufumi; Minato, Kazuo; Koyama, Shin-ichi

    2011-01-01

    This paper treats technology readiness level (TRL) assessment of Partitioning and Transmutation (P-T) toward closed fuel cycle in JAPAN. The purpose is providing clarified information related to the current maturity of the partitioning and transmutation technologies by applying the methodology of TRL, parallel to attempting to establish common indications among relating technology area. The methodology should be one of useful communication tools between specialists and management level, and also among countries interested in the P-T technologies. The generic TRL in this study is based on the GNEP (Global Nuclear Energy Partnership)'s definition: TRL 3 shows the status that critical function is proved and elemental technologies are identified, TRL 4 represents that relating technologies are validated at bench scale in laboratory environment, and TRL 5 achieves the completion of development related to the subsystem and elemental technologies. Detailed indications are established through discussion of the relating specialists. Reviewed technological area includes P-T and minor actinide (MA) cycle: Fast Breeder Reactor (FBR) and Accelerator driven system (ADS) for MA transmutation, partitioning processes, and MA-bearing fuels. The assessments reveal that TRL spreads around TRL 3 to TRL 4 because each system requires more the development of elemental technologies. Transmutation core of FBR is assessed to be TRL 4 in that MA bearing integral test is required additionally, and ADS becomes TRL 3 because the elemental technologies were identified and the requirements were specified. Consequently, the common key issue is how the nuclear calculation methodology will be validated for MA-bearing-fuelled core, since several percentages of MA changes the void reactivity and the Doppler Effect significantly, which are inherently important in reactor safety. It should be that critical experiments with several kg of americium or more are difficult in the existing experimental

  18. News for analytical chemists

    DEFF Research Database (Denmark)

    Andersen, Jens Enevold Thaulov; Karlberg, Bo

    2009-01-01

    The EuCheMS Division of Analytical Chemistry (DAC) maintains a website with informations on groups of analytical chemistry at European universities (www.dac-euchems. org). Everyone may contribute to the database and contributors are responsible for an annual update of the information. The service...... is offered free of charge. The report on activities of DAC during 2008 was published in journals of analytical chemistry where Manfred Grasserbauer contributed with his personal view on analytical chemistry in the assessment of climate changes and sustainable application of the natural resources to human...... directed to various topics of analytical chemistry. Although affected by the global financial crisis, the Euroanalysis Conference will be held on 6 to 10 September in Innsbruck, Austria. For next year, the programme for the analytical section of the 3rd European Chemistry Congress is in preparation...

  19. High-freezing-point fuel studies

    Science.gov (United States)

    Tolle, F. F.

    1980-01-01

    Considerable progress in developing the experimental and analytical techniques needed to design airplanes to accommodate fuels with less stringent low temperature specifications is reported. A computer technique for calculating fuel temperature profiles in full tanks was developed. The computer program is being extended to include the case of partially empty tanks. Ultimately, the completed package is to be incorporated into an aircraft fuel tank thermal analyser code to permit the designer to fly various thermal exposure patterns, study fuel temperatures versus time, and determine holdup.

  20. Analytical Chemistry in Russia.

    Science.gov (United States)

    Zolotov, Yuri

    2016-09-06

    Research in Russian analytical chemistry (AC) is carried out on a significant scale, and the analytical service solves practical tasks of geological survey, environmental protection, medicine, industry, agriculture, etc. The education system trains highly skilled professionals in AC. The development and especially manufacturing of analytical instruments should be improved; in spite of this, there are several good domestic instruments and other satisfy some requirements. Russian AC has rather good historical roots.

  1. Science Update: Analytical Chemistry.

    Science.gov (United States)

    Worthy, Ward

    1980-01-01

    Briefly discusses new instrumentation in the field of analytical chemistry. Advances in liquid chromatography, photoacoustic spectroscopy, the use of lasers, and mass spectrometry are also discussed. (CS)

  2. Analytical Electron Microscope

    Data.gov (United States)

    Federal Laboratory Consortium — The Titan 80-300 is a transmission electron microscope (TEM) equipped with spectroscopic detectors to allow chemical, elemental, and other analytical measurements to...

  3. Fuel cell-fuel cell hybrid system

    Science.gov (United States)

    Geisbrecht, Rodney A.; Williams, Mark C.

    2003-09-23

    A device for converting chemical energy to electricity is provided, the device comprising a high temperature fuel cell with the ability for partially oxidizing and completely reforming fuel, and a low temperature fuel cell juxtaposed to said high temperature fuel cell so as to utilize remaining reformed fuel from the high temperature fuel cell. Also provided is a method for producing electricity comprising directing fuel to a first fuel cell, completely oxidizing a first portion of the fuel and partially oxidizing a second portion of the fuel, directing the second fuel portion to a second fuel cell, allowing the first fuel cell to utilize the first portion of the fuel to produce electricity; and allowing the second fuel cell to utilize the second portion of the fuel to produce electricity.

  4. Computer controlled quality of analytical measurements

    International Nuclear Information System (INIS)

    Clark, J.P.; Huff, G.A.

    1979-01-01

    A PDP 11/35 computer system is used in evaluating analytical chemistry measurements quality control data at the Barnwell Nuclear Fuel Plant. This computerized measurement quality control system has several features which are not available in manual systems, such as real-time measurement control, computer calculated bias corrections and standard deviation estimates, surveillance applications, evaluaton of measurement system variables, records storage, immediate analyst recertificaton, and the elimination of routine analysis of known bench standards. The effectiveness of the Barnwell computer system has been demonstrated in gathering and assimilating the measurements of over 1100 quality control samples obtained during a recent plant demonstration run. These data were used to determine equaitons for predicting measurement reliability estimates (bias and precision); to evaluate the measurement system; and to provide direction for modification of chemistry methods. The analytical chemistry measurement quality control activities represented 10% of the total analytical chemistry effort

  5. Caracterización mecánica de recubrimientos de aluminio por CVD-FBR sobre aceros inoxidables y resistencia a la oxidación en vapor de agua

    OpenAIRE

    Diego Pérez-Muñoz; José Luddey Marulanda-Arévalo; Juan Manuel Meza-Meza

    2015-01-01

    Los recubrimientos de aluminio depositados sobre el acero inoxidable austenítico AISI 317 por Deposición Química de Vapor en Lecho Fluidizado (CVD-FBR) presentan a altas temperaturas una reducción de la velocidad de corrosión de más de 80 veces. Se realizó la caracterización mecánica de los recubrimientos por medio de microdureza, nanoindentación, para conocer cómo se vieron afectas las propiedades mecánicas (en especial la dureza y el módulo de Young) del recubrimiento y del sustrato luego d...

  6. Development of an analytical framework to assess the role of new technologies for liquid and gaseous fuels. Volume IV. Decision analysis of national research and development funding for unconventional oil technologies. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Nesbitt, D.M.; North, D.W.; Phillips, R.L.

    1980-12-01

    This analysis explores the issue of whether developing new technology to exploit the light oil, heavy oil, and tar sands resources is in the national interest. For purposes of this analysis, we will term light oil, heavy oil, and tar sands recovery technologies collectively as enhanced oil recovery technologies (EOR). When we use the term EOR, it will be understood that we are referring to all three of the aforementioned resources and the technologies necessary to exploit them. EOR does not consist of a single technology but rather of a number of related technologies for extracting oil unreachable by primary and secondary techniques or bitumen from tar sands. EOR technologies considered to be currently proven are relatively primitive and apply only to a limited number of reservoirs, principally those containing heavy oil. Although more advanced EOR techniques have removed 90% of the oil from core samples in the laboratory, they have not yet been demonstrated in field tests. Substantial scientific and engineering progress will be needed to reproduce laboratory results under field conditions. Finally, the diversity of known reservoirs indicates that, in all likelihood, no single EOR technology will be usable on more than a small fraction of known reservoirs. The appropriate level of national expenditure on EOR R and D is the focus of this analysis. This report describes an analytical framework and a set of assumptions designed to gain insight regarding the appropriate level of national (public plus private) expenditures on EOR R and D. The analysis described herein has been designed to address this question only; it does not address the problem of the optimal government/industry split or the best EOR technology mix.

  7. Analytical applications for delayed neutrons

    International Nuclear Information System (INIS)

    Eccleston, G.W.

    1983-01-01

    Analytical formulations that describe the time dependence of neutron populations in nuclear materials contain delayed-neutron dependent terms. These terms are important because the delayed neutrons, even though their yields in fission are small, permit control of the fission chain reaction process. Analytical applications that use delayed neutrons range from simple problems that can be solved with the point reactor kinetics equations to complex problems that can only be solved with large codes that couple fluid calculations with the neutron dynamics. Reactor safety codes, such as SIMMER, model transients of the entire reactor core using coupled space-time neutronics and comprehensive thermal-fluid dynamics. Nondestructive delayed-neutron assay instruments are designed and modeled using a three-dimensional continuous-energy Monte Carlo code. Calculations on high-burnup spent fuels and other materials that contain a mix of uranium and plutonium isotopes require accurate and complete information on the delayed-neutron periods, yields, and energy spectra. A continuing need exists for delayed-neutron parameters for all the fissioning isotopes

  8. FUEL ELEMENT

    Science.gov (United States)

    Bean, R.W.

    1963-11-19

    A ceramic fuel element for a nuclear reactor that has improved structural stability as well as improved cooling and fission product retention characteristics is presented. The fuel element includes a plurality of stacked hollow ceramic moderator blocks arranged along a tubular raetallic shroud that encloses a series of axially apertured moderator cylinders spaced inwardly of the shroud. A plurality of ceramic nuclear fuel rods are arranged in the annular space between the shroud and cylinders of moderator and appropriate support means and means for directing gas coolant through the annular space are also provided. (AEC)

  9. Advanced biological treatment of aqueous effluent from the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Pitt, W.W. Jr.; Hancher, C.W.; Patton, B.D.; Shumate, S.E. II.

    1980-01-01

    Many of the processing steps in the nuclear fuel cycle generate aqueous effluent streams bearing contaminants that can, because of their chemical or radiological properties, pose an environmental hazard. Concentration of such contaminants must be reduced to acceptable levels before the streams can be discharged to the environment. Two classes of contaminants, nitrates and heavy metals, are addressed in this study. Specific techniques aimed at the removal of nitrates and radioactive heavy metals by biological processes are being developed, tested, and demonstrated. Although cost comparisons between biological processes and current treatment methods will be presented, these comparisons may be misleading because biological processes yield environmentally better end results which are difficult to price. The fluidized-bed biological denitrification process is an environmentally acceptable and economically sound method for the disposal of nonreusable sources of nitrate effluents. A very high denitrification rate can be obtained in a FBR as the result of a high concentration of denitrification bacteria in the bioreactor and the stagewise operation resulting from plug flow in the reactor. The overall denitrification rate in an FBR ranges from 20- to 100-fold greater than that observed for an STR bioreactor. It has been shown that the system can be operated using Ca 2+ , Na + , or NH 4 + cations at nitrate concentrations up to 1 g/liter without inhibition. Biological sorption of uranium and other radionuclides (particularly the actinides) from dilute aqueous waste streams shows considerable promise as a means of recovering these valuable resources and reducing the environmental impact, however, further development efforts are required

  10. Elements of nuclear reactor fueling theory

    International Nuclear Information System (INIS)

    Egan, M.R.

    1984-01-01

    Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behavior and physical requirements of operating cycle sequences and fueling strategies having practical use in the management of nuclear fuel. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and maneuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy, and which govern fueling decisions normally made by the fuel manager. It is also demonstrated in this application that the simple batch size effect is not valid for non-integer fueling strategies, even in the simplest sequence configurations, and that it systematically underestimates the fueling requirements of degenerate sequences in general

  11. Learning Analytics Considered Harmful

    Science.gov (United States)

    Dringus, Laurie P.

    2012-01-01

    This essay is written to present a prospective stance on how learning analytics, as a core evaluative approach, must help instructors uncover the important trends and evidence of quality learner data in the online course. A critique is presented of strategic and tactical issues of learning analytics. The approach to the critique is taken through…

  12. Smart city analytics

    DEFF Research Database (Denmark)

    Hansen, Casper; Hansen, Christian; Alstrup, Stephen

    2017-01-01

    is very useful when full records are not accessible or available. Smart city analytics does not necessarily require full city records. To our knowledge this preliminary study is the first to predict large increases in home care for smart city analytics....

  13. The Analytical Hierarchy Process

    DEFF Research Database (Denmark)

    Barfod, Michael Bruhn

    2007-01-01

    The technical note gathers the theory behind the Analytical Hierarchy Process (AHP) and present its advantages and disadvantages in practical use.......The technical note gathers the theory behind the Analytical Hierarchy Process (AHP) and present its advantages and disadvantages in practical use....

  14. Quine's "Strictly Vegetarian" Analyticity

    NARCIS (Netherlands)

    Decock, L.B.

    2017-01-01

    I analyze Quine’s later writings on analyticity from a linguistic point of view. In Word and Object Quine made room for a “strictly vegetarian” notion of analyticity. In later years, he developed this notion into two more precise notions, which I have coined “stimulus analyticity” and “behaviorist

  15. Of the Analytical Engine

    Indian Academy of Sciences (India)

    with me, at breakfast, the various powers of the Analytical Engine. After a long conversa- tion on the subject, he inquired what the machine could do if, .... The following conditions relate to the algebraic portion of the Analytical Engine: (e) The number of literal constants must be unlimited. (f) The number of variables must be ...

  16. European Analytical Column

    DEFF Research Database (Denmark)

    Karlberg, B.; Grasserbauer, M.; Andersen, Jens Enevold Thaulov

    2009-01-01

    The European Analytical Column has once more invited a guest columnist to give his views on various matters related to analytical chemistry in Europe. This year, we have invited Professor Manfred Grasserbauer of the Vienna University of Technology to present some of the current challenges for Eur...

  17. Analytic Moufang-transformations

    International Nuclear Information System (INIS)

    Paal, Eh.N.

    1988-01-01

    The paper is aimed to be an introduction to the concept of an analytic birepresentation of an analytic Moufang loop. To describe the deviation of (S,T) from associativity, the associators (S,T) are defined and certain constraints for them, called the minimality conditions of (S,T) are established

  18. Microfluidic fuel cells and batteries

    CERN Document Server

    Kjeang, Erik

    2014-01-01

    Microfluidic fuel cells and batteries represent a special type of electrochemical power generators that can be miniaturized and integrated in a microfluidic chip. Summarizing the initial ten years of research and development in this emerging field, this SpringerBrief is the first book dedicated to microfluidic fuel cell and battery technology for electrochemical energy conversion and storage. Written at a critical juncture, where strategically applied research is urgently required to seize impending technology opportunities for commercial, analytical, and educational utility, the intention is

  19. Quo vadis, analytical chemistry?

    Science.gov (United States)

    Valcárcel, Miguel

    2016-01-01

    This paper presents an open, personal, fresh approach to the future of Analytical Chemistry in the context of the deep changes Science and Technology are anticipated to experience. Its main aim is to challenge young analytical chemists because the future of our scientific discipline is in their hands. A description of not completely accurate overall conceptions of our discipline, both past and present, to be avoided is followed by a flexible, integral definition of Analytical Chemistry and its cornerstones (viz., aims and objectives, quality trade-offs, the third basic analytical reference, the information hierarchy, social responsibility, independent research, transfer of knowledge and technology, interfaces to other scientific-technical disciplines, and well-oriented education). Obsolete paradigms, and more accurate general and specific that can be expected to provide the framework for our discipline in the coming years are described. Finally, the three possible responses of analytical chemists to the proposed changes in our discipline are discussed.

  20. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Sei; Ando, Ryohei; Mitsutake, Toru.

    1995-01-01

    The present invention concerns a fuel assembly suitable to a BWR-type reactor and improved especially with the nuclear characteristic, heat performance, hydraulic performance, dismantling or assembling performance and economical property. A part of poison rods are formed as a large-diameter/multi-region poison rods having a larger diameter than a fuel rod. A large number of fuel rods are disposed surrounding a large diameter water rod and a group of the large-diameter/multi-region poison rods in adjacent with the water rod. The large-diameter water rod has a burnable poison at the tube wall portion. At least a portion of the large-diameter poison rods has a coolant circulation portion allowing coolants to circulate therethrough. Since the large-diameter poison rods are disposed at a position of high neutron fluxes, a large neutron multiplication factor suppression effect can be provided, thereby enabling to reduce the number of burnable poison rods relative to fuels. As a result, power peaking in the fuel assembly is moderated and a greater amount of plutonium can be loaded. In addition the flow of cooling water which tends to gather around the large diameter water rod can be controlled to improve cooling performance of fuels. (N.H.)

  1. TMI criticality studies: lower vessel rubble and analytical benchmarking

    International Nuclear Information System (INIS)

    Westfall, R.M.; Knight, J.R.; Fox, P.B.; Hermann, O.W.; Turner, J.C.

    1986-05-01

    A bounding strategy has been adopted for assuring subcriticality during all TMI-2 defueling operations. The strategy is based upon establishing a safe soluble boron level for the entire reactor core in an optimum reactivity configuration. This paper presents the determination of a fuel rubble model which yields a maximum infinite lattice multiplication factor and the subsequent application of cell-averaged constants in finite system analyses. Included in the analyses are the effects of fuel burnup determined from a simplified power history of the reactor. A discussion of the analytical methods employed and the determination of an analytical bias with benchmark critical experiments completes the presentation. 14 refs., 17 tabs

  2. Elastic plastic analysis of fuel element assemblies - hexagonal claddings and fuel rods

    International Nuclear Information System (INIS)

    Mamoun, M.M.; Wu, T.S.; Chopra, P.S.; Rardin, D.C.

    1979-01-01

    Analytical studies have been conducted to investigate the structural, thermal, and mechanical behavior of fuel rods, claddings and fuel element assemblies of several designs for a conceptual Safety Test Facility (STF). One of the design objectives was to seek a geometrical configuration for a clad by maximizing the volume fraction of fuel and minimizing the resultant stresses set-up in the clad. The results of studies conducted on various geometrical configurations showed that the latter design objective can be achieved by selecting a clad of an hexagonal geometry. The analytical studies necessitated developing solutions for determining the stresses, strains, and displacements experienced by fuel rods and an hexagonal cladding subjected to thermal fuel-bowing loads acting on its internal surface, the external pressure of the coolant, and elevated temperatures. This paper presents some of the initially formulated analytical methods and results. It should be emphasized that the geometrical configuration considered in this paper may not necessarily be similar to that of the final design. Several variables have been taken into consideration including cladding thickness, the dimensions of the fuel rod, the temperature of the fuel and cladding, the external pressure of the cooling fluid, and the mechanical strength properties of fuel and cladding. A finite-element computer program, STRAW Code, has also been employed to generate several numerical results which have been compared with those predicted by employing the initially formulated solutions. The theoretically predicted results are in good agreement with those of the STRAW Code. (orig.)

  3. Google analytics integrations

    CERN Document Server

    Waisberg, Daniel

    2015-01-01

    A roadmap for turning Google Analytics into a centralized marketing analysis platform With Google Analytics Integrations, expert author Daniel Waisberg shows you how to gain a more meaningful, complete view of customers that can drive growth opportunities. This in-depth guide shows not only how to use Google Analytics, but also how to turn this powerful data collection and analysis tool into a central marketing analysis platform for your company. Taking a hands-on approach, this resource explores the integration and analysis of a host of common data sources, including Google AdWords, AdSens

  4. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

    Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Abe, Hideaki; Sakai, Takao; Ishida, Tomio; Yokota, Norikatsu.

    1992-01-01

    The lower ends of a plurality of plate-like shape memory alloys are secured at the periphery of the upper inside of the handling head of a fuel assembly. As the shape memory alloy, a Cu-Zn alloy, a Ti-Pd alloy or a Fe-Ni alloy is used. When high temperature coolants flow out to the handling head, the shape memory alloy deforms by warping to the outer side more greatly toward the upper portion thereof with the temperature increase of the coolants. As the result, the shape of the flow channel of the coolants is changed so as to enlarge at the exit of the upper end of the fuel assembly. Then, the pressure loss of the coolants in the fuel assembly is decreased by the enlargement. Accordingly, the flow rate of the coolants in the fuel assembly is increased to lower the temperature of the coolants. Further, high temperature coolants and low temperature coolants are mixed sufficiently just above the fuel assembly. This can suppress the temperature fluctuation of the mixed coolants in the upper portion of the reactor core, thereby enabling to decrease a fatigue and failures of the structural components in the upper portion of the reactor core. (I.N.)

  6. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    Page, R.D.

    1976-03-01

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  7. Chemometrics in analytical spectroscopy

    National Research Council Canada - National Science Library

    Adams, Mike J

    1995-01-01

    This book provides students and practising analysts with a tutorial guide to the use and application of the more commonly encountered techniques used in processing and interpreting analytical spectroscopic data...

  8. Analytical strategies for phosphoproteomics

    DEFF Research Database (Denmark)

    Thingholm, Tine E; Jensen, Ole N; Larsen, Martin R

    2009-01-01

    sensitive and specific strategies. Today, most phosphoproteomic studies are conducted by mass spectrometric strategies in combination with phospho-specific enrichment methods. This review presents an overview of different analytical strategies for the characterization of phosphoproteins. Emphasis...

  9. Enzymes in Analytical Chemistry.

    Science.gov (United States)

    Fishman, Myer M.

    1980-01-01

    Presents tabular information concerning recent research in the field of enzymes in analytic chemistry, with methods, substrate or reaction catalyzed, assay, comments and references listed. The table refers to 128 references. Also listed are 13 general citations. (CS)

  10. On complex functions analyticity

    CERN Document Server

    Karavashkin, S B

    2002-01-01

    We analyse here the conventional definitions of analyticity and differentiability of functions of complex variable. We reveal the possibility to extend the conditions of analyticity and differentiability to the functions implementing the non-conformal mapping. On this basis we formulate more general definitions of analyticity and differentiability covering those conventional. We present some examples of such functions. By the example of a horizontal belt on a plane Z mapped non-conformally onto a crater-like harmonic vortex, we study the pattern of trajectory variation of a body motion in such field in case of field power function varying in time. We present the technique to solve the problems of such type with the help of dynamical functions of complex variable implementing the analytical non-conformal mapping

  11. Mobility Data Analytics Center.

    Science.gov (United States)

    2016-01-01

    Mobility Data Analytics Center aims at building a centralized data engine to efficiently manipulate : large-scale data for smart decision making. Integrating and learning the massive data are the key to : the data engine. The ultimate goal of underst...

  12. Process Analytical Chemistry

    OpenAIRE

    Trevisan, Marcello G.; Poppi, Ronei J.

    2006-01-01

    Process Analytical Chemistry (PAC) is an important and growing area in analytical chemistry, that has received little attention in academic centers devoted to the gathering of knowledge and to optimization of chemical processes. PAC is an area devoted to optimization and knowledge acquisition of chemical processes, to reducing costs and wastes and to making an important contribution to sustainable development. The main aim of this review is to present to the Brazilian community the developmen...

  13. Analytical Calculations for CAMEA

    OpenAIRE

    Markó, Márton

    2014-01-01

    CAMEA is a novel instrument concept, thus the performance has not been explored. Furthermore it is a complex instrument using many analyser arrays in a wide angular range. The performance of the instrument has been studied by use of three approaches: McStas simulations, analytical calculations, and prototyping. Due to the complexity of the instrument all of the previously mentioned methods can have faults misleading us during the instrument development. We use Monte Carlo and analytical model...

  14. Encyclopedia of analytical surfaces

    CERN Document Server

    Krivoshapko, S N

    2015-01-01

    This encyclopedia presents an all-embracing collection of analytical surface classes. It provides concise definitions  and description for more than 500 surfaces and categorizes them in 38 classes of analytical surfaces. All classes are cross references to the original literature in an excellent bibliography. The encyclopedia is of particular interest to structural and civil engineers and serves as valuable reference for mathematicians.

  15. Intermediate algebra & analytic geometry

    CERN Document Server

    Gondin, William R

    1967-01-01

    Intermediate Algebra & Analytic Geometry Made Simple focuses on the principles, processes, calculations, and methodologies involved in intermediate algebra and analytic geometry. The publication first offers information on linear equations in two unknowns and variables, functions, and graphs. Discussions focus on graphic interpretations, explicit and implicit functions, first quadrant graphs, variables and functions, determinate and indeterminate systems, independent and dependent equations, and defective and redundant systems. The text then examines quadratic equations in one variable, system

  16. ENVIRONMENTAL ANALYTICAL CHEMISTRY OF ...

    Science.gov (United States)

    Within the scope of a number of emerging contaminant issues in environmental analysis, one area that has received a great deal of public interest has been the assessment of the role of pharmaceuticals and personal care products (PPCPs) as stressors and agents of change in ecosystems as well as their role in unplanned human exposure. The relationship between personal actions and the occurrence of PPCPs in the environment is clear-cut and comprehensible to the public. In this overview, we attempt to examine the separations aspect of the analytical approach to the vast array of potential analytes among this class of compounds. We also highlight the relationship between these compounds and endocrine disrupting compounds (EDCs) and between PPCPs and EDCs and the more traditional environmental analytes such as the persistent organic pollutants (POPs). Although the spectrum of chemical behavior extends from hydrophobic to hydrophilic, the current focus has shifted to moderately and highly polar analytes. Thus, emphasis on HPLC and LC/MS has grown and MS/MS has become a detection technique of choice with either electrospray ionization or atmospheric pressure chemical ionization. This contrasts markedly with the bench mark approach of capillary GC, GC/MS and electron ionization in traditional environmental analysis. The expansion of the analyte list has fostered new vigor in the development of environmental analytical chemistry, modernized the range of tools appli

  17. Reforming of fuel inside fuel cell generator

    Science.gov (United States)

    Grimble, Ralph E.

    1988-01-01

    Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream I and spent fuel stream II. Spent fuel stream I is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream I and exhaust stream II, and exhaust stream I is vented. Exhaust stream II is mixed with spent fuel stream II to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells.

  18. Thermal properties of heterogeneous fuels

    International Nuclear Information System (INIS)

    Staicu, D.; Beauvy, M.

    1998-01-01

    Fresh or irradiated nuclear fuels are composites or solid solutions more or less heterogeneous, and their thermal conductivities are strongly dependent on the microstructure. The effective thermal conductivities of these heterogeneous solids must be determined for the modelling of the behaviour under irradiation. Different methods (analytical or numerical) published in the literature can be used for the calculation of this effective thermal conductivity. They are analysed and discussed, but finally only few of them are really useful because the assumptions selected are often not compatible with the complex microstructures observed in the fuels. Numerical calculations of the effective thermal conductivity of various fuels based on the microstructure information provided in our laboratory by optical microscopy or electron micro-probe analysis images, have been done for the validation of these methods. The conditions necessary for accurate results on effective thermal conductivity through these numerical calculations are discussed. (author)

  19. CANDU fuel

    International Nuclear Information System (INIS)

    MacEwan, J.R.; Notley, M.J.F.; Wood, J.C.; Gacesa, M.

    1982-09-01

    The direction of CANDU fuel development was set in 1957 with the decision to build pressure tube reactors. Short - 50 cm long - rodded bundles of natural UO 2 clad in Zircaloy were adopted to facilitate on-power fuelling to improve uranium utilization. Progressive improvements were made during 25 years of development, involving 650 man years and 180 million dollars. Today's CANDU bundle is based on the knowledge gained from extensive irradiation testing and experience in power reactors. The main thrust of future development is to demonstrate that the present bundle is suitable, with minor modifications, for thorium fuels

  20. Hanford analytical sample projections FY 1996 - FY 2001. Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    Joyce, S.M.

    1997-07-02

    This document summarizes the biannual Hanford sample projections for fiscal year 1997-2001. Sample projections are based on inputs submitted to Analytical Services covering Environmental Restoration, Tank Wastes Remediation Systems, Solid Wastes, Liquid Effluents, Spent Nuclear Fuels, Transition Projects, Site Monitoring, Industrial Hygiene, Analytical Services and miscellaneous Hanford support activities. In addition to this revision, details on Laboratory scale technology (development), Sample management, and Data management activities were requested. This information will be used by the Hanford Analytical Services program and the Sample Management Working Group to assure that laboratories and resources are available and effectively utilized to meet these documented needs.

  1. Analytic manifolds in uniform algebras

    International Nuclear Information System (INIS)

    Tonev, T.V.

    1988-12-01

    Here we extend Bear-Hile's result concerning the version of famous Bishop's theorem for one-dimensional analytic structures in two directions: for n-dimensional complex analytic manifolds, n>1, and for generalized analytic manifolds. 14 refs

  2. Croatian Analytical Terminology

    Directory of Open Access Journals (Sweden)

    Kastelan-Macan; M.

    2008-04-01

    Full Text Available Results of analytical research are necessary in all human activities. They are inevitable in making decisions in the environmental chemistry, agriculture, forestry, veterinary medicine, pharmaceutical industry, and biochemistry. Without analytical measurements the quality of materials and products cannot be assessed, so that analytical chemistry is an essential part of technical sciences and disciplines.The language of Croatian science, and analytical chemistry within it, was one of the goals of our predecessors. Due to the political situation, they did not succeed entirely, but for the scientists in independent Croatia this is a duty, because language is one of the most important features of the Croatian identity. The awareness of the need to introduce Croatian terminology was systematically developed in the second half of the 19th century, along with the founding of scientific societies and the wish of scientists to write their scientific works in Croatian, so that the results of their research may be applied in economy. Many authors of textbooks from the 19th and the first half of the 20th century contributed to Croatian analytical terminology (F. Rački, B. Šulek, P. Žulić, G. Pexidr, J. Domac, G. Janeček , F. Bubanović, V. Njegovan and others. M. DeŢelić published the first systematic chemical terminology in 1940, adjusted to the IUPAC recommendations. In the second half of 20th century textbooks in classic analytical chemistry were written by V. Marjanović-Krajovan, M. Gyiketta-Ogrizek, S. Žilić and others. I. Filipović wrote the General and Inorganic Chemistry textbook and the Laboratory Handbook (in collaboration with P. Sabioncello and contributed greatly to establishing the terminology in instrumental analytical methods.The source of Croatian nomenclature in modern analytical chemistry today are translated textbooks by Skoog, West and Holler, as well as by Günnzler i Gremlich, and original textbooks by S. Turina, Z.

  3. ARIANNE. Analytical uncertainties. Simulation of influential factors in the inventory of the final web cam

    International Nuclear Information System (INIS)

    Morales Prieto, M.; Ortega Saiz, P.

    2011-01-01

    Analysis of analytical uncertainties of the methodology of simulation of processes for obtaining isotopic ending inventory of spent fuel, the ARIANE experiment explores the part of simulation of burning.

  4. Fuel Cells

    Science.gov (United States)

    Hawkins, M. D.

    1973-01-01

    Discusses the theories, construction, operation, types, and advantages of fuel cells developed by the American space programs. Indicates that the cell is an ideal small-scale power source characterized by its compactness, high efficiency, reliability, and freedom from polluting fumes. (CC)

  5. Transport fuel

    DEFF Research Database (Denmark)

    Ronsse, Frederik; Jørgensen, Henning; Schüßler, Ingmar

    2014-01-01

    Worldwide, the use of transport fuel derived from biomass increased four-fold between 2003 and 2012. Mainly based on food resources, these conventional biofuels did not achieve the expected emission savings and contributed to higher prices for food commod - ities, especially maize and oilseeds...

  6. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

    Purpose: To reduce the size of the reactor core upper mechanisms and the reactor container, as well as decrease the nuclear power plant construction costs in reactors using liquid metals as the coolants. Constitution: Isotope capturing devices comprising a plurality of pipes are disposed to the gas plenum portion of a nuclear fuel rod main body at the most downstream end in the flowing direction of the coolants. Each of the capturing devices is made of nickel, nickel alloys, stainless steel applied with nickel plating on the surface, nickel alloys applied with nickel plating on the surface or the like. Thus, radioactive nuclides incorporated in the coolants are surely captured by the capturing devices disposed at the most downstream end of the nuclear fuel main body as the coolants flow along the nuclear fuel main body. Accordingly, since discharging of radioactive nuclides to the intermediate fuel exchange system can be prevented, the maintenance or reparing work for the system can be facilitated. (Moriyama, K.)

  7. Modeling the Thermal Rocket Fuel Preparation Processes in the Launch Complex Fueling System

    Directory of Open Access Journals (Sweden)

    A. V. Zolin

    2015-01-01

    Full Text Available It is necessary to carry out fuel temperature preparation for space launch vehicles using hydrocarbon propellant components. A required temperature is reached with cooling or heating hydrocarbon fuel in ground facilities fuel storages. Fuel temperature preparing processes are among the most energy-intensive and lengthy processes that require the optimal technologies and regimes of cooling (heating fuel, which can be defined using the simulation of heat exchange processes for preparing the rocket fuel.The issues of research of different technologies and simulation of cooling processes of rocket fuel with liquid nitrogen are given in [1-10]. Diagrams of temperature preparation of hydrocarbon fuel, mathematical models and characteristics of cooling fuel with its direct contact with liquid nitrogen dispersed are considered, using the numerical solution of a system of heat transfer equations, in publications [3,9].Analytical models, allowing to determine the necessary flow rate and the mass of liquid nitrogen and the cooling (heating time fuel in specific conditions and requirements, are preferred for determining design and operational characteristics of the hydrocarbon fuel cooling system.A mathematical model of the temperature preparation processes is developed. Considered characteristics of these processes are based on the analytical solutions of the equations of heat transfer and allow to define operating parameters of temperature preparation of hydrocarbon fuel in the design and operation of the filling system of launch vehicles.The paper considers a technological system to fill the launch vehicles providing the temperature preparation of hydrocarbon gases at the launch site. In this system cooling the fuel in the storage tank before filling the launch vehicle is provided by hydrocarbon fuel bubbling with liquid nitrogen. Hydrocarbon fuel is heated with a pumping station, which provides fuel circulation through the heat exchanger-heater, with

  8. Information theory in analytical chemistry

    National Research Council Canada - National Science Library

    Eckschlager, Karel; Danzer, Klaus

    1994-01-01

    Contents: The aim of analytical chemistry - Basic concepts of information theory - Identification of components - Qualitative analysis - Quantitative analysis - Multicomponent analysis - Optimum analytical...

  9. Diesel fuel to dc power: Navy & Marine Corps Applications

    Energy Technology Data Exchange (ETDEWEB)

    Bloomfield, D.P. [Analytic Power Corp., Boston, MA (United States)

    1996-12-31

    During the past year Analytic Power has tested fuel cell stacks and diesel fuel processors for US Navy and Marine Corps applications. The units are 10 kW demonstration power plants. The USN power plant was built to demonstrate the feasibility of diesel fueled PEM fuel cell power plants for 250 kW and 2.5 MW shipboard power systems. We designed and tested a ten cell, 1 kW USMC substack and fuel processor. The complete 10 kW prototype power plant, which has application to both power and hydrogen generation, is now under construction. The USN and USMC fuel cell stacks have been tested on both actual and simulated reformate. Analytic Power has accumulated operating experience with autothermal reforming based fuel processors operating on sulfur bearing diesel fuel, jet fuel, propane and natural gas. We have also completed the design and fabrication of an advanced regenerative ATR for the USMC. One of the significant problems with small fuel processors is heat loss which limits its ability to operate with the high steam to carbon ratios required for coke free high efficiency operation. The new USMC unit specifically addresses these heat transfer issues. The advances in the mill programs have been incorporated into Analytic Power`s commercial units which are now under test.

  10. Analytical challenges in characterization of high purity materials

    Indian Academy of Sciences (India)

    Unknown

    K L RAMAKUMAR. Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400 085, India ... 2. Analytical challenges in spark source mass spectrometry (SSMS). 2.1 Determination of hydrogen. Despite some of its limitations, SSMS with conventional .... Inductively coupled plasma mass spectrometry. (ICPMS).

  11. 40 CFR 86.214-94 - Analytical gases.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Analytical gases. 86.214-94 Section 86.214-94 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED... Later Model Year Gasoline-Fueled New Light-Duty Vehicles, New Light-Duty Trucks and New Medium-Duty...

  12. Doing social media analytics

    Directory of Open Access Journals (Sweden)

    Phillip Brooker

    2016-07-01

    Full Text Available In the few years since the advent of ‘Big Data’ research, social media analytics has begun to accumulate studies drawing on social media as a resource and tool for research work. Yet, there has been relatively little attention paid to the development of methodologies for handling this kind of data. The few works that exist in this area often reflect upon the implications of ‘grand’ social science methodological concepts for new social media research (i.e. they focus on general issues such as sampling, data validity, ethics, etc.. By contrast, we advance an abductively oriented methodological suite designed to explore the construction of phenomena played out through social media. To do this, we use a software tool – Chorus – to illustrate a visual analytic approach to data. Informed by visual analytic principles, we posit a two-by-two methodological model of social media analytics, combining two data collection strategies with two analytic modes. We go on to demonstrate each of these four approaches ‘in action’, to help clarify how and why they might be used to address various research questions.

  13. Thorium fuel cycle management

    International Nuclear Information System (INIS)

    Zajac, R.; Darilek, P.; Breza, J.; Necas, V.

    2010-01-01

    In this presentation author deals with the thorium fuel cycle management. Description of the thorium fuels and thorium fuel cycle benefits and challenges as well as thorium fuel calculations performed by the computer code HELIOS are presented.

  14. Nonstationary temperature field in the fuel element

    International Nuclear Information System (INIS)

    Vehauc, A.; Spasojevic, D.

    1970-03-01

    Nonstationary temperature field in the fuel element was examined for spatial and time distribution of the specific power generated in the fuel element. Analytical method was developed for calculating the temperature variation in the fuel element of a nuclear reactor for a typical shape of the heat generation function. The method is based on series expansion of the temperature field by self functions and application of Laplace transformation in time coordinate. For numerical calculation of the temperature distribution a computer code was developed based on the proposed method and applied on the ZUSE-Z-23 computer [sr

  15. Repairing fuel for reinsertion

    International Nuclear Information System (INIS)

    Krukshenk, A.

    1986-01-01

    Eqiupment for nuclear reactor fuel assembly repairing produced by Westinghouse and Brawn Bovery companies is described. Repair of failed fuel assemblies replacement of defect fuel elements gives a noticeable economical effect. Thus if the cost of a new fuel assembly is 450-500 thousand dollars, the replacement of one fuel element in it costs approximately 40-60 thousand dollars. In simple cases repairing includes either removal of failed fuel elements from a fuel assembly and its reinsertion with the rest of fuel elements into the reactor core (reactor refueling), or replacement of unfailed fuel elements from one fuel assembly to a new one (fuel assembly overhaul and reconditioning)

  16. Fuel and target programs for the transmutation at Phenix and other reactors; Programmes combustibles et cibles pour la transmutation dans Phenix et autres reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Gaillard-Groleas, G

    2002-07-01

    The fuels and targets program for transmutation, performed in the framework of the axis 1 of the December 1991 law about the researches on the management of long-lived radioactive wastes, is in perfect consistency with the transmutation scenario studies carried out in the same framework. These studies put forward the advantage of fast breeder reactors (FBR) in the incineration of minor actinides and long-lived fission products. The program includes exploratory and technological demonstration studies covering the different design options. It aims at enhancing our knowledge of the behaviour of materials under irradiation and at ensuring the mastery of processes. The goals of the different experiments foreseen at Phenix reactor are presented. The main goal is to supply a set of results allowing to precise the conditions of the technical feasibility of minor actinides and long-lived fission products incineration in FBRs. (J.S.)

  17. SIMMER-III analytic thermophysical property model

    International Nuclear Information System (INIS)

    Morita, K; Tobita, Y.; Kondo, Sa.; Fischer, E.A.

    1999-05-01

    An analytic thermophysical property model using general function forms is developed for a reactor safety analysis code, SIMMER-III. The function forms are designed to represent correct behavior of properties of reactor-core materials over wide temperature ranges, especially for the thermal conductivity and the viscosity near the critical point. The most up-to-date and reliable sources for uranium dioxide, mixed-oxide fuel, stainless steel, and sodium available at present are used to determine parameters in the proposed functions. This model is also designed to be consistent with a SIMMER-III model on thermodynamic properties and equations of state for reactor-core materials. (author)

  18. Advanced business analytics

    CERN Document Server

    Lev, Benjamin

    2015-01-01

    The book describes advanced business analytics and shows how to apply them to many different professional areas of engineering and management. Each chapter of the book is contributed by a different author and covers a different area of business analytics. The book connects the analytic principles with business practice and provides an interface between the main disciplines of engineering/technology and the organizational, administrative and planning abilities of management. It also refers to other disciplines such as economy, finance, marketing, behavioral economics and risk analysis. This book is of special interest to engineers, economists and researchers who are developing new advances in engineering management but also to practitioners working on this subject.

  19. Competing on talent analytics.

    Science.gov (United States)

    Davenport, Thomas H; Harris, Jeanne; Shapiro, Jeremy

    2010-10-01

    Do investments in your employees actually affect workforce performance? Who are your top performers? How can you empower and motivate other employees to excel? Leading-edge companies such as Google, Best Buy, Procter & Gamble, and Sysco use sophisticated data-collection technology and analysis to answer these questions, leveraging a range of analytics to improve the way they attract and retain talent, connect their employee data to business performance, differentiate themselves from competitors, and more. The authors present the six key ways in which companies track, analyze, and use data about their people-ranging from a simple baseline of metrics to monitor the organization's overall health to custom modeling for predicting future head count depending on various "what if" scenarios. They go on to show that companies competing on talent analytics manage data and technology at an enterprise level, support what analytical leaders do, choose realistic targets for analysis, and hire analysts with strong interpersonal skills as well as broad expertise.

  20. Advances in analytical chemistry

    Science.gov (United States)

    Arendale, W. F.; Congo, Richard T.; Nielsen, Bruce J.

    1991-01-01

    Implementation of computer programs based on multivariate statistical algorithms makes possible obtaining reliable information from long data vectors that contain large amounts of extraneous information, for example, noise and/or analytes that we do not wish to control. Three examples are described. Each of these applications requires the use of techniques characteristic of modern analytical chemistry. The first example, using a quantitative or analytical model, describes the determination of the acid dissociation constant for 2,2'-pyridyl thiophene using archived data. The second example describes an investigation to determine the active biocidal species of iodine in aqueous solutions. The third example is taken from a research program directed toward advanced fiber-optic chemical sensors. The second and third examples require heuristic or empirical models.

  1. Nuclear fuel performance evaluation. Final report

    International Nuclear Information System (INIS)

    Boerresen, S.; Pomeroy, D.L.; Rolstad, E.; Sauar, T.O.

    1977-06-01

    An evaluation has been made of the ability of Scandpower's empirical fuel performance model POSHO (''Power Shock'') to predict the probability of fuel pin failures resulting from pellet-clad interaction in commercial nuclear power plants. POSHO provides an analytical method to calculate the failure probabilities associated with power level maneuvers for different fuel assembly designs. Application of the method provides a basis for risk-benefit decisions concerning operational procedures, fuel designs and fuel management strategies. One boiling water reactor (BWR) and one pressurized water reactor (PWR) were selected for study to compare model predictions with actual failures, as determined from post irradiation examination of the fuel and activity release data. The fuel duty cycles were reconstructed from operating records and nodal power histories were created by using Scandpower's FMS computer programs. Nodal power histories, coupled with the relative pin power distribution in each node, were processed by the fuel failure prediction model, which tracks the interaction power level for each pin group in each node and calculates the power shocks and the probability for pellet-clad interaction cracks. The results of these calculations are processed statistically to give the expected number of cracks, the number of failed fuel pins in each assembly and the total number of failed assemblies in the core. Fuel performance in the BWR, Quad Cities Unit Two, was calculated by the model in approximate agreement with the observed performance. Fuel performance in the PWR, Maine Yankee, was calculated in approximate agreement for two of the three fuel designs. The high failure rate in the third design, Type B fuel, was not calculated by the POSHO pellet-clad interaction model

  2. Analytic number theory

    CERN Document Server

    Iwaniec, Henryk

    2004-01-01

    Analytic Number Theory distinguishes itself by the variety of tools it uses to establish results, many of which belong to the mainstream of arithmetic. One of the main attractions of analytic number theory is the vast diversity of concepts and methods it includes. The main goal of the book is to show the scope of the theory, both in classical and modern directions, and to exhibit its wealth and prospects, its beautiful theorems and powerful techniques. The book is written with graduate students in mind, and the authors tried to balance between clarity, completeness, and generality. The exercis

  3. Social network data analytics

    CERN Document Server

    Aggarwal, Charu C

    2011-01-01

    Social network analysis applications have experienced tremendous advances within the last few years due in part to increasing trends towards users interacting with each other on the internet. Social networks are organized as graphs, and the data on social networks takes on the form of massive streams, which are mined for a variety of purposes. Social Network Data Analytics covers an important niche in the social network analytics field. This edited volume, contributed by prominent researchers in this field, presents a wide selection of topics on social network data mining such as Structural Pr

  4. An analytic thomism?

    Directory of Open Access Journals (Sweden)

    Daniel Alejandro Pérez Chamorro.

    2012-12-01

    Full Text Available For 50 years the philosophers of the Anglo-Saxon analytic tradition (E. Anscombre, P. Geach, A. Kenny, P. Foot have tried to follow the Thomas Aquinas School which they use as a source to surpass the Cartesian Epistemology and to develop the virtue ethics. Recently, J. Haldane has inaugurated a program of “analytical thomism” which main result until the present has been his “theory of identity mind/world”. Nevertheless, none of Thomás’ admirers has still found the means of assimilating his metaphysics of being.

  5. Foundations of predictive analytics

    CERN Document Server

    Wu, James

    2012-01-01

    Drawing on the authors' two decades of experience in applied modeling and data mining, Foundations of Predictive Analytics presents the fundamental background required for analyzing data and building models for many practical applications, such as consumer behavior modeling, risk and marketing analytics, and other areas. It also discusses a variety of practical topics that are frequently missing from similar texts. The book begins with the statistical and linear algebra/matrix foundation of modeling methods, from distributions to cumulant and copula functions to Cornish--Fisher expansion and o

  6. Design, fabrication and operating experience of Monju ex-vessel fuel storage tank

    International Nuclear Information System (INIS)

    Yokota, Yoshio; Yamagishi, Yoshiaki; Kuroha, Mitsuo; Inoue, Tatsuya

    1995-01-01

    In FBRs there are two methods of storing and cooling the spent fuel - the in-vessel storage and the ex-vessel storage. Because of the sodium leaks through the tank at the beginning of pre-operation, the utilization of the ex-vessel fuel storage tank (EVST) of some FBR plant has been changed from the ex-vessel fuel storage to the interim fuel transfer tank. This led to reactor designers focusing on the material, structure and fabrication of the carbon steel sodium storage tanks worldwide. The Monju EVST was at the final stage of the design, when the leaks occurred. The lesson learned from that experience and the domestic fabrication technology are reflected to the design and fabrication of the Monju EVST. This paper describes the design, fabrication and R and D results for the tank, and operating experience in functional test. The items to be examined are as follows: (1) Overall structure of the tank and design philosophy on the function, (2) Structure of the cover shielding plug and its design philosophy, (3) Structures of the rotating rack and its bearings, and their design philosophy, (4) Cooling method and its design philosophy, (5) Structure and fabrication of the cooling coil support inside EVST with comparison of leaked case, (6) R and D effort for items above. The fabrication of the Monju EVST started in August 1986 and it was shipped to the site in March 1990. Installation was completed in November 1990, and sodium fill after pre-heating started in 1991. The operation has been continued since September 1992. In 1996 when the first spent fuel is stored, its total functions will be examined. (author)

  7. Status of sodium cooled fast reactors with closed fuel cycle in India

    International Nuclear Information System (INIS)

    Raj, B.

    2007-01-01

    Fast reactors form the second stage of India's 3-stage nuclear power programme. The seed for India's fast reactor programme was sown through the construction of the Fast Breeder Test Reactor (FBTR) at IGCAR, Kalpakkam, that was commissioned in 1985. FBTR has operated with an unique, indigenously developed plutonium rich mixed carbide fuel, which has reached a burn up as high as 155 GWd/t without any fuel failure in the core. The sodium systems in the reactor have performed excellently. The availability of the reactor has been as high as 92% in the recent campaigns. The fuel discharged from FBTR up to 100 GWd/t has been reprocessed successfully. The experience gained in the construction, commissioning and operation of FBTR has provided the necessary confidence to launch a Prototype FBR of 500 MWe capacity (PFBR). This reactor will be fuelled by uranium, plutonium mixed oxide. The reactor construction started in 2003 and the reactor is scheduled to be commissioned by 2010. The design of the reactor has incorporated the worldwide operating experience from the FBRs and has addressed various safety issues reported in literature, besides introducing a number of innovative features which have reduced the unit energy cost and contributed to its enhanced safety. Simultaneous with the construction of the reactor, the fuel cycle of the reactor has been addressed in a comprehensive manner and construction of a fuel cycle facility has been initiated. Subsequent to the PFBR, 4 more reactors with identical design are proposed to be constructed. Various elements of reactor design are being carefully analysed with the aim of introducing innovative features towards further reduction in unit energy cost and enhancing safety in these reactors

  8. A New Method of Piping Work by Freezing Fuel Oil to Repair a Fuel Oil Pipeline

    Science.gov (United States)

    Okada, Masashi; Tateno, Masayoshi; Minowa, Kazuki; Murayama, Kouichi

    When a pipe is cut off to repair fuel oil pipelines, the oil has to be drained from the pipelines. If the oil inside the pipe is frozen at both sides of a cutting plane, it is not necessary to drain the oil from the pipelines. In the present paper, such a freezing method is studied analytically and experimentally to establish a suitable construction method, where liquid-nitrogen (LN2) is used as a coolant and fuel oil-C is used as a typical example. From the result, thermal conductivity and thermal diffusivity of the fuel oil-C in a low temperature range were measured as a function of temperature in addition to the pour point and glass transition point. Furthermore, in order to compare the agreement between analysis and experiment, an analytical method was performed under various conditions. Finally, temperatures in analytical values were agreed well with experimental ones, and suitable position and time for cutting are clarified.

  9. Modeling of hybrid vehicle fuel economy and fuel engine efficiency

    Science.gov (United States)

    Wu, Wei

    "Near-CV" (i.e., near-conventional vehicle) hybrid vehicles, with an internal combustion engine, and a supplementary storage with low-weight, low-energy but high-power capacity, are analyzed. This design avoids the shortcoming of the "near-EV" and the "dual-mode" hybrid vehicles that need a large energy storage system (in terms of energy capacity and weight). The small storage is used to optimize engine energy management and can provide power when needed. The energy advantage of the "near-CV" design is to reduce reliance on the engine at low power, to enable regenerative braking, and to provide good performance with a small engine. The fuel consumption of internal combustion engines, which might be applied to hybrid vehicles, is analyzed by building simple analytical models that reflect the engines' energy loss characteristics. Both diesel and gasoline engines are modeled. The simple analytical models describe engine fuel consumption at any speed and load point by describing the engine's indicated efficiency and friction. The engine's indicated efficiency and heat loss are described in terms of several easy-to-obtain engine parameters, e.g., compression ratio, displacement, bore and stroke. Engine friction is described in terms of parameters obtained by fitting available fuel measurements on several diesel and spark-ignition engines. The engine models developed are shown to conform closely to experimental fuel consumption and motored friction data. A model of the energy use of "near-CV" hybrid vehicles with different storage mechanism is created, based on simple algebraic description of the components. With powertrain downsizing and hybridization, a "near-CV" hybrid vehicle can obtain a factor of approximately two in overall fuel efficiency (mpg) improvement, without considering reductions in the vehicle load.

  10. Fuel element

    International Nuclear Information System (INIS)

    Hirose, Yasuo.

    1982-01-01

    Purpose: To increase the plenum space in a fuel element used for a liquid metal cooled reactor. Constitution: A fuel pellet is secured at one end with an end plug and at the other with a coil spring in a tubular container. A mechanism for fixing the coil spring composed of a tubular unit is mounted by friction with the inner surface of the tubular container. Accordingly, the recoiling force of the coil spring can be retained by fixing mechanism with a small volume, and since a large amount of plenum space can be obtained, the internal pressure rise in the cladding tube can be suppressed even if large quantities of fission products are discharged. (Kamimura, M.)

  11. Fuel trading

    International Nuclear Information System (INIS)

    2015-01-01

    A first part of this report proposes an overview of trends and predictions. After a synthesis on the sector changes and trends, it indicates and comments the most recent predictions for the consumption of refined oil products and for the turnover of the fuel wholesale market, reports the main highlights concerning the sector's life, and gives a dashboard of the sector activity. The second part proposes the annual report on trends and competition. It presents the main operator profiles and fuel categories, the main determining factors of the activity, the evolution of the sector context between 2005 and 2015 (consumptions, prices, temperature evolution). It analyses the evolution of the sector activity and indicators (sales, turnovers, prices, imports). Financial performances of enterprises are presented. The economic structure of the sector is described (evolution of the economic fabric, structural characteristics, French foreign trade). Actors are then presented and ranked in terms of turnover, of added value, and of result

  12. Analytics: Changing the Conversation

    Science.gov (United States)

    Oblinger, Diana G.

    2013-01-01

    In this third and concluding discussion on analytics, the author notes that we live in an information culture. We are accustomed to having information instantly available and accessible, along with feedback and recommendations. We want to know what people think and like (or dislike). We want to know how we compare with "others like me."…

  13. Social Learning Analytics

    Science.gov (United States)

    Buckingham Shum, Simon; Ferguson, Rebecca

    2012-01-01

    We propose that the design and implementation of effective "Social Learning Analytics (SLA)" present significant challenges and opportunities for both research and enterprise, in three important respects. The first is that the learning landscape is extraordinarily turbulent at present, in no small part due to technological drivers.…

  14. Explanatory analytics in OLAP

    NARCIS (Netherlands)

    Caron, E.A.M.; Daniëls, H.A.M.

    2013-01-01

    In this paper the authors describe a method to integrate explanatory business analytics in OLAP information systems. This method supports the discovery of exceptional values in OLAP data and the explanation of such values by giving their underlying causes. OLAP applications offer a support tool for

  15. Ada & the Analytical Engine.

    Science.gov (United States)

    Freeman, Elisabeth

    1996-01-01

    Presents a brief history of Ada Byron King, Countess of Lovelace, focusing on her primary role in the development of the Analytical Engine--the world's first computer. Describes the Ada Project (TAP), a centralized World Wide Web site that serves as a clearinghouse for information related to women in computing, and provides a Web address for…

  16. History of analytic geometry

    CERN Document Server

    Boyer, Carl B

    2012-01-01

    Designed as an integrated survey of the development of analytic geometry, this study presents the concepts and contributions from before the Alexandrian Age through the eras of the great French mathematicians Fermat and Descartes, and on through Newton and Euler to the "Golden Age," from 1789 to 1850.

  17. Analytic number theory

    CERN Document Server

    Matsumoto, Kohji

    2002-01-01

    The book includes several survey articles on prime numbers, divisor problems, and Diophantine equations, as well as research papers on various aspects of analytic number theory such as additive problems, Diophantine approximations and the theory of zeta and L-function Audience Researchers and graduate students interested in recent development of number theory

  18. Analytical Chemistry Laboratory

    Science.gov (United States)

    Anderson, Mark

    2013-01-01

    The Analytical Chemistry and Material Development Group maintains a capability in chemical analysis, materials R&D failure analysis and contamination control. The uniquely qualified staff and facility support the needs of flight projects, science instrument development and various technical tasks, as well as Cal Tech.

  19. Social Data Analytics Tool

    DEFF Research Database (Denmark)

    Hussain, Abid; Vatrapu, Ravi

    2014-01-01

    This paper presents the design, development and demonstrative case studies of the Social Data Analytics Tool, SODATO. Adopting Action Design Framework [1], the objective of SODATO [2] is to collect, store, analyze, and report big social data emanating from the social media engagement of and socia...

  20. User Behavior Analytics

    Energy Technology Data Exchange (ETDEWEB)

    Turcotte, Melissa [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Moore, Juston Shane [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-28

    User Behaviour Analytics is the tracking, collecting and assessing of user data and activities. The goal is to detect misuse of user credentials by developing models for the normal behaviour of user credentials within a computer network and detect outliers with respect to their baseline.

  1. Environmental analytical chemistry

    International Nuclear Information System (INIS)

    Gasco Sanchez, L.

    1990-01-01

    The environmental analytical chemistry has a big relation with the stochastic methods. The environmetry is an interdisciplinary science that is formed by the computer science, statistics science and environmental science. Today we must apply the logic of the laboratory and with the environmetry we can apply better the chemical analysis into the environmental control and pollutants control

  2. Analytics for Customer Engagement

    NARCIS (Netherlands)

    Bijmolt, Tammo H. A.; Leeflang, Peter S. H.; Block, Frank; Eisenbeiss, Maik; Hardie, Bruce G. S.; Lemmens, Aurelie; Saffert, Peter

    In this article, we discuss the state of the art of models for customer engagement and the problems that are inherent to calibrating and implementing these models. The authors first provide an overview of the data available for customer analytics and discuss recent developments. Next, the authors

  3. Analytics in Higher Education

    Science.gov (United States)

    Universities UK, 2016

    2016-01-01

    Learning analytics provide a set of powerful tools to inform and support learners. They enable institutions and individuals to better understand and predict personal learning needs and performance. Universities already collect vast amounts of data about their student populations, but often this is underutilised. The current "state of the…

  4. Solid TRU fuels and fuel cycle technology

    International Nuclear Information System (INIS)

    Ogawa, Toru; Suzuki, Yasufumi

    1997-01-01

    Alloys and nitrides are candidate solid fuels for transmutation. However, the nitride fuels are preferred to the alloys because they have more favorable thermal properties which allows to apply a cold-fuel concept. The nitride fuel cycle technology is briefly presented

  5. Caracterización mecánica de recubrimientos de aluminio por CVD-FBR sobre aceros inoxidables y resistencia a la oxidación en vapor de agua

    Directory of Open Access Journals (Sweden)

    Diego Pérez-Muñoz

    2015-09-01

    Full Text Available Los recubrimientos de aluminio depositados sobre el acero inoxidable austenítico AISI 317 por Deposición Química de Vapor en Lecho Fluidizado (CVD-FBR presentan a altas temperaturas una reducción de la velocidad de corrosión de más de 80 veces. Se realizó la caracterización mecánica de los recubrimientos por medio de microdureza, nanoindentación, para conocer cómo se vieron afectas las propiedades mecánicas (en especial la dureza y el módulo de Young del recubrimiento y del sustrato luego de ser sometidos a la oxidación a alta temperatura. También se hicieron análisis por medio de Microscopia Electrónica de Barrido (MEB, para observar los cambios microestructurales, y de Microscopia de Fuerza Atómica (MFA, para observar cómo varía la topografía y el gradiente de rugosidad en función de la distancia recorrida por la punta del cantiléver durante los barridos.

  6. Analysis of Double-encapsulated Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Medvedev, Pavel G [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Perez, Danielle Marie [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory

    2014-09-01

    In an LWR fuel rod, the cladding encapsulates the fuel, contains fission products, and transfers heat directly to the water coolant. In some situations, it may be advantageous to separate the cladding from the coolant through use of a secondary cladding or capsule. This may be done to increase confidence that the fuel or fission products will not mix with the coolant, to provide a mechanism for controlling the rod temperature, or to place multiple experimental rodlets within a single housing. With an axisymmetric assumption, it is possible to derive closed-form expressions for the temperature profile in a fuel rod using radially-constant thermal conductivity in the fuel. This is true for both a traditional fuel-cladding rod and a double-encapsulated fuel (fuel, cladding, capsule) configuration. Likewise, it is possible to employ a fuel performance code to analyse both a traditional and a double-encapsulated fuel. In the case of the latter, two sets of gap heat transfer conditions must be imposed. In this work, we review the equations associated with radial heat transfer in a cylindrical system, present analytic and computational results for a postulated power and gas mixture history for IFA-744, and describe the analysis of the AFC-2A, 2B metallic fuel alloy experiments at the Advanced Test Reactor, including the effect of a release of fission products into the cladding-capsule gap. The computational results for these two cases were obtained using BISON, a fuel performance code under development at Idaho National Laboratory.

  7. Blockages in LMFBR fuel assemblies: a review

    International Nuclear Information System (INIS)

    Han, J.T.; Fontana, M.H.

    1977-01-01

    Experimental and analytical investigations performed in the United States, Germany, Great Britain, and Japan on the effects of partial flow blockages in liquid-metal fast breeder reactor fuel assemblies are reviewed and the results presented. Generalized models are developed from experimental data obtained for blockages of various sizes, shapes, and porosity, with and without pins, utilizing water and sodium as the coolant. Generally, the recirculating flow in the wake behind a blockage is a relatively effective heat transfer mechanism. Experiments where sodium boiling was made to occur behind the blockages indicate that boiling is stable for the configurations tested; these results are predicted by analytical models. Blockages at the inlet of fuel assemblies tend to have insignificant effects in the fuel assembly unless flow is reduced grossly and therefore would be detectable. Blockages in the heat generating zone have to be quite large to cause sodium boiling under normal reactor operating conditions

  8. Blockages in LMFBR fuel assemblies: a review

    Energy Technology Data Exchange (ETDEWEB)

    Han, J T; Fontana, M H

    1977-01-01

    Experimental and analytical investigations performed in the United States, Germany, Great Britain, and Japan on the effects of partial flow blockages in liquid-metal fast breeder reactor fuel assemblies are reviewed and the results presented. Generalized models are developed from experimental data obtained for blockages of various sizes, shapes, and porosity, with and without pins, utilizing water and sodium as the coolant. Generally, the recirculating flow in the wake behind a blockage is a relatively effective heat transfer mechanism. Experiments where sodium boiling was made to occur behind the blockages indicate that boiling is stable for the configurations tested; these results are predicted by analytical models. Blockages at the inlet of fuel assemblies tend to have insignificant effects in the fuel assembly unless flow is reduced grossly and therefore would be detectable. Blockages in the heat generating zone have to be quite large to cause sodium boiling under normal reactor operating conditions.

  9. Fuel rod and fuel assembly

    International Nuclear Information System (INIS)

    Takekawa, Tetsuya.

    1993-01-01

    Burnable poisons are contained in a portion of a pellet constituting a fuel rod. A distribution density of the burnable poison-containing pellets and a concentration of the burnable poisons in the pellet are varied depending on the axial position of the fuel rod. That is, the distribution density of the burnable poison containing-pellets is increased at the central portion of the fuel rod and it is decreased at both ends thereof, and a concentration of the burnable poisons of the burnable poison containing-pellet disposed at the end portions thereof is decreased to less than a concentration of the burnable poison-containing pellet at the central portion. With such a constitution, a central peaking at an early stage of the combustion cycle is decreased. Accordingly, power at the central portion is increased than that in the end portions at the latter half of the cycle, to flatten the power distribution. Further, a burnable poison concentration of the pellets at the end portions is decreased to promote burning of burnable poisons at the end portions which are less burnable relatively, thereby enabling to prevent worsening of neutron economy. (T.M.)

  10. Fuel element loading system

    International Nuclear Information System (INIS)

    Arya, S.P; s.

    1978-01-01

    A nuclear fuel element loading system is described which conveys a plurality of fuel rods to longitudinal passages in fuel elements. Conveyor means successively position the fuel rods above the longitudinal passages in axial alignment therewith and adapter means guide the fuel rods from the conveyor means into the longitudinal passages. The fuel elements are vibrated to cause the fuel rods to fall into the longitudinal passages through the adapter means

  11. Proceedings of the 6. international conference on stability and handling of liquid fuels. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Giles, H.N. [ed.] [Deputy Assistant Secretary for Strategic Petroleum Reserve, Washington, DC (United States). Operations and Readiness Office

    1998-12-01

    Volume 2 of these proceedings contain 42 papers arranged under the following topical sections: Fuel blending and compatibility; Middle distillates; Microbiology; Alternative fuels; General topics (analytical methods, tank remediation, fuel additives, storage stability); and Poster presentations (analysis methods, oxidation kinetics, health problems).

  12. Thermochemical data and its use in modeling chemical behavior in mixed-oxide fuel pins

    International Nuclear Information System (INIS)

    Gibby, R.L.; Woodley, R.E.; Adamson, M.G.; Johnson, C.E.

    1979-01-01

    The status of US activities to obtain fuel chemistry data is reviewed. Analytical expressions addressing basic needs of all fuel chemistry models are presented. Fission product concentrations during irradiation, oxygen-to-metal (O/M) at beginning-of-life and at burnup, and the potential in fuel-cladding gap at burnup are described

  13. Nuclear analytical chemistry

    International Nuclear Information System (INIS)

    Brune, D.; Forkman, B.; Persson, B.

    1984-01-01

    This book covers the general theories and techniques of nuclear chemical analysis, directed at applications in analytical chemistry, nuclear medicine, radiophysics, agriculture, environmental sciences, geological exploration, industrial process control, etc. The main principles of nuclear physics and nuclear detection on which the analysis is based are briefly outlined. An attempt is made to emphasise the fundamentals of activation analysis, detection and activation methods, as well as their applications. The book provides guidance in analytical chemistry, agriculture, environmental and biomedical sciences, etc. The contents include: the nuclear periodic system; nuclear decay; nuclear reactions; nuclear radiation sources; interaction of radiation with matter; principles of radiation detectors; nuclear electronics; statistical methods and spectral analysis; methods of radiation detection; neutron activation analysis; charged particle activation analysis; photon activation analysis; sample preparation and chemical separation; nuclear chemical analysis in biological and medical research; the use of nuclear chemical analysis in the field of criminology; nuclear chemical analysis in environmental sciences, geology and mineral exploration; and radiation protection

  14. Analytical applications of aptamers

    Science.gov (United States)

    Tombelli, S.; Minunni, M.; Mascini, M.

    2007-05-01

    Aptamers are single stranded DNA or RNA ligands which can be selected for different targets starting from a library of molecules containing randomly created sequences. Aptamers have been selected to bind very different targets, from proteins to small organic dyes. Aptamers are proposed as alternatives to antibodies as biorecognition elements in analytical devices with ever increasing frequency. This in order to satisfy the demand for quick, cheap, simple and highly reproducible analytical devices, especially for protein detection in the medical field or for the detection of smaller molecules in environmental and food analysis. In our recent experience, DNA and RNA aptamers, specific for three different proteins (Tat, IgE and thrombin), have been exploited as bio-recognition elements to develop specific biosensors (aptasensors). These recognition elements have been coupled to piezoelectric quartz crystals and surface plasmon resonance (SPR) devices as transducers where the aptamers have been immobilized on the gold surface of the crystals electrodes or on SPR chips, respectively.

  15. Analytical chemists and dinosaurs

    International Nuclear Information System (INIS)

    Brooks, R.R.

    1987-01-01

    The role of the analytical chemist in the development of the extraterrestrial impact theory for mass extinctions at the terminal Cretaceous Period is reviewed. High iridium concentrations in Cretaceous/Tertiary boundary clays have been linked to a terrestrial impact from an iridium-rich asteroid or large meteorite som 65 million years ago. Other evidence in favour of the occurrence of such an impact has been provided by the detection of shocked quartz grains originating from impact and of amorphous carbon particles similar to soot, derived presumably from wordwide wildfires at the terminal Cretaceous. Further evidence provided by the analytical chemist involves the determination of isotopic ratios such as 144 Nd/ 143 Nd, 187 Os/ 186 Os, and 87 Sr/ 86 Sr. Countervailing arguments put forward by the gradualist school (mainly palaeontological) as opposed to the catastrophists (mainly chemists and geochemists) are also presented and discussed

  16. Multifunctional nanoparticles: Analytical prospects

    International Nuclear Information System (INIS)

    Dios, Alejandro Simon de; Diaz-Garcia, Marta Elena

    2010-01-01

    Multifunctional nanoparticles are among the most exciting nanomaterials with promising applications in analytical chemistry. These applications include (bio)sensing, (bio)assays, catalysis and separations. Although most of these applications are based on the magnetic, optical and electrochemical properties of multifunctional nanoparticles, other aspects such as the synergistic effect of the functional groups and the amplification effect associated with the nanoscale dimension have also been observed. Considering not only the nature of the raw material but also the shape, there is a huge variety of nanoparticles. In this review only magnetic, quantum dots, gold nanoparticles, carbon and inorganic nanotubes as well as silica, titania and gadolinium oxide nanoparticles are addressed. This review presents a narrative summary on the use of multifuncional nanoparticles for analytical applications, along with a discussion on some critical challenges existing in the field and possible solutions that have been or are being developed to overcome these challenges.

  17. Analytical caustic surfaces

    Science.gov (United States)

    Schmidt, R. F.

    1987-01-01

    This document discusses the determination of caustic surfaces in terms of rays, reflectors, and wavefronts. Analytical caustics are obtained as a family of lines, a set of points, and several types of equations for geometries encountered in optics and microwave applications. Standard methods of differential geometry are applied under different approaches: directly to reflector surfaces, and alternatively, to wavefronts, to obtain analytical caustics of two sheets or branches. Gauss/Seidel aberrations are introduced into the wavefront approach, forcing the retention of all three coefficients of both the first- and the second-fundamental forms of differential geometry. An existing method for obtaining caustic surfaces through exploitation of the singularities in flux density is examined, and several constant-intensity contour maps are developed using only the intrinsic Gaussian, mean, and normal curvatures of the reflector. Numerous references are provided for extending the material of the present document to the morphologies of caustics and their associated diffraction patterns.

  18. Multifunctional nanoparticles: analytical prospects.

    Science.gov (United States)

    de Dios, Alejandro Simón; Díaz-García, Marta Elena

    2010-05-07

    Multifunctional nanoparticles are among the most exciting nanomaterials with promising applications in analytical chemistry. These applications include (bio)sensing, (bio)assays, catalysis and separations. Although most of these applications are based on the magnetic, optical and electrochemical properties of multifunctional nanoparticles, other aspects such as the synergistic effect of the functional groups and the amplification effect associated with the nanoscale dimension have also been observed. Considering not only the nature of the raw material but also the shape, there is a huge variety of nanoparticles. In this review only magnetic, quantum dots, gold nanoparticles, carbon and inorganic nanotubes as well as silica, titania and gadolinium oxide nanoparticles are addressed. This review presents a narrative summary on the use of multifunctional nanoparticles for analytical applications, along with a discussion on some critical challenges existing in the field and possible solutions that have been or are being developed to overcome these challenges. 2010 Elsevier B.V. All rights reserved.

  19. Big Data Analytics

    DEFF Research Database (Denmark)

    Buch, Rasmus Brødsgaard; Beheshti-Kashi, Samaneh; Nielsen, Thomas Alexander Sick

    2018-01-01

    With the growth of textual data and the simultaneous advancements in Text Analytics enabling the exploitation of this huge amount of unstructured data, companies are provided with the opportunity to tap into the previously hidden knowledge. However, how to use this valuable source, still...... is not unveiled for various domains, such as also for the transportation sector. Accordingly, this research aims at examining the potential of textual data in transportation. For this purpose, a case study was designed on public opinion towards the adoption of driverless cars. This case study was framed together...... and tweets using topic modelling, document classification and sentiment analysis. These analyses have for instance shown that Text Analytics may be a supplementary tool to surveys, since they may extract additional knowledge which may not be captured through the application of surveys. In this case...

  20. Industrial Analytics Corporation

    Energy Technology Data Exchange (ETDEWEB)

    Industrial Analytics Corporation

    2004-01-30

    The lost foam casting process is sensitive to the properties of the EPS patterns used for the casting operation. In this project Industrial Analytics Corporation (IAC) has developed a new low voltage x-ray instrument for x-ray radiography of very low mass EPS patterns. IAC has also developed a transmitted visible light method for characterizing the properties of EPS patterns. The systems developed are also applicable to other low density materials including graphite foams.

  1. Analytical and physical electrochemistry

    CERN Document Server

    Girault, Hubert H

    2004-01-01

    The study of electrochemistry is pertinent to a wide variety of fields, including bioenergetics, environmental sciences, and engineering sciences. In addition, electrochemistry plays a fundamental role in specific applications as diverse as the conversion and storage of energy and the sequencing of DNA.Intended both as a basic course for undergraduate students and as a reference work for graduates and researchers, Analytical and Physical Electrochemistry covers two fundamental aspects of electrochemistry: electrochemistry in solution and interfacial electrochemistry. By bringing these two subj

  2. Inorganic Analytical Chemistry

    DEFF Research Database (Denmark)

    Berg, Rolf W.

    The book is a treatise on inorganic analytical reactions in aqueous solution. It covers about half of the elements in the periodic table, i.e. the most important ones : H, Li, B, C, N, O, Na, Mg, Al, P, S, Cl, K, Ca, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, As, Se, Br, Sr, Mo, Ag, Cd, Sn, Sb, I, Ba, W,...

  3. Competing on analytics.

    Science.gov (United States)

    Davenport, Thomas H

    2006-01-01

    We all know the power of the killer app. It's not just a support tool; it's a strategic weapon. Companies questing for killer apps generally focus all their firepower on the one area that promises to create the greatest competitive advantage. But a new breed of organization has upped the stakes: Amazon, Harrah's, Capital One, and the Boston Red Sox have all dominated their fields by deploying industrial-strength analytics across a wide variety of activities. At a time when firms in many industries offer similar products and use comparable technologies, business processes are among the few remaining points of differentiation--and analytics competitors wring every last drop of value from those processes. Employees hired for their expertise with numbers or trained to recognize their importance are armed with the best evidence and the best quantitative tools. As a result, they make the best decisions. In companies that compete on analytics, senior executives make it clear--from the top down--that analytics is central to strategy. Such organizations launch multiple initiatives involving complex data and statistical analysis, and quantitative activity is managed atthe enterprise (not departmental) level. In this article, professor Thomas H. Davenport lays out the characteristics and practices of these statistical masters and describes some of the very substantial changes other companies must undergo in order to compete on quantitative turf. As one would expect, the transformation requires a significant investment in technology, the accumulation of massive stores of data, and the formulation of company-wide strategies for managing the data. But, at least as important, it also requires executives' vocal, unswerving commitment and willingness to change the way employees think, work, and are treated.

  4. Supramolecular analytical chemistry.

    Science.gov (United States)

    Anslyn, Eric V

    2007-02-02

    A large fraction of the field of supramolecular chemistry has focused in previous decades upon the study and use of synthetic receptors as a means of mimicking natural receptors. Recently, the demand for synthetic receptors is rapidly increasing within the analytical sciences. These classes of receptors are finding uses in simple indicator chemistry, cellular imaging, and enantiomeric excess analysis, while also being involved in various truly practical assays of bodily fluids. Moreover, one of the most promising areas for the use of synthetic receptors is in the arena of differential sensing. Although many synthetic receptors have been shown to yield exquisite selectivities, in general, this class of receptor suffers from cross-reactivities. Yet, cross-reactivity is an attribute that is crucial to the success of differential sensing schemes. Therefore, both selective and nonselective synthetic receptors are finding uses in analytical applications. Hence, a field of chemistry that herein is entitled "Supramolecular Analytical Chemistry" is emerging, and is predicted to undergo increasingly rapid growth in the near future.

  5. Research on plant of metal fuel fabrication using casting process

    International Nuclear Information System (INIS)

    Senda, Yasuhide; Mori, Yukihide

    2003-12-01

    This document presents the plant concept of metal fuel fabrication system (38tHM/y) using casting process in electrolytic recycle, which based on recent studies of its equipment design and quality control system. And we estimate the cost of its construction and operation, including costs of maintenance, consumed hardware and management of waste. The content of this work is as follows. (1) Designing of fuel fabrication equipment: We make material flow diagrams of the fuel fabrication plant and rough designs of the injection casting furnace, demolder and inspection equipment. (2) Designing of resolution system of liquid waste, which comes from analytical process facility. Increased analytical items, we rearrange analytical process facility, estimate its chemicals and amount of waste. (3) Arrangement of equipments: We made a arrangement diagram of the metal fuel fabrication equipments in cells. (4) Estimation of cost data: We estimated cost to construct the facility and to operate it. (author)

  6. Nuclear Fuel elements

    International Nuclear Information System (INIS)

    Hirakawa, Hiromasa.

    1979-01-01

    Purpose: To reduce the stress gradient resulted in the fuel can in fuel rods adapted to control the axial power distribution by the combination of fuel pellets having different linear power densities. Constitution: In a fuel rod comprising a first fuel pellet of a relatively low linear power density and a second fuel pellet of a relatively high linear power density, the second fuel pellet is cut at its both end faces by an amount corresponding to the heat expansion of the pellet due to the difference in the linear power density to the adjacent first fuel pellet. Thus, the second fuel pellet takes a smaller space than the first fuel pellet in the fuel can. This can reduce the stress produced in the portion of the fuel can corresponding to the boundary between the adjacent fuel pellets. (Kawakami, Y.)

  7. Costing of spent nuclear fuel storage

    International Nuclear Information System (INIS)

    2009-01-01

    This report deals with economic analysis and cost estimation, based on exploration of relevant issues, including a survey of analytical tools for assessment and updated information on the market and financial issues associated with spent fuel storage. The development of new storage technologies and changes in some of the circumstances affecting the costs of spent fuel storage are also incorporated. This report aims to provide comprehensive information on spent fuel storage costs to engineers and nuclear professionals as well as other stakeholders in the nuclear industry. This report is meant to provide informative guidance on economic aspects involved in selecting a spent fuel storage system, including basic methods of analysis and cost data for project evaluation and comparison of storage options, together with financial and business aspects associated with spent fuel storage. After the review of technical options for spent fuel storage in Section 2, cost categories and components involved in the lifecycle of a storage facility are identified in Section 3 and factors affecting costs of spent fuel storage are then reviewed in the Section 4. Methods for cost estimation and analysis are introduced in Section 5, and other financial and business aspects associated with spent fuel storage are discussed in Section 6.

  8. Overview of chemical characterization of FBTR fuel

    International Nuclear Information System (INIS)

    Venkatesan, V.; Nandi, C.; Patil, A.B.; Prakash, Amrit; Khan, K.B.; Arun Kumar

    2015-01-01

    Uranium Plutonium mixed carbide fuel is the driver fuel for Fast Breeder Test Reactor (FBTR) at IGCAR. The fuel is being fabricated at Radiometallurgy Division, BARC by conventional powder metallurgy route. During the fabrication of fuel, chemical quality control of process intermediates is very important to reach stringent specification of the final fuel product. Different steps are involved in the fabrication of uranium-plutonium carbide (MC) for FBTR. The main steps in the fabrication of MC fuel pellets are carbothermic reduction (CR) of mixture of uranium oxide, plutonium oxide and graphite powder to prepare MC clinkers, crushing and milling of MC clinkers and consolidation of MC powders into fuel pellets and sintering. As a part of process control, analysis of uranium (U), plutonium (Pu), carbon in oxide graphite mixture and U, Pu, carbon, oxygen, nitrogen, MC, M 2 C 3 contents in mixed carbide powder (MC clinkers) are carried out at our laboratory. Analysis of U, Pu, carbon, oxygen, nitrogen, MC and M 2 C 3 contents in mixed carbide sintered pellets are carried out as a part of quality control. This paper describes an overview of analytical instruments used during chemical quality control of mixed carbide fuel

  9. Business analytics a practitioner's guide

    CERN Document Server

    Saxena, Rahul

    2013-01-01

    This book provides a guide to businesses on how to use analytics to help drive from ideas to execution. Analytics used in this way provides "full lifecycle support" for business and helps during all stages of management decision-making and execution.The framework presented in the book enables the effective interplay of business, analytics, and information technology (business intelligence) both to leverage analytics for competitive advantage and to embed the use of business analytics into the business culture. It lays out an approach for analytics, describes the processes used, and provides gu

  10. Hanford analytical sample projections FY 1998--FY 2002

    Energy Technology Data Exchange (ETDEWEB)

    Joyce, S.M.

    1998-02-12

    Analytical Services projections are compiled for the Hanford site based on inputs from the major programs for the years 1998 through 2002. Projections are categorized by radiation level, protocol, sample matrix and program. Analyses requirements are also presented. This document summarizes the Hanford sample projections for fiscal years 1998 to 2002. Sample projections are based on inputs submitted to Analytical Services covering Environmental Restoration, Tank Waste Remediation Systems (TWRS), Solid Waste, Liquid Effluents, Spent Nuclear Fuels, Transition Projects, Site Monitoring, Industrial Hygiene, Analytical Services and miscellaneous Hanford support activities. In addition, details on laboratory scale technology (development) work, Sample Management, and Data Management activities are included. This information will be used by Hanford Analytical Services (HAS) and the Sample Management Working Group (SMWG) to assure that laboratories and resources are available and effectively utilized to meet these documented needs.

  11. CANSAR. Analytical irradiation for PCI analysis

    International Nuclear Information System (INIS)

    Bruet, M.; Lemaignan, C.

    1984-01-01

    The aim of ''CANSAR'' analytical irradiation is to evaluate the various mechanisms expected to be active during PCI failures (local concentration of fission products, fuel expansion, stress concentration induced by fuel fragment relocation, etc.). Two identical test pins, similar to classical PWR pins, but shorter, will be power-ramped in parallel. They will be filled with fuel pellets machined in various ways in order to simulate pellet fracture, relocation and preferential fission product migration path. One pin is highly instrumented with fission gas analysis, centre-line temperature and strain gauges on the cladding. The other can be unloaded between the pile cycles to perform other measurements such as diameter change, eddy currents, hot cell γ scanning. The gauges are necessary to obtain valuable information on cladding stresses. However, they induce significant modification of the thermal and mechanical behaviour of the cladding. Extensive finite element computation has been undertaken to estimate the temperature shift and the cladding reinforcement due to the gauges. Details of this work performed to design and implement the experiment will be presented. This included, in particular, high precision machining of UO 2 sectors to obtain ''precracked'' pellets and computation of the thermo-mechanical behaviour of the cladding with the gauges. (author)

  12. Recovery of uranium from analytical waste solution

    International Nuclear Information System (INIS)

    Kumar, Pradeep; Anitha, M.; Singh, D.K.

    2016-01-01

    Dispersion fuels are considered as advance fuel for the nuclear reactor. Liquid waste containing significant quantity of uranium gets generated during chemical characterization of dispersion fuel. The present paper highlights the effort in devising a counter current solvent extraction process based on the synergistic mixture of D2EHPA and Cyanex 923 to recover uranium from such waste solutions. A typical analytical waste solution was found to have the following composition: U 3 O 8 (∼3 g/L), Al: 0.3 g/L, V: 15 ppm, Phosphoric acid: 3M, sulphuric acid : 1M and nitric acid : 1M. The aqueous solution is composed of mixture of either 3M phosphoric acid and 1M sulphuric acid or 1M sulphuric acid and 1M nitric acid, keeping metallic concentrations in the above mentioned range. Different organic solvents were tested. Based on the higher extraction of uranium with synergistic mixture of 0.5M D2EHPA + 0.125M Cyanex 923, it was selected for further investigation in the present work

  13. Fuel damage during off-normal transients in metal-fueled fast reactors

    International Nuclear Information System (INIS)

    Kramer, J.M.; Bauer, T.H.

    1990-01-01

    Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs

  14. Perspective on the French closed fuel cycle: Open towards energy future and sustainability

    International Nuclear Information System (INIS)

    Tinturier, Bernard; Debes, Michel; Delbecq, Jean-Michel

    2006-01-01

    Energy sustainability and nuclear energy nowadays are far reaching issues with many implications and as a consequence, any long term sustainable strategy needs to be flexible. In France, nuclear energy (427 TWh in 2004, 80% of national electricity production) is a major asset for clean electricity production and for meeting Kyoto protocol objective in France. The decision to build a future EPR reactor in France has been taken. Regarding back end and fuel cycle, the current reprocessing and recycling strategy, with the existing industrial system (Cogema La Hague and Melox), has proven to be reliable and efficient. It enables to meet sustainability requirements, now and in the long run: ensuring a good management of high level waste through vitrification, reducing the amount of nuclear spent fuel in interim storage, recycling valuable nuclear material (Pu), keeping the possibility to use Pu concentrated in MOX spent fuel to start FBR in the future. To maintain this possibility for the far future, EDF considers that the Generation IV program is of major importance in order to develop future fast reactors able to use plutonium and to ensure a full utilization of uranium resource, while optimizing high level waste management. EDF strategy is to keep the nuclear option open in the future, with a two-steps approach for the renewal of the current nuclear fleet: first, around 2020, with the launching of generation III reactors like EPR, and second, depending on the energy demand, launching of Generation IV systems, around 2040 or beyond. To meet this energy prospect, R and D efforts must be devoted to fast breeder reactors (sodium cooled, which benefits already from industrial experience, and gas cooled, under consideration for R and D). Globally, this strategy is open to future progress and optimisation as needed to meet long term energy sustainability. It appears the necessity of a good consistency between all the components of the nuclear system: reactors, fuel cycle

  15. Elements of nuclear reactor fueling theory

    International Nuclear Information System (INIS)

    Egan, M.R.

    1984-01-01

    Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behaviour and physical requirements of operating cycle sequences and fueling strategies having practical use in fuel management. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and manoeuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy. Numerical evaluations of degenerate equilibrium cycle sequences are then performed for a typical PWR core, and accompanying fuel cycle costs are calculated. The impact of design and operational limits as constraints on the performance mappings for this reactor are also studied with respect to achieving improved cost performance from the once-through fuel cycle. The dynamics of transition cycle sequences are then examined using the generalized theory. Proof of the existence of non-degenerate equilibrium cycle sequences is presented when the mechanics of the fixed reload batch size strategy are developed analytically for transition sequences. Finally, an analysis of the fixed reload enrichment strategy demonstrates the potential for convergence of the transition sequence to a fully degenerate equilibrium sequence. (author)

  16. Low enriched uranium fuel conversion and fuel shipping guide

    International Nuclear Information System (INIS)

    1997-01-01

    The analysis of reactor core physics and thermal hydraulics was completed in 1993. A supplement to the Final Safety Analysis Report describing the results of these analyses was submitted to the Nuclear Regulatory Commission along with proposed Technical Specifications in May, 1993. Discussions with the NRC staff led to a submittal of revised proposed Technical Specifications in February, 1994. The analytical work is complete. A second portion of the grant was to develop a fuel shipping guide for university research reactors. Such a guide was developed and is available for use by the research reactor community

  17. ANALYTIC SOLUTIONS OF MATRIX RICCATI EQUATIONS WITH ANALYTIC COEFFICIENTS

    NARCIS (Netherlands)

    Curtain, Ruth; Rodman, Leiba

    2010-01-01

    For matrix Riccati equations of platoon-type systems and of systems arising from PDEs, assuming the coefficients are analytic or rational functions in a suitable domain, analyticity of the stabilizing solution is proved under various hypotheses. General results on analytic behavior of stabilizing

  18. Constant strength fuel-fuel cell

    International Nuclear Information System (INIS)

    Vaseen, V.A.

    1980-01-01

    A fuel cell is an electrochemical apparatus composed of both a nonconsumable anode and cathode; and electrolyte, fuel oxidant and controls. This invention guarantees the constant transfer of hydrogen atoms and their respective electrons, thus a constant flow of power by submergence of the negative electrode in a constant strength hydrogen furnishing fuel; when said fuel is an aqueous absorbed hydrocarbon, such as and similar to ethanol or methnol. The objective is accomplished by recirculation of the liquid fuel, as depleted in the cell through specific type membranes which pass water molecules and reject the fuel molecules; thus concentrating them for recycle use

  19. IN-CORE FUEL MANAGEMENT: PWR Core Calculations Using MCRAC

    Science.gov (United States)

    PetroviĆ, B. G.

    1991-01-01

    The following sections are included: * INTRODUCTION * IN-CORE FUEL MANAGEMENT CALCULATIONS * In-Core Fuel Management * Methodological Problems of In-Core Fuel Management * In-Core Fuel Management Analytical Tools * PENN STATE FUEL MANAGEMENT PACKAGE * Penn State Fuel Management Package (PFMP) * Assembly Data Description (ADD) * Linking PSU-LEOPARD and MCRAC: An Example * MULTICYCLE REACTOR ANALYSIS CODE (MCRAC) * Main Features and Options of MCRAC code * Core geometry * Diffusion equations * 1.5-group model * Multicycle neutronic analysis * Multicycle cost analysis * Criticality search * Power-dependent xenon feedback calculations * Control rod and burnable absorber simulation * Search for LP with flat BOC power distribution * Artificial ADD option * Variable dimensioning technique * RBI version of MCRAC code * Programming changes in PC version * Fuel interchange option * MCRAC Input/Output * General input description * Sample input * Sample output * EXPERIENCE WITH MCRAC CODE * CONCLUSIONS * REFERENCES

  20. Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

    International Nuclear Information System (INIS)

    Boltax, A.; Biancheria, A.

    1977-01-01

    Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

  1. Chemical characterization of nuclear fuel materials

    International Nuclear Information System (INIS)

    Ramakumar, K.L.

    2011-01-01

    India is fabricating nuclear fuels for various types of reactors, for example, (U-Pu) MOX fuel of varying Pu content for boiling water reactors (BWRs), pressurized heavy water reactors (PHWRs), prototype fast breeder reactors (PFBRs), (U-Pu) carbide fuel fast breeder test reactor (FBTR), and U-based fuels for research reactors. Nuclear fuel being the heart of the reactor, its chemical and physical characterisation is an important component of this design. Both the fuel materials and finished fuel products are to be characterised for this purpose. Quality control (both chemical and physical) provides a means to ensure that the quality of the fabricated fuel conforms to the specifications for the fuel laid down by the fuel designer. Chemical specifications are worked out for the major and minor constituents which affect the fuel properties and hence its performance under conditions prevailing in an operating reactor. Each fuel batch has to be subjected to comprehensive chemical quality control for trace constituents, stoichiometry and isotopic composition. A number of advanced process and quality control steps are required to ensure the quality of the fuels. Further more, in the case of Pu-based fuels, it is necessary to extract maximum quality data by employing different evaluation techniques which would result in minimum scrap/waste generation of valuable plutonium. The task of quality control during fabrication of nuclear fuels of various types is both challenging and difficult. The underlying philosophy is total quality control of the fuel by proper mix of process and quality control steps at various stages of fuel manufacture starting from the feed materials. It is also desirable to adapt more than one analytical technique to increase the confidence and reliability of the quality data generated. This is all the most required when certified reference materials are not available. In addition, the adaptation of non-destructive techniques in the chemical quality

  2. MERRA Analytic Services

    Science.gov (United States)

    Schnase, J. L.; Duffy, D. Q.; McInerney, M. A.; Tamkin, G. S.; Thompson, J. H.; Gill, R.; Grieg, C. M.

    2012-12-01

    MERRA Analytic Services (MERRA/AS) is a cyberinfrastructure resource for developing and evaluating a new generation of climate data analysis capabilities. MERRA/AS supports OBS4MIP activities by reducing the time spent in the preparation of Modern Era Retrospective-Analysis for Research and Applications (MERRA) data used in data-model intercomparison. It also provides a testbed for experimental development of high-performance analytics. MERRA/AS is a cloud-based service built around the Virtual Climate Data Server (vCDS) technology that is currently used by the NASA Center for Climate Simulation (NCCS) to deliver Intergovernmental Panel on Climate Change (IPCC) data to the Earth System Grid Federation (ESGF). Crucial to its effectiveness, MERRA/AS's servers will use a workflow-generated realizable object capability to perform analyses over the MERRA data using the MapReduce approach to parallel storage-based computation. The results produced by these operations will be stored by the vCDS, which will also be able to host code sets for those who wish to explore the use of MapReduce for more advanced analytics. While the work described here will focus on the MERRA collection, these technologies can be used to publish other reanalysis, observational, and ancillary OBS4MIP data to ESGF and, importantly, offer an architectural approach to climate data services that can be generalized to applications and customers beyond the traditional climate research community. In this presentation, we describe our approach, experiences, lessons learned,and plans for the future.; (A) MERRA/AS software stack. (B) Example MERRA/AS interfaces.

  3. Renewable Fuel Standard Program

    Science.gov (United States)

    Information about regulations, developed by EPA, in collaboration with refiners, renewable fuel producers, and many other stakeholders, that ensure that transportation fuel sold in the United States contains a minimum volume of renewable fuel.

  4. Fuel Property Blend Model

    Energy Technology Data Exchange (ETDEWEB)

    Pitz, William J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mehl, Marco [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wagnon, Scott J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Zhang, Kuiwen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kukkadapu, Goutham [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Westbrook, Charles K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-01-12

    The object of this project is to develop chemical models and associated correlations to predict the blending behavior of bio-derived fuels when mixed with conventional fuels like gasoline and diesel fuels.

  5. Logistic Fuel Processor Development

    National Research Council Canada - National Science Library

    Salavani, Reza

    2004-01-01

    The Air Base Technologies Division of the Air Force Research Laboratory has developed a logistic fuel processor that removes the sulfur content of the fuel and in the process converts logistic fuel...

  6. Fuel pellet loading apparatus

    International Nuclear Information System (INIS)

    1980-01-01

    Apparatus is described for loading a predetermined amount of nuclear fuel pellets into nuclear fuel elements and particularly for the automatic loading of fuel pellets from within a sealed compartment. (author)

  7. Ecological effects of fuel cycle activities

    International Nuclear Information System (INIS)

    Barnthouse, L.; Cada, G.; Kroodsma, R.; Shriner, D.; Tolbert, V.; Turner, R.

    1994-01-01

    The purpose of this paper is to summarize the approach used to characterize ecological impacts of the coal fuel cycle. The same approach is used for many of the impacts in other fuel cycles as well. The principal analytical approach being used in the study is an accounting framework - that is, a series of matrices that map each phase of the fuel cycle to a suite of possible. emissions, each emission to a suite of impact categories, and each impact category to an external cost. This paper summarizes the ecological impacts of all phases of the coal fuel cycle, defines the ecological impact categories used in the study's 'accounting framework', and discusses alternative approaches to quantification. Externalities associated with CO 2 -induced global climate change are beyond the scope of this paper and are not discussed

  8. Analysis of fuel oxygenates in the environment

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, T.C.; Berg, M.; Haderlein, S.B. [Swiss Federal Inst. for Environmental Science and Technology (EAWAG) (Switzerland); Swiss Federal Inst. of Technology (ETH), Duebendorf (Switzerland); Duong, Hong-Anh [Vietnam National Univ.,Hanoi (Viet Nam). Center for Environmental Chemistry

    2001-07-01

    This paper presents an overview of currently available analytical methods for fuel oxygenates such as methyl tert-butyl ether and ethanol and highlights the advantages and disadvantages of the different methods. The occurrence of fuel oxygenates in water and air is explored, and sampling and enrichment of oxygenates in water are described covering water sampling, direct aqueous injection into a chromatographic column, headspace analysis, purge and trap enrichment, and solid phase microextraction. Methods for sampling and enrichment of oxygenates in air, separation of fuel oxygenates, preparation of standards and calibration, and detection using flame ionisation and photoionisation detection, mass spectrometry, atomic emission detection, and Fourier transform infrared spectroscopy are examined. Environmentally relevant physiochemical parameters of fuel oxygenates are tabulated, and injection and enrichment techniques for water analysis are compared.

  9. Advanced analytical techniques

    International Nuclear Information System (INIS)

    Mrochek, J.E.; Shumate, S.E.; Genung, R.K.; Bahner, C.T.; Lee, N.E.; Dinsmore, S.R.

    1976-01-01

    The development of several new analytical techniques for use in clinical diagnosis and biomedical research is reported. These include: high-resolution liquid chromatographic systems for the early detection of pathological molecular constituents in physiologic body fluids; gradient elution chromatography for the analysis of protein-bound carbohydrates in blood serum samples, with emphasis on changes in sera from breast cancer patients; electrophoretic separation techniques coupled with staining of specific proteins in cellular isoenzymes for the monitoring of genetic mutations and abnormal molecular constituents in blood samples; and the development of a centrifugal elution chromatographic technique for the assay of specific proteins and immunoglobulins in human blood serum samples

  10. Local analytic geometry

    CERN Document Server

    Abhyankar, Shreeram Shankar

    1964-01-01

    This book provides, for use in a graduate course or for self-study by graduate students, a well-motivated treatment of several topics, especially the following: (1) algebraic treatment of several complex variables; (2) geometric approach to algebraic geometry via analytic sets; (3) survey of local algebra; (4) survey of sheaf theory. The book has been written in the spirit of Weierstrass. Power series play the dominant role. The treatment, being algebraic, is not restricted to complex numbers, but remains valid over any complete-valued field. This makes it applicable to situations arising from

  11. Analytic aspects of convexity

    CERN Document Server

    Colesanti, Andrea; Gronchi, Paolo

    2018-01-01

    This book presents the proceedings of the international conference Analytic Aspects in Convexity, which was held in Rome in October 2016. It offers a collection of selected articles, written by some of the world’s leading experts in the field of Convex Geometry, on recent developments in this area: theory of valuations; geometric inequalities; affine geometry; and curvature measures. The book will be of interest to a broad readership, from those involved in Convex Geometry, to those focusing on Functional Analysis, Harmonic Analysis, Differential Geometry, or PDEs. The book is a addressed to PhD students and researchers, interested in Convex Geometry and its links to analysis.

  12. Analytical elements of mechanics

    CERN Document Server

    Kane, Thomas R

    2013-01-01

    Analytical Elements of Mechanics, Volume 1, is the first of two volumes intended for use in courses in classical mechanics. The books aim to provide students and teachers with a text consistent in content and format with the author's ideas regarding the subject matter and teaching of mechanics, and to disseminate these ideas. The book opens with a detailed exposition of vector algebra, and no prior knowledge of this subject is required. This is followed by a chapter on the topic of mass centers, which is presented as a logical extension of concepts introduced in connection with centroids. A

  13. Analytical chemistry in space

    CERN Document Server

    Wainerdi, Richard E

    1970-01-01

    Analytical Chemistry in Space presents an analysis of the chemical constitution of space, particularly the particles in the solar wind, of the planetary atmospheres, and the surfaces of the moon and planets. Topics range from space engineering considerations to solar system atmospheres and recovered extraterrestrial materials. Mass spectroscopy in space exploration is also discussed, along with lunar and planetary surface analysis using neutron inelastic scattering. This book is comprised of seven chapters and opens with a discussion on the possibilities for exploration of the solar system by

  14. Fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Enomoto, Hirofumi.

    1989-05-22

    This invention aims to maintain a long-term operation with stable cell output characteristics by uniformly supplying an electrolyte from the reserver to the matrix layer over the entire matrix layer, and further to prevent the excessive wetting of the catalyst layer by smoothly absorbing the volume change of the electrolyte, caused by the repeated stop/start-up of the fuel cell, within the reserver system. For this purpose, in this invention, an electrolyte transport layer, which connects with an electrolyte reservor formed at the electrode end, is partly formed between the electrode material and the catalyst layer; a catalyst layer, which faces the electrolyte transport layer, has through-holes, which connect to the matrix, dispersely distributed. The electrolyte-transport layer is a thin sheet of a hydrophilic fibers which are non-wovens of such fibers as carbon, silicon carbide, silicon nitride or inorganic oxides. 11 figs.

  15. Quality Indicators for Learning Analytics

    Science.gov (United States)

    Scheffel, Maren; Drachsler, Hendrik; Stoyanov, Slavi; Specht, Marcus

    2014-01-01

    This article proposes a framework of quality indicators for learning analytics that aims to standardise the evaluation of learning analytics tools and to provide a mean to capture evidence for the impact of learning analytics on educational practices in a standardised manner. The criteria of the framework and its quality indicators are based on…

  16. Learning Analytics: Readiness and Rewards

    Science.gov (United States)

    Friesen, Norm

    2013-01-01

    This position paper introduces the relatively new field of learning analytics, first by considering the relevant meanings of both "learning" and "analytics," and then by looking at two main levels at which learning analytics can be or has been implemented in educational organizations. Although integrated turnkey systems or…

  17. Division of Analytical Chemistry, 1998

    DEFF Research Database (Denmark)

    Hansen, Elo Harald

    1999-01-01

    The article recounts the 1998 activities of the Division of Analytical Chemistry (DAC- formerly the Working Party on Analytical Chemistry, WPAC), which body is a division of the Federation of European Chemical Societies (FECS). Elo Harald Hansen is the Danish delegate, representing The Danish...... Chemical Society/The Society for Analytical Chemistry....

  18. Fuel storage

    International Nuclear Information System (INIS)

    Palacios, C.; Alvarez-Miranda, A.

    2009-01-01

    ENSA is a well known manufacturer of multi-system primary components for the nuclear industry and is totally prepared to satisfy future market requirements in this industry. At the same time that ENSA has been gaining a reputation world wider for the supply of primary components, has been strengthening its commitment and experience in supplying spent fuel components, either pool racks or storage and transportation casks, and offers not only fabrication but also design capabilities for its products. ENSA has supplied Spent Fuel Pool Racks, in spain, Finland, Taiwan, Korea, China, and currently it is in the process of licensing its own rack design in the United States of America for the ESBWR along with Ge-Hitachi. ENSA has supplied racks for 20 pools and 22 different reactors and it has also manufactured racks under all available technologies and developed a design known as Interlock Cell Matrix whose main features are outlined in this article. Another ENSA achievement in rack technology is the use of remote control for re-racking activities instead of using divers, which improves the ALARA requirements. Regarding casks for storage and transportation, ENSA also has al leading worldwide position, with exports prevailing over the Spanish market where ENSA has supplied 16 storage and transportation casks to the Spanish nuclear power Trillo. In some cases, ENSA acts as subcontractor for other clients. Foreign markets are still a major challenge for ENSA. ENSA-is well known for its manufacturing capabilities in the nuclear industry, but has been always involved in design activities through its engineering division, which carries out different tasks: components Design; Tooling Design; Engineering and Documentation; Project Engineering; Calculations, Design and Development Engineering. (Author)

  19. Nuclear fuel replacement device

    International Nuclear Information System (INIS)

    Ritz, W.C.; Robey, R.M.; Wett, J.F.

    1984-01-01

    A fuel handling arrangement for a liquid metal cooled nuclear reactor having a single rotating plug eccentric to the fuel core and a fuel handling machine radially movable along a slot in the plug with a transfer station disposed outside the fuel core but covered by the eccentric plug and within range of movement of said fuel handling machine to permit transfer of fuel assemblies between the core and the transfer station. (author)

  20. CANDU fuel performance

    International Nuclear Information System (INIS)

    Ivanoff, N.V.; Bazeley, E.G.; Hastings, I.J.

    1982-01-01

    CANDU fuel has operated successfully in Ontario Hydro's power reactors since 1962. In the 19 years of experience, about 99.9% of all fuel bundles have performed as designed. Most defects occurred before 1979 and subsequent changes in fuel design, fuel management, reactor control, and manufacturing quality control have reduced the current defect rate to near zero. Loss of power production due to defective fuel has been negligible. The outstanding performance continues while maintaining a low unit energy cost for fuel