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Sample records for producing fissile fuels

  1. Electronuclear fissile fuel production. Linear accelerator fuel regenerator and producer LAFR and LAFP

    International Nuclear Information System (INIS)

    Steinberg, M.; Powell, J.R.; Takahashi, H.; Grand, P.; Kouts, H.J.C.

    1978-04-01

    A linear accelerator fuel generator is proposed to enrich naturally occurring fertile U-238 or thorium 232 with fissile Pu-239 or U-233 for use in LWR power reactors. High energy proton beams in the range of 1 to 3 GeV energy are made to impinge on a centrally located dispersed liquid lead target producing spallation neutrons which are then absorbed by a surrounding assembly of fabricated LWR fuel elements. The accelerator-target design is reviewed and a typical fuel cycle system and economic analysis is presented. One 300 MW beam (300 ma-1 GeV) linear accelerator fuel regenerator can provide fuel for 3 to 1000 MW(e) LWR power reactors over its 30-year lifetime. There is a significant saving in natural uranium requirement which is a factor of 4.5 over the present LWR fuel requirement assuming the restraint of no fissile fuel recovery by reprocessing. A modest increase (approximately 10%) in fuel cycle and power production cost is incurred over the present LWR fuel cycle cost. The linear accelerator fuel regenerator and producer assures a long-term supply of fuel for the LWR power economy even with the restraint of the non-proliferation policy of no reprocessing. It can also supply hot-denatured thorium U-233 fuel operating in a secured reprocessing fuel center

  2. Fissile Content Assay of Spent Fuel Using LSDS System

    International Nuclear Information System (INIS)

    Jeon, Ju Young; Lee, Yong Deok; Park, Chang Je

    2016-01-01

    About 1.5 % fissile materials still exist in the spent fuel. Therefore, for reutilization of fissile materials in spent fuel at SFR, resource material is produced through the pyro process. Fissile material contents in the resource material must be analyzed before fabricating SFR fuel for reactor safety and economics. The new technology for an isotopic fissile material content assay is under development at KAERI using a lead slowing down spectrometer (LSDS). LSDS is very sensitive to distinguish fission signals from each fissile isotope in spent and recycled fuel. In an assay of fissile content of spent fuel and recycled fuel, an intense radiation background gives limits the direct analysis of fissile materials. However, LSDS is not influenced by such a radiation background in a fissile assay. Based on the decided LSDS geometry set up, a self shielding parameter was calculated at the fuel assay zone by introducing spent fuel or pyro produced nuclear material. When nuclear material is inserted into the assay area, the spent fuel assembly or pyro recycled fuel material perturbs the spatial distribution of slowing down neutrons in lead and the prompt fast fission neutrons produced by fissile materials are also perturbed. The self shielding factor is interpreted as how much of the absorption is created inside the fuel area when it is in the lead. The self shielding effect provides a non-linear property in the isotopic fissile assay. When the self shielding is severe, the assay system becomes more complex and needs a special parameter to treat this non linear effect. Additionally, an assay of isotopic fissile content will contribute to an accuracy improvement of the burn-up code and increase the transparency and credibility for spent fuel storage and usage, as internationally increasing demand. The fissile contents result came out almost exactly with relative error ∼ 2% in case of Pu239, Pu241 for two different plutonium contents. In this study, meaningful results were

  3. Determining fissile content of nuclear fuel elements

    International Nuclear Information System (INIS)

    Arya, S.P.; Grossman, L.N.; Schoenig, F.C.

    1980-01-01

    This invention relates to the determination of the fissile fuel content of fuel for nuclear reactors. A nondestructive method is described for determining rapidly, accurately and simultaneously the fissile content, enrichment and location of fuel material which may also contain amounts of burnable poison, by detecting the γ-rays emitted from the fuel material due to natural radioactive decay. (U.K.)

  4. Rational non-Pu fuel-cycle composed simple power-stations and fissile producers

    International Nuclear Information System (INIS)

    Furukawa, K.; Mitachi, K.; Kato, Y.; Lecocq, A.

    1989-01-01

    In the next century, the fission breeder concept would not be practical for solving global energy problems. As a measure, a new rational is needed. In this paper the breeding fuel cycle system is proposed to establish the improvement in issues of safety, power-size flexibility, anti-terrorism and radio-waste, economy, etc. securing the simple operation, maintenance and chemical processing

  5. Irradiation performance of HTGR recycle fissile fuel

    International Nuclear Information System (INIS)

    Homan, F.J.; Long, E.L. Jr.

    1976-08-01

    The irradiation performance of candidate HTGR recycle fissile fuel under accelerated testing conditions is reviewed. Failure modes for coated-particle fuels are described, and the performance of candidate recycle fissile fuels is discussed in terms of these failure modes. The bases on which UO 2 and (Th,U)O 2 were rejected as candidate recycle fissile fuels are outlined, along with the bases on which the weak-acid resin (WAR)-derived fissile fuel was selected as the reference recycle kernel. Comparisons are made relative to the irradiation behavior of WAR-derived fuels of varying stoichiometry and conclusions are drawn about the optimum stoichiometry and the range of acceptable values. Plans for future testing in support of specification development, confirmation of the results of accelerated testing by real-time experiments, and improvement in fuel performance and reliability are described

  6. Design of LSDS for Isotopic Fissile Assay in Spent Fuel

    International Nuclear Information System (INIS)

    Lee, Yongdeok; Park, Changje; Kim, Hodong; Song, Kee Chan

    2013-01-01

    A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded

  7. DESIGN OF LSDS FOR ISOTOPIC FISSILE ASSAY IN SPENT FUEL

    Directory of Open Access Journals (Sweden)

    YONGDEOK LEE

    2013-12-01

    Full Text Available A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI, the system involves a Sodium Fast Reactor (SFR linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS. The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.

  8. LSDS Development for Isotopic Fissile Assay in Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Lee, Jung Won; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-07-01

    As an option to reduce a spent fuel and reuse an existing fissile material in spent fuel, sodium fast reactor SFR program linked with pyro-processing is under development in KAERI. A uranium-TRU mixture through a pyro-process is used to fabricate SFR fuel. An assay of isotopic fissile content plays an important role in an optimum design of storage site and reuse of fissile materials of spent fuel. Lead slowing down spectrometer LSDS is being developed in KAERI to analyze isotopic fissile material content. LSDS has several features: direct fissile assay, near real time fissile assay, no influence from radiation background, fissile isotopic assay and applicable to spent fuel and recycled fuel. Based on the designed geometry, neutron energy resolution was investigated. The neutron energy spectrum was analyzed as well. Spent fuel emits large number of neutrons by spontaneous fission. Neutron generator must overcome the neutron background to get the pure fission signals from fissile materials. Neutron generator is planned to have compact system with one section electron linac which is easy maintenance, less cost and high neutron yield. The LSD has the power to resolve the fission characteristics from each fissile material. This feature can analyze the content of isotopic fissile. From 1keV to 0.1eV energy range, the energy resolution is enough to get the individual fissile fission signatures. The dominant fission signature is shown below 1eV for each fissile isotope. The neutron generation system with target was designed to get fission signals by fissile materials. The system was decided to overcome neutron backgrounds and to get good counting statistics. Finally, an accurate fissile material content will contribute to safety of spent fuel reuse in future nuclear energy system and optimum design of spent fuel storage site. Additionally, an accurate fissile material content will increase international transparence and credibility for the reuse of PWR spent fuel.

  9. LSDS Development for Isotopic Fissile Assay in Spent Fuel

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Lee, Jung Won; Song, Kee Chan

    2011-01-01

    As an option to reduce a spent fuel and reuse an existing fissile material in spent fuel, sodium fast reactor SFR program linked with pyro-processing is under development in KAERI. A uranium-TRU mixture through a pyro-process is used to fabricate SFR fuel. An assay of isotopic fissile content plays an important role in an optimum design of storage site and reuse of fissile materials of spent fuel. Lead slowing down spectrometer LSDS is being developed in KAERI to analyze isotopic fissile material content. LSDS has several features: direct fissile assay, near real time fissile assay, no influence from radiation background, fissile isotopic assay and applicable to spent fuel and recycled fuel. Based on the designed geometry, neutron energy resolution was investigated. The neutron energy spectrum was analyzed as well. Spent fuel emits large number of neutrons by spontaneous fission. Neutron generator must overcome the neutron background to get the pure fission signals from fissile materials. Neutron generator is planned to have compact system with one section electron linac which is easy maintenance, less cost and high neutron yield. The LSD has the power to resolve the fission characteristics from each fissile material. This feature can analyze the content of isotopic fissile. From 1keV to 0.1eV energy range, the energy resolution is enough to get the individual fissile fission signatures. The dominant fission signature is shown below 1eV for each fissile isotope. The neutron generation system with target was designed to get fission signals by fissile materials. The system was decided to overcome neutron backgrounds and to get good counting statistics. Finally, an accurate fissile material content will contribute to safety of spent fuel reuse in future nuclear energy system and optimum design of spent fuel storage site. Additionally, an accurate fissile material content will increase international transparence and credibility for the reuse of PWR spent fuel

  10. Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1997-01-01

    A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs

  11. New Technology For Fissile Assay In Spent Fuel Using LSDS

    International Nuclear Information System (INIS)

    Lee, Yongdeok; Park, Changje; Park, Geunil; Lee, Jungwon; Song, Keechan

    2012-01-01

    The principle of LSDS is very simple. The interrogated neutron induces energy dependent characteristic fission from fissile materials in spent fuel. The fission threshold detector screens the prompt fast fission neutrons from background and fissionable materials. However, intense source neutron is necessary to overcome radiation background. The detected signals have a direct relationship to the content of each fissile material. The isotopic fissile assay using LSDS is applicable for optimum design of spent fuel storage and management, quality assurance of recycled nuclear material, maximization of burnup credit. Another important application is verity burnup code and provide correction factor for improving the fissile material content, fission product correction factor for improving the fissile material content, fission product content and theoretical burnup. Additionally, the isotopic fissile content assay will increase the transparence and credibility for spent fuel storage and its re-utilization, as internationally demanded

  12. Fissile fuel doubling time characteristics for reactor lifetime fuel logistics

    International Nuclear Information System (INIS)

    Heindler, M.; Harms, A.A.

    1978-01-01

    The establishment of nuclear fuel requirements and their efficient utilization requires a detailed knowledge of some aspects of fuel dynamics and processing during the reactor lifetime. It is shown here that the use of the fuel stockpile inventory concept can serve effectively for this fuel management purpose. The temporal variation of the fissile fuel doubling time as well as nonequilibrium core conditions are among the characteristics which thus become more evident. These characteristics - rather than a single figure-of-merit - clearly provide an improved description of the expansion capacity and/or fuel requirements of a nuclear reactor energy system

  13. Status of LSDS Development for Isotopic Fissile Assay in Used Fuel

    International Nuclear Information System (INIS)

    Lee, Y.D.; Ahn, S.; Kim, H.-D.; Song, K.C.; Park, C.J.

    2015-01-01

    Because of the large amount accumulation of spent fuel, a research to solve the spent fuel problem is actively performed in Korea. One option is to develop the SFR linked with the pyro process to reuse the existing fissile materials in spent fuel. Therefore, an accurate isotopic fissile content assay becomes a key factor in the reuse of fissile material for safety and safeguards purpose. There are several commercial non-destructive technologies for nuclear material assay. However, technology for direct isotopic fissile content assay in spent fuel is not developed yet. Internationally, a verification of special nuclear material in spent fuel, mainly U-235, Pu239, Pu241, is very important for the safeguards objective. These fissile materials can be misused for nuclear weapon purpose, not for peaceful use. As a future nuclear system is developed,, improved safeguards technology must be developed for an approval of fissile materials. A direct measurement of fissile materials is very important to provide a continuous of knowledge on nuclear materials. LSDS (Lead Slowing Down Spectrometer) has an advantage to assay an isotopic fissile content directly, without any help of burnup code and history. LSDS system is under development in KAERI (Korea Atomic Energy Research Institute) for spent fuel and recycled fuel. A linear assay model was setup for U235, Pu239 and Pu241. The dominant individual fission characteristic is appeared between 0.1 eV and 1 keV range. An electron linear accelerator for compact and low cost is under development to produce high source neutron effectively and efficiently. The LSDS is also applicable for optimum design of spent fuel storage and management. The advanced fissile assay technology will contribute to increase the transparency and credibility internationally on a reuse of fissile materials in future nuclear energy system development. (author)

  14. Fissile fuel dynamics of breeder/converter reactors

    International Nuclear Information System (INIS)

    Harms, A.A.

    1978-01-01

    The long-term fissile fuel dynamics for a hierarchy of fission reactors covering the range from pure-burners to super-breeders is examined. It is found that the breeding gains of the core and blanket can be used to identify several distinct fissile fuel histories and elucidate the importance of fuel cycle characteristics such as the time dependence of the fissile fuel doubling time. On this basis, a self-sufficient fission reactor is introduced and its determining characteristics are identified. (author)

  15. Quantitative Fissile Assay In Used Fuel Using LSDS System

    Science.gov (United States)

    Lee, YongDeok; Jeon, Ju Young; Park, Chang-Je

    2017-09-01

    A quantitative assay of isotopic fissile materials (U235, Pu239, Pu241) was done at Korea Atomic Energy Research Institute (KAERI), using lead slowing down spectrometer (LSDS). The optimum design of LSDS was performed based on economics, easy maintenance and assay effectiveness. LSDS system consists of spectrometer, neutron source, detection and control. LSDS system induces fissile fission and fast neutrons are collected at fission chamber. The detected signal has a direct relation to the mass of existing fissile isotopes. Many current commercial assay technologies have a limitation in direct application on isotopic fissile assay of spent fuel, except chemical analysis. In the designed system, the fissile assay model was setup and the correction factor for self-shield was obtained. The isotopic fissile content assay was performed by changing the content of Pu239. Based on the fuel rod, the isotopic content was consistent with 2% uncertainty for Pu239. By applying the covering (neutron absorber), the effective shielding was obtained and the activation was calculated on the target. From the assay evaluation, LSDS technique is very powerful and direct to analyze the isotopic fissile content. LSDS is applicable for nuclear fuel cycle and spent fuel management for safety and economics. Additionally, an accurate fissile content will contribute to the international transparency and credibility on spent fuel.

  16. Isotopic fissile assay of spent fuel in a lead slowing-down spectrometer system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Deok; Jeon, Ju Young [Dept. of Fuel Cycle Technology, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Chang Je [Dept. of Nuclear Engineering, Sejong University, Seoul (Korea, Republic of)

    2017-04-15

    A lead slowing-down spectrometer (LSDS) system is under development to analyze isotopic fissile content that is applicable to spent fuel and recycled material. The source neutron mechanism for efficient and effective generation was also determined. The source neutron interacts with a lead medium and produces continuous neutron energy, and this energy generates dominant fission at each fissile, below the unresolved resonance region. From the relationship between the induced fissile fission and the fast fission neutron detection, a mathematical assay model for an isotopic fissile material was set up. The assay model can be expanded for all fissile materials. The correction factor for self-shielding was defined in the fuel assay area. The corrected fission signature provides well-defined fission properties with an increase in the fissile content. The assay procedure was also established. The assay energy range is very important to take into account the prominent fission structure of each fissile material. Fission detection occurred according to the change of the Pu239 weight percent (wt%), but the content of U235 and Pu241 was fixed at 1 wt%. The assay result was obtained with 2∼3% uncertainty for Pu239, depending on the amount of Pu239 in the fuel. The results show that LSDS is a very powerful technique to assay the isotopic fissile content in spent fuel and recycled materials for the reuse of fissile materials. Additionally, a LSDS is applicable during the optimum design of spent fuel storage facilities and their management. The isotopic fissile content assay will increase the transparency and credibility of spent fuel storage.

  17. Development of a fissile particle for HTGR fuel recycle

    International Nuclear Information System (INIS)

    Homan, F.J.; Long, E.L. Jr.; Lindemer, T.B.; Beatty, R.L.; Tiegs, T.N.

    1976-12-01

    Recycle fissile fuel particles for high-temperature gas-cooled reactors (HTGRs) have been under development since the mid-1960s. Irradiation performance on early UO 2 and Th 0 . 8 U 0 . 2 O 2 kernels is described in this report, and the performance limitations associated with the dense oxide kernels are presented. The development of the new reference fuel kernel, the weak-acid-resin-derived (WAR) UO 2 --UC 2 , is discussed in detail, including an extensive section on the irradiation performance of this fuel in HFIR removable beryllium capsules HRB-7 through -10. The conclusion is reached that the irradiation performance of the WAR fissile fuel kernel is better than that of any coated particle fuel yet tested. Further, the present fissile kernel is adequate for steam cycle HTGRs as well as for many advanced applications such as gas turbine and process heat HTGRs

  18. Spectrum analysis in lead spectrometer for isotopic fissile assay in used fuel

    International Nuclear Information System (INIS)

    Lee, Y.D.; Park, C.J.; Kim, H.D.; Song, K.C.

    2014-01-01

    The LSDS system is under development for analyzing isotopic fissile content applicable in a hot cell for the pyro process. The fuel assay area and nuclear material composition were selected for simulation. The source mechanism for efficient neutron generation was also determined. A neutron is produced at the Ta target by hitting it from accelerated electron. The parameters for an electron accelerator are being researched for cost effectiveness, easy maintenance, and compact size. The basic principle of LSDS is that isotopic fissile has its own fission structure below the unresolved resonance region. The source neutron interacts with a lead medium and produces continuous neutron energy, which generates dominant fission at each fissile. Therefore, a spectrum analysis is very important at a lead medium and fuel area for system working. The energy spectrum with respect to slowing down energy and the energy resolution were investigated in lead. A spectrum analysis was done by the existence of surrounding detectors. In particular, high resonance energy was considered. The spectrum was well organized at each slowing down energy and the energy resolution was acceptable to distinguish isotopic fissile fissions. Additionally, LSDS is applicable for the optimum design of spent fuel storage and management.The isotopic fissile content assay will increase the transparency and credibility for spent fuel storage and its re-utilization, as demanded internationally. (author)

  19. The burnable poisons utilization for fissile enriched CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Serghiuta, D; Nainer, O [Team 3 Solutions, Don Mills, ON (Canada)

    1996-12-31

    Utilization of burnable poison for the fissile enriched fueled CANDU 6 Mk1 core is investigated. The main incentives for this analysis are the reduction of void reactivity effects, the maximization of the fissile content of fresh fuel bundles, and the achievement of better power shape control, in order to preserve the power envelope of the standard 37 rod fuel bundle. The latter allows also the preservation of construction parameters of the standard core (for example: number and location of reactivity devices). It also permits the use of regular shift fueling schemes. The paper makes analyses of MOX weapons-grade plutonium and 1.2% SEU fueled CANDU 6 Mk 1 cores. (author). 6 refs., 4 tabs., 10 figs.

  20. Development for fissile assay in recycled fuel using lead slowing down spectrometer

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Je Park, C.; Kim, Ho-Dong; Song, Kee Chan

    2013-01-01

    A future nuclear energy system is under development to turn spent fuels produced by PWRs into fuels for a SFR (Sodium Fast Reactor) through the pyrochemical process. The knowledge of the isotopic fissile content of the new fuel is very important for fuel safety. A lead slowing down spectrometer (LSDS) is under development to analyze the fissile material content (Pu 239 , Pu 241 and U 235 ) of the fuel. The LSDS requires a neutron source, the neutrons will be slowed down through their passage in a lead medium and will finally enter the fuel and will induce fission reactions that will be analysed and the isotopic content of the fuel will be then determined. The issue is that the spent fuel emits intense gamma rays and neutrons by spontaneous fission. The threshold fission detector screens the prompt fast fission neutrons and as a result the LSDS is not influenced by the high level radiation background. The energy resolution of LSDS is good in the range 0.1 eV to 1 keV. It is also the range in which the fission reaction is the most discriminating for the considered fissile isotopes. An electron accelerator has been chosen to produce neutrons with an adequate target through (e - ,γ)(γ,n) reactions

  1. Fusion-Fission Hybrid for Fissile Fuel Production without Processing

    Energy Technology Data Exchange (ETDEWEB)

    Fratoni, M; Moir, R W; Kramer, K J; Latkowski, J F; Meier, W R; Powers, J J

    2012-01-02

    Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of {sup 232}Th and fission of {sup 233}U in situ without reprocessing or 'closed' cycles based on irradiation of {sup 232}Th followed by reprocessing, and recycling of {sup 233}U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE1) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in

  2. Valorization of the energy potential of fossil and fissile fuels for heat production: dual-purpose power plants and heat-producing nuclear reactors

    International Nuclear Information System (INIS)

    Lavite, Michel.

    1975-07-01

    The heat market is analyzed briefly within the French context: present structures and characteristics of the market, current means of heat production, predictable trend of the demand. The possible applications of nuclear energy to heat production, through the agency of combined electricity-steam stations or heat-producing stations, are then examined. Nuclear solutions are compared with others from the technico-economic and ecological wiewpoints and an estimate fo their respective impacts on the energy balance is attempted [fr

  3. Economic evaluation of fissile fuel production using resistive magnet tokamaks

    International Nuclear Information System (INIS)

    Doyle, J.C. Jr.

    1985-06-01

    The application of resistive magnet tokamaks to fissile fuel production has been studied. Resistive magnets offer potential advantages over superconducting magnets in terms of robustness, less technology development required and possibility of demountable joints. Optimization studies within conservatively specified constraints for a compact machine result in a major radius of 3.81 m and 618 MW fusion power and a blanket space envelope of 0.35 m inboard and 0.75 m outboard. This machine is called the Resistive magnet Tokamak Fusion Breeder (RTFB). A computer code was developed to estimate the cost of the resistive magnet tokamak breeder. This code scales from STARFIRE values where appropriate and calculates costs of other systems directly. The estimated cost of the RTFB is $3.01 B in 1984 dollars. The cost of electricity on the same basis as STARFIRE is 42.4 mills/kWhre vs 44.9 mills/kWhre for STARFIRE (this does not include the fuel value or fuel cycle costs for the RTFB). The breakeven cost of U 3 O 8 is $150/lb when compared to a PWR on the once through uranium fuel cycle with no inflation and escalation. On the same basis, the breakeven cost for superconducting tokamak and tandem mirror fusion breeders is $160/lb and $175/lb. Thus, the RTFB appears to be competitive in breakeven U 3 O 8 cost with superconducting magnet fusion breeders and offers the potential advantages of resistive magnet technology

  4. Fissile fuel assembly for a sub-moderated nuclear reactor

    International Nuclear Information System (INIS)

    Millot, J.P.; Dejeux, Pol.; Alibran, Patrice.

    1983-01-01

    Each of the core assemblies is composed of a prismatic case made of a neutron absorbing material, inside which very long rods containing the fissile material are arranged parallel to the height of the case and according to a regular network in the straight sections of the case. At least one piece in a fertile material exposed to the neutrons emitted by the fissile material of the assembly is arranged on each one of the side faces of the case. The invention applies in particular to sub-moderated reactors, cooled and moderated by pressurized water [fr

  5. Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

    International Nuclear Information System (INIS)

    Ozdemir, Levent; Acar, Banu Bulut; Zabunoglu, Okan H.

    2011-01-01

    When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of 239 Pu and 241 Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

  6. Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

    International Nuclear Information System (INIS)

    Ipek, O.

    1997-01-01

    The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

  7. Experimental verification of neutron emission method for measuring of fissile material content in spent fuel

    International Nuclear Information System (INIS)

    Abou-Zaid, A.A.; Pytel, K.

    1999-01-01

    A non-destructive method of measurement of fissile nuclides content remained in spent fuel from research reactor is presented. The method, called the neutron emission one, is based on counting of fission neutrons emitted from fissile isotopes: 235 U, 239 Pu, 241 Pu. Fissions are induced mainly by neutrons supplied by the external neutron source. Another effects contribute also to the measured neutron population, e. g. source neutrons from penetrating the fuel without being captured and scattered, neutrons (α,n) reactions and from spontaneous fissions of actinides. Complexity of phenomena occurring within the measurement facility required the detailed numerical simulation and experimental studies prior design of ultimate measurement stand. In the previous paper, the results of Monte Carlo simulation on optimisation of measuring stand for neutron emission method were presented. On the basis of those results, the experimental stand for Maria reactor fuel investigation has been designed and manufactured. The present paper, being the continuation of previous one, contains the description of experimental facility and the results of measurements for the fresh fuel (without burnup) and the fuel mock-up (without fissile materials). Although some discrepancies were found between Monte Carlo and experimental results, the main conclusions concerning the optimal geometry of measuring facility have been confirmed. (author)

  8. Comparison of thorium-based fuels with different fissile components in existing BWRs

    International Nuclear Information System (INIS)

    Bjoerk, Klara Insulander; Fhager, Valentin; Demaziere, Christophe

    2009-01-01

    Three different types of thorium based BWR fuel have been developed, in each of which thorium was combined with a different fissile component, the three components being reactor grade plutonium, uranium enriched to 20% in uranium 235 and pure uranium 233. A BWR nuclear bundle design, based on the geometrical fuel assembly design GE14, was developed for each of these fissile components. The properties and performance of the corresponding fuel assemblies were investigated via full core calculations carried out for an existing BWR and compared with the ones of an ordinary Low Enriched Uranium (LEU) fuel, which was developed for reference. The fuel assemblies and cores were designed to meet existing fuel design criteria, and were then analyzed with regards to reactivity coefficients, delayed neutron fractions, control rod worths and shutdown margins. The results show that all three alternatives seem to be feasible, although some difficulties remain with complying with the thermal limits, and with the moderator temperature and coolant void coefficients of the U-233 containing fuel being positive under some circumstances. (author)

  9. Harmonisation of criticality assessments of packages for the transport of fissile nuclear fuel cycle materials

    International Nuclear Information System (INIS)

    Farrington, L.

    2004-01-01

    The transport of fissile nuclear fuel cycle materials is an international business, and for international shipments the regulations require a package to be certified by each country through or into which the consignment is to be transported. This raises a number of harmonisation issues, which have an important bearing on transport activities. National authorities carry out independent reviews of the criticality safety of packages containing fissile materials but the underlying assumptions used in the calculations can differ, and the outcome is that implementation of the regulations is not uniform. A single design may require multiple criticality analyses to obtain base approval and foreign validations. When several competent authorities are involved, the approval and validation process of package design can often become a time-consuming, expensive and unpredictably lengthy process that can have a significant detrimental effect upon the businesses involved. The characteristics of the fissile nuclear fuel cycle materials transported by the various countries have much in common and so have the designs of the packages to contain them. A greater degree of standardisation should allow criticality safety to be assessed consistently and efficiently with benefits for the nuclear transport industry and the regulatory bodies. (author)

  10. Harmonisation of criticality assessments of packages for the transport of fissile nuclear fuel cycle materials

    International Nuclear Information System (INIS)

    Farrington, L.

    2004-01-01

    The transport of fissile nuclear fuel cycle materials is an international business and for international shipments the regulations require a package to be certified by each country through or into which the consignment is to be transported. This raises a number of harmonisation issues, which have an important bearing on transport activities. National authorities carry out independent reviews of criticality safety of packages containing fissile materials but the underlying assumptions used in the calculations can differ, and the outcome is that implementation of the regulations is not uniform. A single design may require multiple criticality analyses to obtain base approval and foreign validations. When several Competent Authorities are involved, the approval and validation process of package design can often become time consuming, expensive and an unpredictably lengthy process that can have a significant detrimental effect upon the businesses involved. The characteristics of the fissile nuclear fuel cycle materials transported by the various countries have much in common and so have the designs of the packages to contain them. A greater degree of standardisation should allow criticality safety to be assessed consistently and efficiently with benefits for the nuclear transport industry and the regulatory bodies

  11. Effect of fissile isotope burnup on criticality safety for stored disintegrated fuel rods

    International Nuclear Information System (INIS)

    Heaberlin, S.W.; Selby, G.P.

    1978-09-01

    If the fuel rods were to disintegrate and water added, a criticality could occur in a 13-in. PWR canister with fresh fuel enriched to 3.5 wt % 235 U. The question is, ''If credit could be taken for burnup, could this indicate a subcritical condition.'' In attempting to answer this question, a series of calculations were performed. A set of isotopic concentrations were generated for 5,000, 10,000, 15,000, and 20,000 MWD/MTU burnup levels. Four reflector materials, water, concrete and two types of soil, were considered. Results indicate that allowing credit for fissile isotope burnup does not completely remove the concern for criticality safety in the event of rod disintegration. Reactivities which are ''subcritical'' (k/sub eff/ = 0.95) would not occur for three of the four reflector materials at even the 20,000 MWD/MTU burnup level in the 13-in. canister. The water reflected canister would achieve the k/sub eff/ = 0.95 level near 18,000 MWD/MTU. A smaller canister could be postulated. If a quarter inch gap is allowed, a Westinghouse 17 x 17 PWR assembly requires a 12 1 / 4 inch diameter canister. For such a canister with water reflection the ''subcritical'' (k/sub eff/ = 0.95) level would be reached near 15,000 MWD/MTU. The soil reflected canisters would reach this level between 18,000 and 19,000 MWD/MTU. Considering the difficulties in taking credit for burnup, such modest gains in apparent safety are not encouraging. This situation might be improved, however, if credit were also taken for neutron absorption by fission product poisons produced during burnup. It is strongly recommended that other approaches to a solution of the criticality safety problem be considered

  12. Physics concept on the constellation type fissile fuels and its application to the prospective Th-232U Reactor

    International Nuclear Information System (INIS)

    Zhang, Jiahua

    1994-01-01

    In contrast with the conventional nuclear reactor which usually fuelled with on single fissile nuclide, a constellation type fissile fuels reactor consists of a parent nuclide such as 232 Th or 238 U and its whole family of neutron generated daughter nuclides. All of them are regarded as fissile fuels but of quite different fission ability. The concentration of each daughter nuclide is determined by its saturate concentration ratio with the parent nuclide. In such fuel system, the whole fuel consumed by neutron reaction almost completely results in fission products. In this article, some properties of such fuel system, determination of the saturate concentration of each daughter nuclide and applicability to Th- 233 U fueled reactor will be discussed. 3 refs., 1 tab., 2 figs

  13. Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor

    International Nuclear Information System (INIS)

    Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon

    2005-01-01

    A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams

  14. Partitioning of fissile and radio-toxic materials from spent nuclear fuel

    International Nuclear Information System (INIS)

    Bychkov, A.V.; Skiba, O.V.; Kormilitsyn, M.V.

    2007-01-01

    these elements as fuel components, they could be involved in the recycling together with the main actinides, and they could be jointly extracted in the partitioning processes. It is also possible to design some special reactor systems for energy generation. For instance, Np, Am and Cm could be considered as fuel components for fast reactors. It would be possible to apply similar approaches even to the burning of uranium isotopes ( 232,234,236 U), which should be produced in a concentrated form during the re-enrichment. So the future development of innovative technologies should be directed from a complete reprocessing towards partitioning of fissile and radio-toxic materials from the spent nuclear fuel. The objectives of technology optimisation can be stated as follows: (1) reprocessing/partitioning with the view of non-proliferation, (2) partitioning with a minimal effect on the environment (3) partitioning using advanced economical methods. The criteria for the partitioning in future (after the year 2050) can be taken from the INPRO methodology. (authors)

  15. Analysis of triso packing fraction and fissile material to DB-MHR using LWR reprocessed fuel

    International Nuclear Information System (INIS)

    Silva, Clarysson A.M. da; Pereira, Claubia; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Gual, Maritza R.

    2013-01-01

    Gas-cooled and graphite-moderated reactor is being considered the next generation of nuclear power plants because of its characteristic to operate with reprocessed fuel. The typical fuel element consists of a hexagonal block with coolant and fuel channels. The fuel pin is manufactured into compacted ceramic-coated particles (TRISO) which are used to achieve both a high burnup and a high degree of passive safety. This work uses the MCNPX 2.6.0 to simulate the active core of Deep Burn Modular Helium Reactor (DB-MHR) employing PWR (Pressurized Water Reactor) reprocessed fuel. However, before a complete study of DB-MHR fuel cycle and recharge, it is necessary to evaluate the neutronic parameters to some values of TRISO Packing Fractions (PF) and Fissile Material (FM). Each PF and FM combination would generate the best behaviour of neutronic parameters. Therefore, this study configures several PF and FM combinations considering the heterogeneity of TRISO layers and lattice. The results present the best combination of PF and FM values according with the more appropriated behaviour of the neutronic parameters during the burnup. In this way, the optimized combination can be used to future works of MHR fuel cycle and recharge. (author)

  16. Fissile fuel breeding and minor actinide transmutation in the life engine

    International Nuclear Information System (INIS)

    Sahin, Suemer; Khan, Mohammad Javed; Ahmed, Rizwan

    2011-01-01

    zone (50 cm), containing MA as fissionable fuel. A 3rd ODS layer (2 cm) separates the molten salt zone on the right side from the graphite reflector (30 cm). Calculations have been conducted for a fusion driver power of 500 MW th in S 8 -P 3 approximation using 238-neutron groups. Minor actinides (MA) out of the nuclear waste of LWRs are used as fissile carbide fuel in TRISO particles with volume fractions of 0, 2, 3, 4 and 5% have been dispersed homogenously in the Flibe coolant. For these cases, tritium breeding at startup is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. In the course of plant operation, TBR and fissile neutron multiplication factor decrease gradually. For a self-sustained reactor, TBR > 1.05 can be kept for all cases over 8 years. Higher fissionable fuel content in the molten salt leads also to higher blanket energy multiplication, namely M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5% TRISO volume fraction at start up, respectively. For all investigated cases, fissile burn up exceeds 400 000 MW D/MT. Major damage mechanisms have been calculated as DPA = 50 and He = 176 appm per year. This implies a replacement of the first wall every 3 years.

  17. Self Shielding in Nuclear Fissile Assay Using LSDS

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Song, Kee Chan

    2012-01-01

    The new technology for isotopic fissile material contents assay is under development at KAERI using lead slowing down spectrometer(LSDS). LSDS is very sensitive to distinguish fission signals from each fissile isotope in spent and recycled fuel. The accumulation of spent fuel is current big issue. The amount of spent fuels will reach the maximum storage capacity of the pools soon. Therefore, an interim storage must be searched and it should be optimized in design by applying accurate fissile content. When the storage has taken effect, all the nuclear materials must be also specified and verified for safety, economics and management. Generally, the spent fuel from PWR has unburned ∼1 % U235, produced ∼0.5 % plutonium from decay chain, ∼3 % fission products, ∼ 0.1 % minor actinides (MA) and uranium remainder. About 1.5 % fissile materials still exist in the spent fuel. Therefore, for reutilization of fissile materials in spent fuel at SFR, resource material is produced through pyro process. Fissile material contents in resource material must be analyzed before fabricating SFR fuel for reactor safety and economics. In assay of fissile content of spent fuel and recycled fuel, intense radiation background gives limitation on the direct analysis of fissile materials. However, LSDS is not influenced by such a radiation background in fissile assay. Based on the decided geometry setup, self shielding parameter was calculated at the fuel assay zone by introducing spent fuel or pyro produced nuclear material. When nuclear material is inserted into the assay area, the spent fuel assembly or pyro recycled fuel material perturbs the spatial distribution of the slowing down neutrons in lead and the prompt fast fission neutrons produced by fissile materials are also perturbed. The self shielding factor is interpreted as that how much of absorption is created inside the fuel area when it is in the lead. Self shielding effect provides a non-linear property in the isotopic

  18. Feasibility of fissile mass assay of spent nuclear fuel using 252Cf-source-driven frequency-analysis

    International Nuclear Information System (INIS)

    Mattingly, J.K.; Valentine, T.E.; Mihalczo, J.T.

    1996-01-01

    The feasibility was evaluated using MCNP-DSP, an analog Monte Carlo transport cod to simulate source-driven measurements. Models of an isolated Westinghouse 17x17 PWR fuel assembly in a 1500-ppM borated water storage pool were used. In the models, the fuel burnup profile was represented using seven axial burnup zones, each with isotopics estimated by the PDQ code. Four different fuel assemblies with average burnups from fresh to 32 GWd/MTU were modeled and analyzed. Analysis of the fuel assemblies was simulated by inducing fission in the fuel using a 252 Cf source adjacent to the assembly and correlating source fissions with the response of a bank of 3 He detectors adjacent to the assembly opposite the source. This analysis was performed at 7 different axial positions on each of the 4 assemblies, and the source-detector cross-spectrum signature was calculated for each of these 28 simulated measurements. The magnitude of the cross-spectrum signature follows a smooth upward trend with increasing fissile material ( 235 U and 239 Pu) content, and the signature is independent of the concentration of spontaneously fissioning isotopes (e.g., 244 Cm) and (α,n) sources. Furthermore, the cross-spectrum signature is highly sensitive to changes in fissile material content. This feasibility study indicated that the signature would increase ∼100% in response to an increase of only 0.1 g/cm 3 of fissile material

  19. LSDS Development for Isotopic Fissile Content Assay

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Lee, Jung Won; Song, Kee Chan

    2010-01-01

    Concerning the sustainable energy supply and green house effect, nuclear energy became the most feasible option to meet the energy demand in Korea. However, the production of the spent nuclear fuel is the inevitable situation. Since the first nuclear power plant started to produce the electricity in Korea, the accumulated amount of spent fuels exceeded 10k tomes recently. The accumulation of the spent fuels is the big issue in the society. Therefore, as an option which strengthens the nuclear proliferation resistance and reduces the amount of spent fuels, sodium fast reactor (SFR) program linked with pyro-processing is under development to re-use the PWR spent fuel and produce the energy. In the process, the produced metallic material involves uranium and TRU (transuranic; neptunium, plutonium, and americium). The uranium-TRU is used to fabricate SFR fuel. The burning the recycled fuel in the reactor is to solve the current spent fuel storage problem and to minimize the actinides accumulation having long half-life. Generally, the spent fuel from PWR has unburned ∼1 % U235, produced ∼0.5 % plutonium from decay chain, ∼3 % fission products, ∼ 0.1 % minor actinides (MA) and uranium remainder. About 1.5 % fissile materials still exist in the spent fuel. Therefore, spent fuel is not only waste but energy resource. The direct and isotopic fissile content assay is the crucial technology for the spent fuel reuse. Additionally, the fissile content analysis will contribute to the optimum storage design and safe spent fuel management. Several nondestructive technologies have been developed for the spent fuel assay; gamma ray measurement, passive and active neutron measurements. Spent fuel emits intense gamma rays and neutrons by (a, n) and spontaneous fission. This intense background has the limitation on the direct analysis of fissile materials. Recently, to analyze the individual fissile content, leadslowing down spectrometer (LSDS) has been being developed in Korea

  20. Fuel costs of a light water reactor with fissile material recycling

    International Nuclear Information System (INIS)

    Clauss, J.

    1984-01-01

    In the light of the present prices of natural uranium and separative work and fabrication costs, savings can be achieved by reloading recycled fissile material. As in all recycling techniques, the product recovered cannot meet the whole new requirement. No excessive economic expectations should be associated with fissile material recycling in ligth water reactors. The main advantages of the procedure are the conservation of resources and the safety against proliferation. Besides, the original purpose of reprocessing should not be forgotten, i.e., in addition to the recycling of fissile material, to have a safe and easy method of secular disposal of high level waste (concentrated fission products). (orig.) [de

  1. Predicting fissile content of spent nuclear fuel assemblies with the Passive Neutron Albedo Reactivity technique and Monte Carlo code emulation

    International Nuclear Information System (INIS)

    Conlin, Jeremy Lloyd; Tobin, Stephen J.

    2011-01-01

    There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed. (author)

  2. Operational experience in the non-destructive assay of fissile material in General Electric's nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Stewart, J.P.

    1976-01-01

    Operational experience in the non-destructive assay of fissile material in a variety of forms and containers and incorporation of the assay devices into the accountability measurement system for General Electric's Wilmington Fuel Fabrication Facility measurement control programme is detailed. Description of the purpose and related operational requirements of each non-destructive assay system is also included. In addition, the accountability data acquisition and processing system is described in relation to its interaction with the various non-destructive assay devices and scales used for accountability purposes within the facility. (author)

  3. Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

    Science.gov (United States)

    Andrianova, E. A.; Tsibul'skiy, V. F.

    2017-12-01

    At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and 238U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with 238U alone should be reprocessed in the usual way.

  4. Design study of a fusion-driven tokamak hybrid reactor for fissile fuel production. Final report

    International Nuclear Information System (INIS)

    Rose, R.P.

    1979-05-01

    This study evaluated conceptual approaches for a tokamak fusion-driven fuel producing reactor. The conceptual design of this hybrid reactor was based on using projected state-of-the-art technology for the late 1980s. This reactor would be a demonstration plant and, therefore, first-of-a-kind considerations have been included. The conceptual definitions of two alternatives for the fusion driver were evaluated. A Two-Component Tokamak (TCT) concept, based on the TFTR plasma physics parameters, was compared to a Beam-Driven Thermonuclear (BDTN) concept, based on the USSR T-20 plasma physics parameters

  5. Methods of producing transportation fuel

    Science.gov (United States)

    Nair, Vijay [Katy, TX; Roes, Augustinus Wilhelmus Maria [Houston, TX; Cherrillo, Ralph Anthony [Houston, TX; Bauldreay, Joanna M [Chester, GB

    2011-12-27

    Systems, methods, and heaters for treating a subsurface formation are described herein. At least one method for producing transportation fuel is described herein. The method for producing transportation fuel may include providing formation fluid having a boiling range distribution between -5.degree. C. and 350.degree. C. from a subsurface in situ heat treatment process to a subsurface treatment facility. A liquid stream may be separated from the formation fluid. The separated liquid stream may be hydrotreated and then distilled to produce a distilled stream having a boiling range distribution between 150.degree. C. and 350.degree. C. The distilled liquid stream may be combined with one or more additives to produce transportation fuel.

  6. Producing liquid fuels from biomass

    Science.gov (United States)

    Solantausta, Yrjo; Gust, Steven

    The aim of this survey was to compare, on techno-economic criteria, alternatives of producing liquid fuels from indigenous raw materials in Finland. Another aim was to compare methods under development and prepare a proposal for steering research related to this field. Process concepts were prepared for a number of alternatives, as well as analogous balances and production and investment cost assessments for these balances. Carbon dioxide emissions of the alternatives and the price of CO2 reduction were also studied. All the alternatives for producing liquid fuels from indigenous raw materials are utmost unprofitable. There are great differences between the alternatives. While the production cost of ethanol is 6 to 9 times higher than the market value of the product, the equivalent ratio for substitute fuel oil produced from peat by pyrolysis is 3 to 4. However, it should be borne in mind that the technical uncertainties related to the alternatives are of different magnitude. Production of ethanol from barley is of commercial technology, while biomass pyrolysis is still under development. If the aim is to reach smaller carbon dioxide emissions by using liquid biofuels, the most favorable alternative is pyrolysis oil produced from wood. Fuels produced from cultivated biomass are more expensive ways of reducing CO2 emissions. Their potential of reducing CO2 emissions in Finland is insignificant. Integration of liquid fuel production to some other production line is more profitable.

  7. Direct conversion of surplus fissile materials, spent nuclear fuel, and other materials to high-level-waste glass

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Elam, K.R.

    1995-01-01

    With the end of the cold war the United States, Russia, and other countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. The United States Academy of Sciences (NAS) has recommended that these surplus fissile materials (SFMs) be processed so they are no more accessible than plutonium in spent nuclear fuel (SNF). This spent fuel standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. The NAS recommended investigation of three sets of options for disposition of SFMs while meeting the spent fuel standard: (1) incorporate SFMs with highly radioactive materials and dispose of as waste, (2) partly burn the SFMs in reactors with conversion of the SFMs to SNF for disposal, and (3) dispose of the SFMs in deep boreholes. The US Government is investigating these options for SFM disposition. A new method for the disposition of SFMs is described herein: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptinium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal

  8. Development and production of Zenith fissile elements

    Energy Technology Data Exchange (ETDEWEB)

    George, D; Wheatley, C C.H.; Lloyd, H

    1959-06-15

    The development of a new glass-bonded alumina-uranium oxide composition forming the fissile component of the Zenith fuel elements is described, together with the production of the initial charge containing 15 Kg. of U{sub 235]; the composition is capable of retaining fission product gases at high temperatures. The description includes criticality considerations, details of manufacture, and production statistics of the 11,000 discs produced.

  9. Nondestructive determination of burnup and fissile isotope balance in spent fuel assemblies of water cooled reactors

    International Nuclear Information System (INIS)

    Pinel, J.

    1983-03-01

    Two non-destructive methods for measuring fuel assemblies in storage pools have been developed: a gamma fuel scanning method, using the 134 Cs - 137 Cs and 144 Ce gamma rays, and the measurement of the neutron flux emitted by the fuel assembly. For interpreting the measurement, we have used calculated correlations to establish a connection between the measured phenomena and the parameters to be determined. A measurement campaign involving 58 assemblies from the C.N.A. reactor was conducted in the reprocessing plant of LA HAGUE. The results obtained show that the objectives can be achevied within an industrial environment [fr

  10. Electronuclear conversion of fertile to fissile material

    International Nuclear Information System (INIS)

    Van Atta, C.M.; Lee, J.D.; Heckrotte, W.

    1976-01-01

    The electronuclear conversion of fertile to fissile material by accelerator-produced neutrons is discussed. Experimental and theoretical results obtained in the MTA program (1949--1954) on the production of low-energy (less than 20-MeV) neutrons by high-energy proton, deuteron, and neutron bombardment of target materials are briefly reviewed. More recent calculations of the cascade process, by which the low-energy neutrons are produced, are discussed. A system is described by which 500- to 600-MeV deuterons incident on a lithium primary target can be converted to high-energy neutrons, which can be multiplied by spallation cascades and nuclear excitation to produce low-energy neutrons in a depleted-uranium or thorium secondary target. Fission events producing heat and additional neutrons are produced. The evaporation and fission neutrons would be captured, and fissile material would be produced. The production rates for 239 Pu and 233 U are estimated for 0.25-A and 0.375-A deuteron beams from an Alvarez linac. The capital and operating costs are estimated, and the resulting costs of fissile materials are calculated. The cost of generating power in reactors using the fissile material so produced as make-up fuel is also estimated. The energy multiplication (power generated in reactors so fueled/power consumed by the accelerator) ranges from about 10 to about 50 depending upon the make-up of the secondary target; depleted uranium, thorium, or a combination of the two. An experimental and theoretical program to facilitate optimization of the parameters of a production installation is described. 13 figures, 14 tables

  11. Thermal energy of nuclear origin produced in non-fissile materials (1962); Energie calorifique d'origine nucleaire degagee dans les materiaux non fissiles (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Naudet, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Millies, P; Berger, J [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1962-07-01

    A first part is devoted to the description of the interaction phenomena between elementary particles and material that may be observed during the irradiation process in a nuclear reactor: nuclear reactions due to neutrons, production of gamma rays and absorption of those gamma rays through various processes. In a second part the phenomena producing calorific energy in irradiated material are quantitatively examined. In the third part results are summed up in a formulary. The fourth part presents tables and figures giving to the reader all the numerical values necessary for practical calculations. (authors) [French] Une premiere partie est consacree a l'examen des principaux phenomenes d'interaction des particules avec la matiere qui interviennent lors d'une irradiation dans un reacteur: reactions nucleaires dues aux neutrons, production des rayons gamma et absorption de ces derniers par les divers processus. Une deuxieme partie etudie quantitativement les phenomenes qui conduisent a l'apparition d'energie calorifique dans le materiau irradie. En troisieme partie, un formulaire resume les resultats etablis. Dans une quatrieme partie, des tableaux et des courbes fournissent a l'experimentateur toutes les valeurs numeriques necessaires aux calculs pratiques. (auteurs)

  12. Accelerator based production of fissile nuclides, threshold uranium price and perspectives

    International Nuclear Information System (INIS)

    Djordjevic, D.; Knapp, V.

    1988-01-01

    Accelerator breeder system characteristics are considered in this work. One such system which produces fissile nuclides can supply several thermal reactors with fissile fuel, so this system becomes analogous to an uranium enrichment facility with difference that fissile nuclides are produced by conversion of U-238 rather than by separation from natural uranium. This concept, with other long-term perspective for fission technology on the basis of development only one simpler technology. The influence of basic system characteristics on threshold uranium price is examined. Conditions for economically acceptable production are established. (author)

  13. Accelerator based production of fissile nuclides, threshold uranium price and perspectives; Akceleratorska proizvodnja fisibilnih nuklida, granicna cijena urana i perspektive

    Energy Technology Data Exchange (ETDEWEB)

    Djordjevic, D [INIS-Inzenjering, Sarajevo (Yugoslavia); Knapp, V [Elektrotehnicki fakultet, zagreb (Yugoslavia)

    1988-07-01

    Accelerator breeder system characteristics are considered in this work. One such system which produces fissile nuclides can supply several thermal reactors with fissile fuel, so this system becomes analogous to an uranium enrichment facility with difference that fissile nuclides are produced by conversion of U-238 rather than by separation from natural uranium. This concept, with other long-term perspective for fission technology on the basis of development only one simpler technology. The influence of basic system characteristics on threshold uranium price is examined. Conditions for economically acceptable production are established. (author)

  14. Method of producing nuclear fuels

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Suzuki, Tokuyuki; Oomura, Hiroshi.

    1985-01-01

    Purpose: To fabricate a nuclear fuel assembly with uniform enrichment degree, in the blanket of a hybrid reactor. Constitution: A vessel charged with powderous source materials is conveyed by a conveying gas through a material charge/discharge tube to the inside of the blanket. Then, plasmas are formed in the inner space of the blanket so as to enrich the source materials by the irradiation of neutrons. After the average degree of enrichment reaches a predetermined level, the material vessel is discharged by the conveying gas onto a conveyor. The powder materials are separated from the source-material vessel and then charged into a source-material hopper. The mixed material of a uniform enrichment degree is supplied to a fuel-assembly-fabrication device. FP gases resulted after the enrichment are effectively separated and removed through an FP gas pipe. (Horiuchi, T.)

  15. 49 CFR 173.420 - Uranium hexafluoride (fissile, fissile excepted and non-fissile).

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Uranium hexafluoride (fissile, fissile excepted....420 Uranium hexafluoride (fissile, fissile excepted and non-fissile). (a) In addition to any other... non-fissile uranium hexafluoride must be offered for transportation as follows: (1) Before initial...

  16. Recovery of fissile materials from nuclear wastes

    Science.gov (United States)

    Forsberg, Charles W.

    1999-01-01

    A process for recovering fissile materials such as uranium, and plutonium, and rare earth elements, from complex waste feed material, and converting the remaining wastes into a waste glass suitable for storage or disposal. The waste feed is mixed with a dissolution glass formed of lead oxide and boron oxide resulting in oxidation, dehalogenation, and dissolution of metal oxides. Carbon is added to remove lead oxide, and a boron oxide fusion melt is produced. The fusion melt is essentially devoid of organic materials and halogens, and is easily and rapidly dissolved in nitric acid. After dissolution, uranium, plutonium and rare earth elements are separated from the acid and recovered by processes such as PUREX or ion exchange. The remaining acid waste stream is vitrified to produce a waste glass suitable for storage or disposal. Potential waste feed materials include plutonium scrap and residue, miscellaneous spent nuclear fuel, and uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, organic material and other carbon-containing material.

  17. Counterstreaming-ion-tokamak fissile breeder

    International Nuclear Information System (INIS)

    Jassby, D.L.; Lee, J.D.

    1976-08-01

    Tokamak plasmas fueled and heated by energetic neutral-atom beams are characterized by total ion energy greatly exceeding the electron energy. For smaller devices the largest fusion reactivity of energetic-ion plasmas is obtained when oppositely injected D 0 and T 0 beams sustain counterstreaming velocity distributions of deuterons and tritons. This scoping study investigates the net fissile and power productions of a tokamak fusion-fission reactor with a counterstreaming-ion fusion driver and a fertile blanket optimized for fissile breeding. The fusion driver has parameters R/sub o/ = 4.7 m, a = 1.0 m, B/sub t/ = 5.6 T, W/sub b/ = 100 keV (D 0 ), n tau/sub E/ = 1.4 x 10 13 cm -3 s, Q = 1.5, 14-MeV neutron production = 175 MW. The blanket contains a fast-fission zone of natural U plus Mo (7 percent), followed by a Li-bearing zone for T breeding. The reactor produces a net power of 480 MWe and supplies sufficient Pu to support a system of LWR's producing 3800 MWe, with an estimated electrical energy cost for the entire system of 27 mills/kWh

  18. Long-term criticality safety concerns associated with surplus fissile material disposition

    International Nuclear Information System (INIS)

    Choi, J.S.

    1995-01-01

    A substantial inventory of surplus fissile material would result from ongoing and planned dismantlement of US and Russian nuclear weapons. This surplus fissile material could be dispositioned by irradiation in nuclear reactors, and the resulting spent MOx fuel would be similar in radiation characteristics to regular LWR spent UO2 fuel. The surplus fissile material could also be immobilized into high-level waste forms, such as borosilicate glass, synroc, or metal-alloy matrix. The MOx spent fuel, or the immobilized waste forms, could then be directly disposed of in a geologic repository. Long-term criticality safety concerns arise because the fissile contents (i.e., Pu-239 and its decay daughter U-235) in these waste forms are higher than in LWR spent UO2 fuel. MOx spent fuel could contain 3 to 4 wt% of reactor-grade plutonium, compared to only 0.9 wt% of plutonium in LWR spent UO2 fuel. At some future time (tens of thousand of years), when the waste forms had deteriorated due to intruding groundwater, the water could mix with the long-lived fissile materials to form into a critical system. If the critical system is self-sustaining, somewhat like the natural-occurring reactor in OKLO, fission products produced could readily be available for dissolution and release out to the accessible environment, adversely affecting public health and safety. This paper will address ongoing activities to evaluate long-term criticality safety concerns associated with disposition of fissile material in a geologic setting. Issues to be addressed include the identification of a worst-case water-intrusion scenario and waste-form geometries which present the most concern for long-term criticality safety; and suggests of technical solutions for such concerns

  19. Process of producing a fuel, etc

    Energy Technology Data Exchange (ETDEWEB)

    1924-12-01

    This invention has for its object a process of producing fuels by separating a light oil from primary tar, characterized by a succession of operations comprising preliminary removal of phenols from the oils, removing sulfur completely by the application of suitable catalysts and an agent to fix the free sulfur as hydrogen sulfide; finally, washing to remove ethylenes, pyridines, and impurities from the treatment.

  20. Process for producing nuclear reactor fuel oxides

    International Nuclear Information System (INIS)

    Goenrich, H.; Druckenbrodt, W.G.

    1981-01-01

    The waste gases of the calcination process furnace in the AVC or AV/PuC process (manufacture of nuclear reactor fuel dioxides) are returned to the furnace in a closed circuit. The NH 3 produced replaces the hydrogen which would otherwise be required for reduction in this process. (orig.) [de

  1. Preliminary evaluation of fuel oil produced from pyrolysis of waste ...

    African Journals Online (AJOL)

    It could be refined further to produce domestic kerosene and gasoline. The physical and structural properties of the fuel oil produced compared favorably with that of Aviation fuel JP-4 (a wide-cut US Air force fuel). Presently African countries are importing aviation fuels. The fuel oil produced from the pyrolysis of waste water ...

  2. Method of producing granulated ceramic nuclear fuels

    International Nuclear Information System (INIS)

    Wilkinson, W.L.

    1976-01-01

    For the production of granulated ceramic nuclear fuels with a grain size spectrum as narrow as possible it is proposed to suspend the nuclear fuel powder in a non-aqueous solvent with small content of hydrogen (e.g. chloridized hydrocarbons) while adding a binding agent and then dry it by means of rays. As binding agent polybutyl methane acrylate in dibutyl phthalate is proposed. The method is described by the example of UO 2 -powder in trichloroethylene. The dry granulated material is produced in one working step. (UWI) [de

  3. Non-electrical uses of thermal energy generated in the production of fissile fuel in fusion--fission reactors: a comparative economic parametric analysis for a hybrid with or without synthetic fuel production

    International Nuclear Information System (INIS)

    Tai, A.S.; Krakowski, R.A.

    1979-01-01

    A parametric analysis has been carried out for testing the sensitivity of the synfuel production cost in relation to crucial economic and technologic quantities (investment costs of hybrid and synfuel plant, energy multiplication of the fission blanket, recirculating power fraction of the fusion driver, etc.). In addition, a minimum synfuel selling price has been evaluated, from which the fission--fusion--synfuel complex brings about a higher economic benefit than does the fusion--fission hybrid entirely devoted to fissile-fuel and electricity generation. Assuming an electricity cost of 2.7 cents/kWh, an annual investment cost per power unit of 4.2 to 6 $/GJ (132 to 189 k$/MWty) for the fission--fusion complex and 1.5 to 3 $/GJ (47 to 95 k$/MWty) for the synfuel plant, the synfuel production net cost (i.e., revenue = cost) varies between 6.5 and 8.6 $/GJ. These costs can compete with those obtained by other processes (natural gas reforming, resid partial oxidation, coal gasification, nuclear fission, solar electrolysis, etc.). This study points out a potential use of the fusion--fission hybrid other than fissile-fuel and electricity generation

  4. Romania, producer and consumer of nuclear fuel

    International Nuclear Information System (INIS)

    Iuhas, Tiberius

    1998-01-01

    A historical sketch of the activity of Romanian Rare Metals Enterprises is presented stressing the valorization of rare metals like: - radioactive metals, uranium and thorium; - dispersed rare metals, molybdenum, monazite; - heavy and refractory metals, titanium and zirconium; rare earths, lanthanides and yttrics. The beginning and developing of research in the nuclear field is in closed relation to the existence on the domestic territory of important uranium ores the mining of which begun early in 1954. The exploitation of Baita-Bihor orebody was followed by that at Ciudanovita, Natra and Dobrei ores in Caras-Severin county. Concomitantly with the ore mining, geological research was developed covering vast areas of country's surface and using advanced investigation tools suitable for increasing depths. The next step in the nuclear fuel program was made by building a uranium concentrate (as ammonium or sodium diuranate) plant. Two purification units for processing the uranium concentrate to sintered uranium dioxide powder were completed and commissioned at Feldioara in 1986. The quality of the uranium dioxide product meets the quality standards requirements for CANDU type nuclear fuel as certified in 1994. Currently, part of the fuel load of Cernavoda reactor is fuel element clusters produced by Nuclear Fuel Plant at Pitesti of sintered powder processed at Feldioara. A list of strategic objectives of the Uranium National Company is presented among which: - maintaining the uranium mining and milling activities in close relation with the fuel requirements of Cernavoda NPP; continuing geological research in promising zones, to find new uranium orebodies, easy to mill cost effectively; decreasing the environmental impact in the geological research areas, in mining and transport affected areas and in the processing plants. The fuel demand of current operation of Cernavoda NPP Unit 1 as well as of future Unit 2 after commissioning are and will be satisfied by the

  5. The incorporation of boron in fissile transport packages for the transport and interim storage of irradiated light water reactor fuels

    International Nuclear Information System (INIS)

    Hunter, I.J.

    1998-01-01

    Boron is widely used in the nuclear industry for capturing neutrons and it is particularly useful in the criticality control of packages for the transport and interim storage of irradiated light water reactor fuels. Such fuels are typically located in an internal frame of stainless steel or aluminium, referred to as a basket, which locates the fuel assemblies in channels. Transnucleaire has designed and supplied more than 100 baskets of varying types during the past 30 years. Boron has been incorporated in many forms. Early designs of baskets used boron in specific zones to contribute to the control of criticality. Later developments in new materials dispersed boron throughout the basket and gave designers more options for the basic forms which make up the channels. New basket concepts have been developed by Transnucleaire to meet the changing market needs for transport and interim storage and boron continues to play an important role as an efficient thermal neutron absorber. (author)

  6. Method of producing encapsulated thermonuclear fuel particles

    International Nuclear Information System (INIS)

    Smith, W.H.; Taylor, W.L.; Turner, H.L.

    1976-01-01

    A method of producing a fuel particle is disclosed, which comprises forming hollow spheroids which have a mass number greater than 50, immersing said spheroids while under the presence of pressure and heat in a gaseous atmosphere containing an isotope, such as deuterium and tritium, so as to diffuse the gas into the spheroid and thereafter cooling said spheroids up to about 77 0 Kelvin to about 4 0 Kelvin. 4 Claims, 3 Drawing Figures

  7. A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushuev, A. V.; Kozhin, A. F., E-mail: alexfkozhin@yandex.ru; Aleeva, T. B.; Zubarev, V. N.; Petrova, E. V.; Smirnov, V. E. [National Research Nuclear University MEPhI (Russian Federation)

    2016-12-15

    An active neutron method for measuring the residual mass of {sup 235}U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual {sup 235}U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of {sup 238}U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

  8. An economic parametric analysis of the synthetic fuel produced by a fusion-fission complex

    International Nuclear Information System (INIS)

    Tai, A.S.; Krakowski, R.A.

    1980-01-01

    A simple analytic model is used to examine economic constraints of a fusion-fission complex in which a portion of a thermal energy is used for producing synthetic fuel (synfuel). Since the values of many quantities are not well-known, a parametric analysis has been carried out for testing the sensitivity of the synfuel production cost in relation to crucial economic and technological quantities (investment costs of hybrid and synfuel plants, energy multiplication of the fission blanket, recirculating power fraction of the fusion driver, etc.). In addition, a minimum synfuel selling price has been evaluated, from which the fission-fusion-synfuel complex brings about a higher economic benefit than does the fusion-fission hybrid entirely devoted to fissile-fuel and electricity generation. This paper describes the energy flow diagram of fusion-fission synfuel concept, express the revenue-to-cost formulation and the breakeven synfuel selling price. The synfuel production cost given by the model is evaluated within a range of values of crucial parameters. Assuming an electric cost of 2.7 cents/kWh, an annual investment cost per energy unit of 4.2 to 6 $/FJ for the fusion-fission complex and 1.5 to 3 $/GJ for the synfuel plant, the synfuel production cost lies between 6.5 and 8.5 $/GJ. These production costs can compete with those evaluated for other processes. The study points out a potential use of the fusion-fission hybrid reactor for other than fissile-fuel and electricity generation. (orig.) [de

  9. Overview of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Knief, R.A.

    1978-01-01

    The nuclear fuel cycle is substantially more complicated than the energy production cycles of conventional fuels because of the very low abundance of uranium 235, the presence of radioactivity, the potential for producing fissile nuclides from irradiation, and the risk that fissile materials will be used for nuclear weapons. These factors add enrichment, recycling, spent fuel storage, and safeguards to the cycle, besides making the conventional steps of exploration, mining, processing, use, waste disposal, and transportation more difficult

  10. Process to produce homogenized reactor fuels

    International Nuclear Information System (INIS)

    Hart, P.E.; Daniel, J.L.; Brite, D.W.

    1980-01-01

    The fuels consist of a mixture of PuO 2 and UO 2 . In order to increase the homogeneity of mechanically mixed fuels the pellets are sintered in a hydrogen atmosphere with a sufficiently low oxygen potential. This results in a reduction of Pu +4 to Pu +3 . By the reduction process water vapor is obtained increasing the pressure within the PuO 2 particles and causing PuO 2 to be pressed into the uranium oxide structure. (DG) [de

  11. Development of lead slowing down spectrometer for isotopic fissile assay

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Park, Chang Je; Ahn, Sang Joon; Kim, Ho Dong

    2014-01-01

    A lead slowing down spectrometer (LSDS) is under development for analysis of isotopic fissile material contents in pyro-processed material, or spent fuel. Many current commercial fissile assay technologies have a limitation in accurate and direct assay of fissile content. However, LSDS is very sensitive in distinguishing fissile fission signals from each isotope. A neutron spectrum analysis was conducted in the spectrometer and the energy resolution was investigated from 0.1eV to 100keV. The spectrum was well shaped in the slowing down energy. The resolution was enough to obtain each fissile from 0.2eV to 1keV. The detector existence in the lead will disturb the source neutron spectrum. It causes a change in resolution and peak amplitude. The intense source neutron production was designed for ∼E12 n's/sec to overcome spent fuel background. The detection sensitivity of U238 and Th232 fission chamber was investigated. The first and second layer detectors increase detection efficiency. Thorium also has a threshold property to detect the fast fission neutrons from fissile fission. However, the detection of Th232 is about 76% of that of U238. A linear detection model was set up over the slowing down neutron energy to obtain each fissile material content. The isotopic fissile assay using LSDS is applicable for the optimum design of spent fuel storage to maximize burnup credit and quality assurance of the recycled nuclear material for safety and economics. LSDS technology will contribute to the transparency and credibility of pyro-process using spent fuel, as internationally demanded.

  12. Materials technology for accelerator production of fissile isotopes

    International Nuclear Information System (INIS)

    Horak, J.A.

    1978-02-01

    The materials used for the accelerator production of fissile isotopes must enable the facility to achieve maximum fuel production at a minimum cost. Neutron production in the target would be maximized by use of thorium cooled with Pb--56 percent Bi or with sodium. The thorium should be ion-plated with approximately 1 mil of nickel or stainless steel for retention of fission products. The target container will have to be replaced at frequent intervals because of the copious quantities of neutronically produced helium and hydrogen in the container. Replacement would coincide with shutdown of the facility for the removal of the fissile material produced. If sodium is used to cool both the target and fertile blanket, a simple basket-type target container could be used. This would greatly reduce radiation effects in the target container. Type 316 stainless steel or V--20 wt percent Ti should perform satisfactorily as a target container. The fertile blanket should be 233 Th or 238 U that is coated with approximately 1 mil of nickel or stainless steel and cooled with sodium. The blanket container could be an austenitic stainless steel such as type 304 or 316; some ferritic alloys may also provide a satisfactory blanket container. 31 references

  13. Process and device to produce fuel briquettes

    Energy Technology Data Exchange (ETDEWEB)

    Caroe, C J

    1980-10-23

    A two-stage process for the production of briquettes consisting essentially of cellulose (sawdust, peanut shells) is proposed. The fuel material (in case with additives) is molded by high pressure to pellets of the size of a few centimeters. The pellets are mixed with flammable binding agents like paraffin, wax, polyethylene etc. and molded at a lower pressure or extruded in a second step. A suited molding device is described. The wax content could be lowered with respect to known processes.

  14. Attempt to produce silicide fuel elements in Indonesia

    International Nuclear Information System (INIS)

    Soentono, S.; Suripto, A.

    1991-01-01

    After the successful experiment to produce U 3 Si 2 powder and U 3 Si 2 -Al fuel plates using depleted U and Si of semiconductor quality, silicide fuel was synthesized using x -Al available at the Fuel Element Production Installation (FEPI) at Serpong, Indonesia. Two full-size U 3 Si 2 -Al fuel elements, having similar specifications to the ones of U 3 O 8 -Al for the RSG-GAS (formerly known as MPR-30), have been produced at the FEPI. All quality controls required have been imposed to the feeds, intermediate, as well as final products throughout the production processes of the two fuel elements. The current results show that these fuel elements are qualified from fabrication point of view, therefore it is expected that they will be permitted to be tested in the RSG-GAS, sometime by the end of 1989, for normal (∝50%) and above normal burn-up. (orig.)

  15. Fissile materials detection

    International Nuclear Information System (INIS)

    Dumesnil, P.

    1977-03-01

    Description is given of three types of apparatus intended for controlling fossile materials in view of avoiding their diversion or preventing said products to be mixed to less dangerous radioactive wastes. The gantry-type apparatus is intended for the detection of small masses of fissile materials moving through a crossing place; the neutron gantry consists of helium 3 detectors of the type 150NH100, located inside polyethylene blocks; as for the gamma gantry, it consists of two large plastic scintillators integrated to the vertical legs of said gantry. The second apparatus is a high-efficiency detector intended for controlling Pu inside waste casks. It can detect 10mg of Pu inside a 100 liters drum for one minute counting. The third apparatus intended for persons and things monitoring is still on study. Such as the gantries it is based on sampled measurement of the background noise [fr

  16. Fissile material proliferation risk

    International Nuclear Information System (INIS)

    Dreicer, J.S.; Rutherford, D.A.

    1996-01-01

    The proliferation risk of a facility depends on the material attractiveness, level of safeguards, and physical protection applied to the material in conjunction with an assessment of the impact of the socioeconomic circumstances and threat environment. Proliferation risk is a complementary extension of proliferation resistance. The authors believe a better determination of nuclear proliferation can be achieved by establishing the proliferation risk for facilities that contain nuclear material. Developing a method that incorporates the socioeconomic circumstances and threat environment inherent to each country enables a global proliferation assessment. To effectively reduce the nuclear danger, a broadly based set of criteria is needed that provides the capability to relatively assess a wide range of nuclear related sites and facilities in different countries and still ensure a global decrease in proliferation risk for fissile material (plutonium and highly enriched uranium)

  17. Preliminary evaluation of fuel oil produced from pyrolysis of low ...

    African Journals Online (AJOL)

    MICHAEL

    The wax content decreases as temperature increases .The highest quantity ... polyethylene are generated . The producers of ... increase in the volume of waste generated daily by its usage in .... aviation industry and other domestic fuel users.

  18. Fossil fuel produced radioactivities and their effect on foodchains

    Energy Technology Data Exchange (ETDEWEB)

    Okamoto, K [New South Wales Univ., Kensington (Australia). Dept. of Applied Mathematics

    1980-10-01

    The environmental impact of radioactivities produced from fossil fuel burning is not necessarily small compared with that of nuclear energy. The effect of these radioactivities on the foodchain through seafoods is discussed.

  19. Agricultural residues as fuel for producer gas generation

    Energy Technology Data Exchange (ETDEWEB)

    Hoeglund, C

    1981-01-01

    This paper reports on results from a series of tests with four different types of agricultural residues as fuel for producer gas generation. The fuels are coconut shells, coconut husks, pelletized wheat-straw and pressed sugar cane. The tests were made with a 73 Hp (50 kW) agricultural tractor diesel engine equipped with a standard gasifier developed for wood chips in Sweden, and run on a testbed at the Swedish National Machinery Testing Institute. The engine was operated on approximately 10% diesel oil and 90% producer gas. The gas composition, its calorific value and temperature, the pressure drop and the engine power were monitored. Detailed elementary analysis of the fuel and gas were carried out. Observations were also made regarding the important aspects of bridging and slagging in the gasifier. The tests confirmed that coconut shells make an excellent fuel for producer gas generation. After 8 hours of running no problems with slags and bridging were experienced. Coconut husks showed no bridging but some slag formation. The gasifier operated satisfactorily for this fuel. Pelletized wheat straw and pressed sugar cane appeared unsuitable as fuel in the unmodified test gasifier (Type F 300) due to slag formation. It is important to note, however, that the present test results are not optimal for any of the fuels used, the gasifier being designed for wood-chips and not for the test-fuels used. Tests using approximately modified gasifiers are planned for the future.

  20. Repository for fissile materials

    International Nuclear Information System (INIS)

    Gablin, K.A.

    1976-01-01

    A repository for holding and storing fissile or other hazardous materials either under or above the ground is provided by enclosing one or more inner containers, such as standard steel drums, in a larger, corrosion-resistant outer shell, with a layer of foamed polyurethane occupying the space therebetween. The polyurethane foam is free of voids at its interfaces with the inner container and outer shell, and adheres to and reinforces same to provide a stress skin structure. Protection is afforded by the chemical and physical characteristics of the polyurethane foam against destructive influences such as water vapor intrusion, package leakage and damaging effects of the environment, such as freezing, electrolysis, chemical and bacterial action. The outer shell is shaped to conform generally to the shape of the inner container and is made of a tube of bituminized fiber material with endcaps of exterior grade plywood treated with wood preservative. A quantity of fluorescein dye is positioned within the inner container for monitoring each package for leakage

  1. 49 CFR 172.441 - FISSILE label.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false FISSILE label. 172.441 Section 172.441... SECURITY PLANS Labeling § 172.441 FISSILE label. (a) Except for size and color, the FISSILE label must be... FISSILE label must be white. [69 FR 3669, Jan. 26, 2004] ...

  2. Warhead and fissile-material declarations

    International Nuclear Information System (INIS)

    von Hippel, F.

    1992-01-01

    Until recently, arms control agreements were limited by the fact that the only available verification capabilities were national technical means, which involved instruments in space or beyond national borders. As a result, the SALT II treaty constrained only the construction of large missile silos, ballistic-missile submarines and long-range bombers - and limited the flight testing of long-range ballistic missiles. Recently, however, on-site verification has been accepted, making it possible in the INF treaty to extend controls to small mobile missiles and their launchers. This paper therefore outlines a comprehensive system of verifiable limits on nuclear warheads. The authors discuss in some detail the verifiability of a halt in the production of fissile materials for nuclear warheads, the verifiability of declarations of the amounts of fissile material produced for warheads prior to the production cutoff, and the establishment of a verifiable accounting system for the numbers and types of nuclear warheads possessed by each side

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

    International Nuclear Information System (INIS)

    Zidan, W.I.

    2000-01-01

    The Egyptian Fuel Manufacturing Pilot Plant, FMPP, is a Material Testing Reactor type (MTR) fuel element facility, for producing the specified fuel elements required for the Egyptian Second Research Reactor, ETRR-2. The plant uses uranium hexafluoride (UF 6 , 19.75% U 235 by wt) as a raw material which is processed through a series of the manufacturing, inspection and test plan to produce the final specified fuel elements. Radiological safety aspects during design, construction, operation, and all reasonably accepted steps should be taken to prevent or reduce the chance of accidents occurrence. (author)

  4. Core fueling to produce peaked density profiles in large tokamaks

    International Nuclear Information System (INIS)

    Mikkelsen, D.R.; McGuire, K.M.; Schmidt, G.L.; Zweben, S.J.

    1994-06-01

    Peaking the density profile increases the usable bootstrap current and the average fusion power density; this could reduce the current drive power and increase the net output of power producing tokamaks. The use of neutral beams and pellet injection to produce peaked density profiles is assessed. We show that with radially ''hollow'' diffusivity profiles (and no particle pinch) moderately peaked density profiles can be produced by particle source profiles which are peaked off-axis. The fueling penetration requirements can therefore be relaxed and this greatly improves the feasibility of generating peaked density profiles in large tokamaks. In particular, neutral beam fueling does not require MeV particle energy. Even with beam voltages of ∼200 keV, however, exceptionally good particle confinement, τ p much-gt τ E is required to achieve net electrical power generation. In system with no power production requirement (e.g., neutron sources) neutral beam fueling should be capable of producing peaked density profiles in devices as large as ITER. Fueling systems with low energy cost per particle (such as cryogenic pellet injection) must be used in power producing tokamaks when τ p ∼ τ E . Simulations with pellet injection speeds of 7 km/sec show the peaking factor, n eo /left-angle n e right-angle, approaching 2

  5. Nuclear fuel assembly with improved spectral shift-producing rods

    International Nuclear Information System (INIS)

    Ferrari, H.M.

    1987-01-01

    This patent describes a nuclear reactor having fuel assemblies and a moderator-coolant liquid flowing through the fuel assemblies, each fuel assembly including an organized array of nuclear fuel rods wherein the moderator-coolant liquid flows along the fuel rods, at least one improved spectral shift-producing rod disposed among the fuel rods. The spectra shift-producing rod consists of: (a) an elongated hollow hermetically-sealed tubular member; (b) a weakened region formed in a portion of the member, the portion being subject to rupture at a given level of internal pressure; and (c) burnable poison material contained in the member which generates gas in the member as operation of the reactor proceeds normally, the material being soluble in the moderator-coolant liquid when brought into contact therewith; (d) the given level of internal pressure being less than the maximum level of internal pressure normally expected to be generated within the member by the poison material by normal operation of the reactor

  6. 1980 Annual status report: fissile materials control and management

    International Nuclear Information System (INIS)

    1981-01-01

    The R and D activities of the JRC in the field of Fissile Material Control and Management are oriented to the development of safeguards systems in the European Community nuclear fuel cycle and to provide means for a more efficient nuclear material management within the nuclear industry

  7. New glass material oxidation and dissolution system facility: Direct conversion of surplus fissile materials, spent nuclear fuel, and other material to high-level-waste glass. Storage and disposition of weapons-usable fissile materials programmatic environmental impact statement data report: Predecisional draft

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Elam, K.R.; Reich, W.J.

    1995-01-01

    With the end of the Cold War, countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. It has been recommended that these surplus fissile materials (SFMs) be processed so that they are no more accessible than plutonium in spent nuclear fuel (SNF). This SNF standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. This report provides for the PEIS the necessary input data on a new method for the disposition of SFMs: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptunium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal. The primary form of this SNF is Hanford-N SNF with preirradiation uranium enrichments between 0.95 and 1.08%. The final product is a plutonium, low-enriched-uranium, HLW, borosilicate glass for disposition in a geological repository. The proposed conversion process is the Glass Material Oxidation and Dissolution System (GMODS), which is a new process. The initial analysis of the GMODS process indicates that a MODS facility for this application would be similar in size and environmental impact to the Defense Waste Processing Facility (DWPF) at the Savannah River Site. Because of this, the detailed information available on DWPF was used as the basis for much of the GMODS input into the SFMs PEIS

  8. Process for producing a fuel suitable for degassing from refuse

    Energy Technology Data Exchange (ETDEWEB)

    Sulzberger, J

    1975-11-20

    Utilization of the heat energy of refuse in waste incineration plants is time-consuming and expensive due to high investment and operation costs. The inventor recommends to process the refuse to a sterile, handy and storable fuel. For this propose the refuse should be crushed, kneaded and pressed. The briquettes produced in this way should be dried.

  9. Performance Study of Dual Fuel Engine Using Producer Gas as Secondary Fuel

    Directory of Open Access Journals (Sweden)

    Deepika Shaw

    2016-06-01

    Full Text Available In the present paper, development of producer gas fuelled 4 stroke diesel engine has been investigated. Producer gas from biomass has been examined and successfully operated with 4 stroke diesel engine. The effects of higher and lower loads were investigated on the dual fuel mode. The experimental investigations revealed that at lower loads dual fuel operation with producer gas shows lower efficiency due to lower combustion rate cause by low calorific value of the producer gas. Beyond 40% load the brake thermal efficiency of dual fuel operation improved due to faster combustion rate of producer gas and higher level of premixing. It can be observed that at lower load and 20% opening of producer gas the gaseous fuel substitution found to be 56% whereas at 100% opening of producer gas it reaches 78% substitution. The CO2 emission increased at high producer gas opening and high load because at 100% producer gas maximum atoms of carbons were there and at high load condition the diesel use increased. At 80% load and producer gas varying from 20% to 100. Power output was almost comparable to diesel power with marginal higher efficiency. Producer gas is one such technology which is environmentally benign and holds large promise for future.

  10. Method of storing fissile mateiral

    International Nuclear Information System (INIS)

    Onoshita, Toshio; Ishitobi, Masuhiro

    1989-01-01

    Upon storing nuclear fissile materials in a storing building, vessels packed with fissile materials are inserted into a containing chamber divided with partition walls comprising neutron absorbers and neutron moderators. Thus, released neutrons permeating the vessel are moderated by the neutron moderators and then absorbed by the neutron absorbers. Accordingly, the neutron absorbing effect by the neutron absorbers is improved, and irradiation of neutrons released from one of vessels to the other of vessels can be suppressed. Accordingly, it is possible to shorten the distance between the vessels in a contained state as much as possible, while securing the critical safety, to improve the containing density during storage. (T.M.)

  11. Producing Liquid Fuels from Coal: Prospects and Policy Issues

    Science.gov (United States)

    2008-01-01

    fraction of the weight of a plant. Most of the material in plants is cellulose , hemicellulose, or lignin . None of these substances is amenable to the...conventional fuel involved in producing the biomass. This is especially the case for non-food-crop biomass, such as corn stover, switchgrass, prairie...conversion of cellulosic materials, starches, or sugars to alcohols. Coal-to-Liquids Technologies 39 Unfortunately, annual variations in weather

  12. A light hydrocarbon fuel processor producing high-purity hydrogen

    Science.gov (United States)

    Löffler, Daniel G.; Taylor, Kyle; Mason, Dylan

    thermal efficiency is better than 67% operating at full load. This fuel processor has been integrated with a 5-kW fuel cell producing electricity and hot water.

  13. The molten salt reactor option for beneficial use of fissile material from dismantled weapons

    International Nuclear Information System (INIS)

    Gat, U.; Engel, J.R.

    1991-01-01

    The Molten Salt Reactor (MSR) option for burning fissile fuel from dismantled weapons is examined and is found very suitable for the beneficial use of this fuel. MSRs can utilize any fissile fuel in continuous operation with no special modifications, as demonstrated in the Molten Salt Reactor Experiment. Thus, MSRs are flexible while maintaining their economy. Furthermore, MSRs require only a minimum of special fuel preparation. They can tolerate denaturing and dilution of their fuel. The size of fuel shipments can be determined to optimize safety and security-all of which supports nonproliferation and resists diversion. In addition, MSRs have inherent safety features that make them acceptable and attractive. They can burn fissile material completely or can convert it to other fuels. MSRs also have the potential for burning the actinides and delivering the waste in an optimal form, thus contributing to the solution of one of the major remaining problems in the deployment of nuclear power

  14. Fissile and fertile nuclear material measurements using a new differential die-away self-interrogation technique

    International Nuclear Information System (INIS)

    Menlove, H.O.; Menlove, S.H.; Tobin, S.J.

    2009-01-01

    This paper presents a new technique for the measurement of fissile and fertile nuclear materials in spent fuel and plutonium-laden materials such as mixed oxide (MOX) fuel. The technique, called differential die-away self-interrogation, is similar to traditional differential die-away analysis, but it does not require a pulsed neutron generator or pulsed beam accelerator, and it can measure the fertile mass in addition to the fissile mass. The new method uses the spontaneous fission neutrons from 244 Cm in spent fuel and 240 Pu effective neutrons in MOX as the 'pulsed' neutron source, with an average of ∼2.7 neutrons per pulse. The time-correlated neutrons from the spontaneous fission and the subsequent induced fissions are analyzed as a function of time to determine the spontaneous fission rate, the induced fast-neutron fissions, and the induced thermal-neutron fissions. The fissile mass is determined from the induced thermal-neutron fissions that are produced by reflected thermal neutrons that originated from the spontaneous fission reaction. The sensitivity of the fissile mass measurement is enhanced by the use of two measurements, with and without a cadmium liner between the sample and a hydrogenous moderator that surrounds the sample. The fertile mass is determined from the multiplicity analysis of the neutrons detected soon after the initial triggering neutron is detected. The method obtains good sensitivity by the optimal design of two different neutron die-away regions: a short die-away for the neutron detector region and a longer die-away for the sample interrogation region.

  15. Fuel characteristics and trace gases produced through biomass burning

    Directory of Open Access Journals (Sweden)

    BAMBANG HERO SAHARJO

    2010-01-01

    Full Text Available Saharjo BH, Sudo S, Yonemura S, Tsuruta H (2010 Fuel characteristics and trace gases produced through biomass burning. Biodiversitas 11: 40-45. Indonesian 1997/1998 forest fires resulted in forest destruction totally 10 million ha with cost damaged about US$ 10 billion, where more than 1 Gt CO2 has been released during the fire episode and elevating Indonesia to one of the largest polluters of carbon in the world where 22% of world’s carbon dioxide produced. It has been found that 80-90% of the fire comes from estate crops and industrial forest plantation area belongs to the companies which using fire illegally for the land preparation. Because using fire is cheap, easy and quick and also support the companies purpose in achieving yearly planted area target. Forest management and land use practices in Sumatra and Kalimantan have evolved very rapidly over the past three decades. Poor logging practices resulted in large amounts of waste will left in the forest, greatly elevating fire hazard. Failure by the government and concessionaires to protect logged forests and close old logging roads led to and invasion of the forest by agricultural settlers whose land clearances practices increased the risk of fire. Several field experiments had been done in order to know the quality and the quantity of trace produced during biomass burning in peat grass, peat soil and alang-alang grassland located in South Sumatra, Indonesia. Result of research show that different characteristics of fuel burned will have the different level also in trace gasses produced. Peat grass with higher fuel load burned produce more trace gasses compared to alang-alang grassland and peat soil.

  16. The bio refinery; producing feed and fuel from grain.

    Science.gov (United States)

    Scholey, D V; Burton, E J; Williams, P E V

    2016-04-15

    It is both possible and practicable to produce feed and fuel from grain. Using the value of grain to produce renewable energy for transport, while using the remaining protein content of the grain as a valuable protein source for livestock and for fish, can be seen as a complimentary and optimal use of all the grain constituents. Consideration must be given to maximise the value of the yeast components, as substantial yeast is generated during the fermentation of the grain starch to produce ethanol. Yeast is a nutritionally rich feed ingredient, with potential for use both as feed protein and as a feed supplement with possible immunity and gut health enhancing properties. Bioprocessing, with the consequent economies of scale, is a process whereby the value of grain can be optimised in a way that is traditional, natural and sustainable for primarily producing protein and oil for feed with a co-product ethanol as a renewable fuel. Copyright © 2015 Elsevier Ltd. All rights reserved.

  17. Multi-Detector Analysis System for Spent Nuclear Fuel Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Reber, Edward Lawrence; Aryaeinejad, Rahmat; Cole, Jerald Donald; Drigert, Mark William; Jewell, James Keith; Egger, Ann Elizabeth; Cordes, Gail Adele

    1999-09-01

    The Spent Nuclear Fuel (SNF) Non-Destructive Analysis (NDA) program at INEEL is developing a system to characterize SNF for fissile mass, radiation source term, and fissile isotopic content. The system is based on the integration of the Fission Assay Tomography System (FATS) and the Gamma-Neutron Analysis Technique (GNAT) developed under programs supported by the DOE Office of Non-proliferation and National Security. Both FATS and GNAT were developed as separate systems to provide information on the location of special nuclear material in weapons configuration (FATS role), and to measure isotopic ratios of fissile material to determine if the material was from a weapon (GNAT role). FATS is capable of not only determining the presence and location of fissile material but also the quantity of fissile material present to within 50%. GNAT determines the ratios of the fissile and fissionable material by coincidence methods that allow the two prompt (immediately) produced fission fragments to be identified. Therefore, from the combination of FATS and GNAT, MDAS is able to measure the fissile material, radiation source term, and fissile isotopics content.

  18. Epithermal interrogation of fissile waste

    International Nuclear Information System (INIS)

    Coop, K.L.; Hollas, C.L.

    1996-01-01

    Self-shielding of interrogating thermal neutrons in lumps of fissile material can be a major source of error in transuranic waste assay using the widely employed differential dieaway technique. We are developing a new instrument, the combined thermal/epithermal neutron (CTEN) interrogation instrument to detect the occurrence of self- shielding and mitigate its effects. Neutrons are moderated in the graphite walls of the CTEN instrument to provide an interrogating flux of epithermal and thermal neutrons. The induced prompt fission neutrons are detected in proportional counters. We report the results of measurements made with the CTEN instrument, using minimal and highly self-shielding plutonium and uranium sources in 55 gallon drums containing a variety of mock waste matrices. Fissile isotopes and waste forms for which the method is most applicable, and limitations associated with the hydrogen content of the waste package/matrix are described

  19. Characteristics of sustainable bio-solid fuel produced from sewage sludge as a conventional fuel substitute

    International Nuclear Information System (INIS)

    Jung, Bongjin; Nam, Wonjun; Lee, Na-Yeon; Kim, Kyung-Hoon

    2010-01-01

    Safely final disposal of sewage sludge which is being increased every year has already become serious problems. As one of the promising technologies to solve this problem, thermal drying method has been attracting wide attention due to energy recovery from sewage sludge. This paper describes several characteristics of sustainable bio-solid fuel, as a conventional fuel substitute, produced from sewage sludge drying and granulation plant having the treatment capacity of 10 ton/ day. This plant has been successfully operated many times and is now designing for scale-up. Average moisture content of twelve kinds of bio-solid fuels produced from the plant normally less than 10 wt% and average shape of them is mainly composed of granular type having a diameter of 2-8 mm for easy handling and transportation to the final market destinations. Average higher heating value, which is one of the important properties to estimate the possibility of available energy, of bio-solid fuels is about 3800 kcal/ kg as dry basis. So they can be utilized to supply energy in the coal power plant and cement kiln etc. as a conventional fuel substitute for a beneficial reuse. Characteristics including proximate analysis, ultimate analysis, contents of heavy metals, wettability etc. of bio-solid fuels have been also analyzed for the environmentally safe re utilization. (author)

  20. Neutron source, linear-accelerator fuel enricher and regenerator and associated methods

    Science.gov (United States)

    Steinberg, Meyer; Powell, James R.; Takahashi, Hiroshi; Grand, Pierre; Kouts, Herbert

    1982-01-01

    A device for producing fissile material inside of fabricated nuclear elements so that they can be used to produce power in nuclear power reactors. Fuel elements, for example, of a LWR are placed in pressure tubes in a vessel surrounding a liquid lead-bismuth flowing columnar target. A linear-accelerator proton beam enters the side of the vessel and impinges on the dispersed liquid lead-bismuth columns and produces neutrons which radiate through the surrounding pressure tube assembly or blanket containing the nuclear fuel elements. These neutrons are absorbed by the natural fertile uranium-238 elements and are transformed to fissile plutonium-239. The fertile fuel is thus enriched in fissile material to a concentration whereby they can be used in power reactors. After use in the power reactors, dispensed depleted fuel elements can be reinserted into the pressure tubes surrounding the target and the nuclear fuel regenerated for further burning in the power reactor.

  1. Non-proliferation, safeguards, and security for the fissile materials disposition program immobilization alternatives

    Energy Technology Data Exchange (ETDEWEB)

    Duggan, R.A.; Jaeger, C.D.; Tolk, K.M. [Sandia National Labs., Albuquerque, NM (United States); Moore, L.R. [Lawrence Livermore National Lab., CA (United States)

    1996-05-01

    The Department of Energy is analyzing long-term storage and disposition alternatives for surplus weapons-usable fissile materials. A number of different disposition alternatives are being considered. These include facilities for storage, conversion and stabilization of fissile materials, immobilization in glass or ceramic material, fabrication of fissile material into mixed oxide (MOX) fuel for reactors, use of reactor based technologies to convert material into spent fuel, and disposal of fissile material using geologic alternatives. This paper will focus on how the objectives of reducing security and proliferation risks are being considered, and the possible facility impacts. Some of the areas discussed in this paper include: (1) domestic and international safeguards requirements, (2) non-proliferation criteria and measures, (3) the threats, and (4) potential proliferation, safeguards, and security issues and impacts on the facilities. Issues applicable to all of the possible disposition alternatives will be discussed in this paper. However, particular attention is given to the plutonium immobilization alternatives.

  2. Criticality Control Fissile of Materials. Proceedings of the Symposium on Criticality Control of Fissile Materials

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-05-15

    Criticality control comprises all the administrative and technical procedures which enable the storage and processing of fissile material to be carried out under conditions of nuclear safety. It is of particular importance in the safe design and operation of chemical and metallurgical plants processing fissile material, in the handling and storage of enriched fuel for reactors, and in transportation of fissile material. The growth of nuclear power, with its increasing use of fissile material and production of plutonium, is leading to an ever widening need for this discipline. This Symposium was held 4 Vulgar-Fraction-One-Half years after the only other international meeting on this topic, at which the first broad exchange of ideas and theories enabled a comparison to be drawn between the various ways in which the subject is handled in the different countries. That meeting showed that criticality safety was often achieved by procedures known to be ultra-safe, as there was a great lack of useful experimental data with which to check theoretical models. Since that time the quantities of material being processed have increased, and with the now urgent necessity of achieving economic, and hence commercially competitive, operation, the procedure of using arbitrary factors of safety is no longer adequate. Plant Managers now require good data on the basis of which they can choose a suitable factor of safety, and design a process to be safe under any foreseeable circumstances. The present Symposium showed the great increase in the amount of available experimental data and its importance in checking the now highly sophisticated computer calculations. There are many diagrams in these Proceedings with curves from which critical parameters for various configurations can be taken. The dearth of data for plutonium systems is causing some difficulty in plutonium processing plants, which are becoming commercially important. The excellent safety record of the atomic energy industry

  3. Improved moulding material for addition to nuclear fuel particles to produce nuclear fuel elements

    International Nuclear Information System (INIS)

    Miertschin, G.N.; Leary, D.F.

    1976-01-01

    A suggestion is made to improve the moulding materials used to produce carbon-contained nuclear fuel particles by a coke-reducing added substance. The nuclear fuel particles are meant for the formation of fuel elements for gas-cooled high-temperature nuclear reactors. The moulding materials are above all for the formation of coated particles which are burnt in situ in nuclear fuel element chambers out of 'green' nuclear fuel bodies. The added substance improves the shape stability of the particles forming and prevents a stiding or bridge formation between the particles or with the surrounding walls. The following are named as added substances: 1) Polystyrene and styrene-butadiene-Co polymers (mol. wt. between 5oo and 1,000,000), 2) aromatic compounds (mol. wt. 75 to 300), 3) saturated hydrocarbon polymers (mol. wt. 5,000 to 1,000,000). Additional release agents further improve the properties in the same direction (e.g. alcohols, fatty acids, amines). (orig.) [de

  4. Solar Reforming of Carbon Dioxide to Produce Diesel Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dennis Schuetzle; Robert Schuetzle

    2010-12-31

    Measurement Methods for Assessing Contaminant Levels in Captured CO2 Streams; (3) An Assessment of Current Commercial Scale Fisher-Tropsch (F-T) Technologies for the Conversion of Syngas to Fuels; (4) An Overview of CO2 Capture Technologies from Various Industrial Sources; and (5) Lifecycle Analysis for the Capture and Conversion of CO2 to Synthetic Diesel Fuel. Commercial scale Sunexus CO2 Solar Reformer plant designs, proposed in this report, should be able to utilize waste CO2 from a wide variety of industrial sources to produce a directly usable synthetic diesel fuel that replaces petroleum derived fuel, thus improving the United States energy security while also sequestering CO2. Our material balance model shows that every 5.0 lbs of CO2 is transformed using solar energy into 6.26 lbs (1.0 U.S. gallon) of diesel fuel and into by-products, which includes water. Details are provided in the mass and energy model in this report.

  5. The back-end management of fissile material at SCK-CEN

    International Nuclear Information System (INIS)

    Noynaert, L.; Massaut, V.; Braeckeveldt, M.

    1999-01-01

    The back-end management of fissile materials at SCK-CEN mainly concerns the HEU spent fuel of the BR2 (MTR) and the LEU and MOX spent fuel of the BR3, the first PWR installed in Western Europe and in decommissioning since 1987. It also concerns the experimental fuels tested in the SCK-CEN facilities. Furthermore as a result of its R and D programs in reprocessing and characterisation of spent fuel, considerable amounts of fissile materials in all kinds of forms and characteristics are stored in the different laboratories. For these, six main types of fissile materials are identified: highly enriched uranium, experimental spent fuel from the fast breeder programmes, MOX fuel, low enriched fuel, natural uranium and lab fissile materials. For the BR2 and BR3 spent fuel, various options, i.e. reprocessing, dry storage in casks and dry storage in canisters were evaluated against criteria, e.g. available techniques, safety, waste production, overall costs and policies. As a result of these studies, it was decided to opt in the case of the HEU from the BR2 reactor for the reprocessing without recovery of uranium while for the LEU and MOX fuel from the BR3 reactor, the dry storage in containers was chosen. For the others, the studies are still in progress. (author)

  6. 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)

  7. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  8. 14 CFR 26.39 - Newly produced airplanes: Fuel tank flammability.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Newly produced airplanes: Fuel tank... Tank Flammability § 26.39 Newly produced airplanes: Fuel tank flammability. (a) Applicability: This... Series 767 Series (b) Any fuel tank meeting all of the criteria stated in paragraphs (b)(1), (b)(2) and...

  9. Fissile solution dynamics: Student research

    Energy Technology Data Exchange (ETDEWEB)

    Hetrick, D.L.

    1994-09-01

    There are two research projects in criticality safety at the University of Arizona: one in dynamic simulation of hypothetical criticality accidents in fissile solutions, and one in criticality benchmarks using transport theory. We have used the data from nuclear excursions in KEWB, CRAC, and SILENE to help in building models for solution excursions. An equation of state for liquids containing gas bubbles has been developed and coupled to point-reactor dynamics in an attempt to predict fission rate, yield, pressure, and kinetic energy. It appears that radiolytic gas is unimportant until after the first peak, but that it does strongly affect the shape of the subsequent power decrease and also the dynamic pressure.

  10. The Molten Salt Reactor option for beneficial use of fissile material from dismantled weapons

    International Nuclear Information System (INIS)

    Gat, U.; Engel, J.R.; Dodds, H.L.

    1991-01-01

    The Molten Salt Reactor (MSR) option for burning fissile fuel from dismantled weapons is examined. It is concluded that MSRs are very suitable for beneficial utilization of the dismantled fuel. The MSRs can utilize any fissile fuel in continuous operation with no special modifications, as demonstrated in the Molten Salt Reactor Experiment. Thus MSRs are flexible while maintaining their economy. MSRs further require a minimum of special fuel preparation and can tolerate denaturing and dilution of the fuel. Fuel shipments can be arbitrarily small, all of which supports nonproliferation and averts diversion. MSRs have inherent safety features which make them acceptable and attractive. They can burn a fuel type completely and convert it to other fuels. MSRs also have the potential for burning the actinides and delivering the waste in an optimal form, thus contributing to the solution of one of the major remaining problems for deployment of nuclear power. 19 refs

  11. DoD use of Domestically-Produced Alternative Fuels and Alternative Fuel Vehicles

    Science.gov (United States)

    2014-04-10

    85 $21,927 Electric $171 Hydrogen $3 Liquefied Natural Gas (LNG) $4 Liquefied Petroleum Gas ( LPG ) $14 Total $25,053 Data source: GSA’s FAST Data...919 407 5,802 GAS PH 13 77 94 10 10 204 HYD DE 5 5 LNG BI 1 1 LPG BI 47 47 LPG DE 1 1 Conventional DSL DE 867 16,174 16,028 5,698 2,508 41,275...includes information on the status of: (1) use and potential use of domestically-produced alternative fuels including but not limited to, natural gas

  12. Nuclear fuel element, and method of producing same

    International Nuclear Information System (INIS)

    Armijo, J.S.; Esch, E.L.

    1986-01-01

    This invention relates to an improvement in nuclear fuel elements having a composite container comprising a cladding sheath provided with a protective barrier of zirconium metal covering the inner surface of the sheath, rendering such fuel elements more resistant to hydrogen accumulation in service. The invention specifically comprises removing substantially all zirconium metal of the barrier layer from the part of the sheath surrounding and defining the plenum region. Thus the protective barrier of zirconium metal covers only the inner surface of the fuel container in the area immediately embracing the fissionable fuel material

  13. Preliminary assessment of a symbiotic fusion--fission power system using the TH/U refresh fuel cycle

    International Nuclear Information System (INIS)

    Bender, D.J.; Lee, J.D.; Moir, R.W.

    1977-10-01

    Studies of the mirror hybrid reactor by LLL/GA have concluded that the most promising role for this reactor concept is that of a producer of fissile fuel for fission reactors. Studies to date have examined primarily the U/Pu fuel cycle with light-water reactors serving as the consumers of the hybrid-bred fissile fuel; the specific scenarios examined required reprocessing and refabrication of the bred fuel before introduction into the fission reactor. This combination of technologies was chosen to illustrate the manner in which the hybrid reactor concept could serve the needs of, and use the technology of, the fission reactor industry as it now exists (and as it was thought it would evolve). However, the current U.S. Administration has expressed strong concerns about proliferation of nuclear weapons capability and terrorist diversion of weapons-grade nuclear materials. These concerns are based on the projected technology for the light-water reactor/fast breeder reactor using the U/Pu fuel cycle and extensive reprocessing/refabrication. A symbiotic nuclear power generation concept (hybrid fissile producer plus fission burner reactors) is described which eliminates those aspects of the present nuclear fuel cycle that (may) represent significant proliferation/diversion risks. Specifically, the proposed concept incorporates the following features: (1)Th/U 233 fuel cycle, (2) no reprocessing or fabrication of fissile material, and (3) no fissile material in a weapons-grade state

  14. Disposition scenarios and safeguardability of fissile materials under START Treaty

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1993-01-01

    Under the Strategic Arms Reduction Treaty (START-I) signed in 1991 and the Lisbon Protocol of 1992, a large inventory of fissile materials will be removed from the weapons fuel cycles of the United States and the Former Soviet Union (FSU). The Lisbon Protocol calls for Ukraine, Kazakstan, and Byelarus to become nonnuclear members of the treaty and for Russia to assume the responsibility of the treaty as a nuclear weapons state. In addition, the START-II Treaty, which was signed in 1993 by the United States and Russia, further reduces deployed nuclear warheads and adds to the inventory of excess special nuclear materials (SNM). Because storage of in-tact warheads has the potential for a open-quotes breakout,close quotes it would be desirable to dismantle the warheads and properly dispose of the SNMs under appropriate safeguards to prevent their reentry into the weapons fuel cycle. The SNM recovered from dismantled warheads can be disposed of in several ways, and the final choices may be up to the country having the title to the SNM. Current plans are to store them indefinitely, leaving serious safeguards concerns. Recognizing that the underlying objective of these treaties is to prevent the fissile materials from reentering the weapons fuel cycle, it is necessary to establish a verifiable disposal scheme that includes safeguards requirements. This paper identifies some realistic scenarios for the disposal of SNM from the weapons fuel cycle and examines the safeguardability of those scenarios

  15. A 252Cf based nondestructive assay system for fissile material

    International Nuclear Information System (INIS)

    Menlove, H.O.; Crane, T.W.

    1978-01-01

    A modulated 252 Cf source assay system 'Shuffler' based on fast-or-thermal-neutron interrogation combined with delayed-neutron counting has been developed for the assay of fissile material. The 252 Cf neutron source is repetitively transferred from the interrogation position to a shielded position while the delayed neutrons are counted in a high efficiency 3 He neutron well-counter. For samples containing plutonium, this well-counter is also used in the passive coincidence mode to assay the effective 240 Pu content. The design of an optimized neutron tailoring assembly for fast-neutron interrogation using a Monte Carlo Neutron Computer Code is described. The Shuffler system has been applied to the assay of fuel pellets, inventory samples, irradiated fuel and plutonium mixed-oxide fuel. The system can assay samples with fissile contents from a few milligrams up to several kilograms using thermal-neutron interrogation for the low mass samples and fast-neutron interrogation for the high mass samples. Samples containing 235 U- 238 U, or 233 U-Th, or UO 2 -PuO 2 fuel mixtures have been assayed with the Shuffler system. (Auth.)

  16. Fissile materials in solution concentration measured by active neutron interrogation

    International Nuclear Information System (INIS)

    Romeyer Dherbey, J.; Passard, Ch.; Cloue, J.; Bignan, G.

    1993-01-01

    The use of the active neutron interrogation to measure the concentration of plutonium contained in flow solutions is particularly interesting for fuel reprocessing plants. Indeed, this method gives a signal which is in a direct relation with the fissile materials concentration. Moreover, it is less sensitive to the gamma dose rate than the other nondestructive methods. Two measure methods have been evolved in CEA. Their principles are given into details in this work. The first one consists to detect fission delayed neutrons induced by a 252 Cf source. In the second one fission prompt neutrons induced by a neutron generator of 14 MeV are detected. (O.M.)

  17. Dual fuel mode operation in diesel engines using renewable fuels: Rubber seed oil and coir-pith producer gas

    Energy Technology Data Exchange (ETDEWEB)

    Ramadhas, A.S.; Jayaraj, S.; Muraleedharan, C. [Department of Mechanical Engineering, National Institute of Technology Calicut, Calicut-673601 (India)

    2008-09-15

    Partial combustion of biomass in the gasifier generates producer gas that can be used as supplementary or sole fuel for internal combustion engines. Dual fuel mode operation using coir-pith derived producer gas and rubber seed oil as pilot fuel was analyzed for various producer gas-air flow ratios and at different load conditions. The engine is experimentally optimized with respect to maximum pilot fuel savings in the dual fuel mode operation. The performance and emission characteristics of the dual fuel engine are compared with that of diesel engine at different load conditions. Specific energy consumption in the dual-fuel mode of operation with oil-coir-pith operation is found to be in the higher side at all load conditions. Exhaust emission was found to be higher in the case of dual fuel mode of operation as compared to neat diesel/oil operation. Engine performance characteristics are inferior in fully renewable fueled engine operation but it suitable for stationary engine application, particularly power generation. (author)

  18. Temperature Profile of the Solution Vessel of an Accelerator-Driven Subcritical Fissile Solution System

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Determan, John C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-14

    Dynamic System Simulation (DSS) models of fissile solution systems have been developed and verified against a variety of historical configurations. DSS techniques have been applied specifically to subcritical accelerator-driven systems using fissile solution fuels of uranium. Initial DSS models were developed in DESIRE, a specialized simulation scripting language. In order to tailor the DSS models to specifically meet needs of system designers they were converted to a Visual Studio implementation, and one of these subsequently to National Instrument’s LabVIEW for human factors engineering and operator training. Specific operational characteristics of subcritical accelerator-driven systems have been examined using a DSS model tailored to this particular class using fissile fuel.

  19. Temperature Profile of the Solution Vessel of an Accelerator-Driven Subcritical Fissile Solution System

    International Nuclear Information System (INIS)

    Klein, Steven Karl; Determan, John C.

    2015-01-01

    Dynamic System Simulation (DSS) models of fissile solution systems have been developed and verified against a variety of historical configurations. DSS techniques have been applied specifically to subcritical accelerator-driven systems using fissile solution fuels of uranium. Initial DSS models were developed in DESIRE, a specialized simulation scripting language. In order to tailor the DSS models to specifically meet needs of system designers they were converted to a Visual Studio implementation, and one of these subsequently to National Instrument's LabVIEW for human factors engineering and operator training. Specific operational characteristics of subcritical accelerator-driven systems have been examined using a DSS model tailored to this particular class using fissile fuel.

  20. Implementation of safeguards and security for fissile materials disposition reactor alternative facilities

    International Nuclear Information System (INIS)

    Jaeger, C.D.; Duggan, R.A.; Tolk, K.M.

    1995-01-01

    A number of different disposition alternatives are being considered and include facilities which provide for long-ten-n and interim storage, convert and stabilize fissile materials for other disposition alternatives, immobilize fissile material in glass and/or ceramic material, fabricate fissile material into mixed oxide (MOX) fuel for reactors, use reactor based technologies to convert material into spent fuel, and dispose of fissile material using a number of geologic alternatives. Particular attention will be given to the reactor alternatives which include existing, partially completed, advanced or evolutionary LWRs and CANDU reactors. The various reactor alternatives are all very similar and include processing which converts Pu to a usable form for fuel fabrication, a MOX fuel fab facility located in either the US or in Europe, US LWRs or the CANDU reactors and ultimate disposal of spent fuel in a geologic repository. This paper focuses on how the objectives of reducing security risks and strengthening arms reduction and nonproliferation will be accomplished and the possible impacts of meeting these objectives on facility operations and design. Some of the areas in this paper include: (1) domestic and international safeguards requirements, (2) non-proliferation criteria and measures, (3) the threat, and (4) potential proliferation risks, the impacts on the facilities, and safeguards and security issues unique to the presence of Category 1 or strategic special nuclear material

  1. Method to produce fuel element blocks for HTR reactors

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1977-01-01

    The patent claim relates to one partial step of the multi-stage pressing process in the production of fuel elements. A binder resin with a softening point at least 15 0 C but preferably 25-40 0 C above the melting point of the lubricant is proposed. The pressed block is expelled from the forging die in the temperature interval between the melting point of the lubricant and the softening point of the binder resin. The purpose of the invention is that the pressed fuel element blocks are expelled from the machine tool without damage at a pressure low enough to protect the mechanical integrity of the coated fuel particles or fertile particles. (UA) [de

  2. Disposition of surplus fissile materials via immobilization

    International Nuclear Information System (INIS)

    Gray, L.W.; Kan, T.; Sutcliffe, W.G.; McKibben, J.M.; Danker, W.

    1995-01-01

    In the Cold War aftermath, the US and Russia have agreed to large reductions in nuclear weapons. To aid in the selection of long-term management options, the USDOE has undertaken a multifaceted study to select options for storage and disposition of surplus plutonium (Pu). One disposition alternative being considered is immobilization. Immobilization is a process in which surplus Pu would be embedded in a suitable material to produce an appropriate form for ultimate disposal. To arrive at an appropriate form, we first reviewed published information on HLW immobilization technologies to identify forms to be prescreened. Surviving forms were screened using multi-attribute utility analysis to determine promising technologies for Pu immobilization. We further evaluated the most promising immobilization families to identify and seek solutions for chemical, chemical engineering, environmental, safety, and health problems; these problems remain to be solved before we can make technical decisions about the viability of using the forms for long-term disposition of Pu. All data, analyses, and reports are being provided to the DOE Office of Fissile Materials Disposition to support the Record of Decision that is anticipated in Summer of 1996

  3. Integrating Wind And Solar With Hydrogen Producing Fuel Cells

    NARCIS (Netherlands)

    Hemmes, K.

    2007-01-01

    The often proposed solution for the fluctuating wind energy supply is the conversion of the surplus of wind energy into hydrogen by means of electrolysis. In this paper a patented alternative is proposed consisting of the integration of wind turbines with internal reforming fuel-cells, capable of

  4. What should ''damaged'' mean in air transport of fissile packages

    International Nuclear Information System (INIS)

    Luna, R.E.; Falci, F.P.; Blackman, D.

    1995-01-01

    It is likely that the ongoing process to produce the 1996 version of the IAEA Regulation for the Safe Transport of Radioactive Materials, IAEA Safety Series 6(SS 6) will result in a more stringent package qualification standard for air transport of large quantities of radioactive materials (RAM) than is included in the 1990 version. During the process to define the scope of the new requirements there was extensive discussion of their impact on, and application to, fissile material package qualification criteria. Since fissile materials are shipped in a variety of packagings ranging from exempt to Type B, each packaging of each type must be evaluated for its ability to maintain subcriticality both alone and in arrays and in both damaged and undamaged condition. In the 1990 version of SS 6 ''damaged'' means the condition of a package after it had undergone the ''tests for demonstrating the ability to withstand accident conditions in transport,'' i.e., Type B qualification tests. These tests conditions are typical of severe accidents in surface modes, but are less severe than air mode qualification test environments to be applied to Type C packages. As a result, questions arose about the need for a corresponding change in the 1996 SS 6 to define ''damaged'' to include the Type C test regime for criticality evaluations of fissile packages in air transport

  5. Energy and climate impacts of producing synthetic hydrocarbon fuels from CO(2).

    Science.gov (United States)

    van der Giesen, Coen; Kleijn, René; Kramer, Gert Jan

    2014-06-17

    Within the context of carbon dioxide (CO2) utilization there is an increasing interest in using CO2 as a resource to produce sustainable liquid hydrocarbon fuels. When these fuels are produced by solely using solar energy they are labeled as solar fuels. In the recent discourse on solar fuels intuitive arguments are used to support the prospects of these fuels. This paper takes a quantitative approach to investigate some of the claims made in this discussion. We analyze the life cycle performance of various classes of solar fuel processes using different primary energy and CO2 sources. We compare their efficacy with respect to carbon mitigation with ubiquitous fossil-based fuels and conclude that producing liquid hydrocarbon fuels starting from CO2 by using existing technologies requires much more energy than existing fuels. An improvement in life cycle CO2 emissions is only found when solar energy and atmospheric CO2 are used. Producing fuels from CO2 is a very long-term niche at best, not the panacea suggested in the recent public discourse.

  6. Method to produce carbon-cladded nuclear fuel particles

    International Nuclear Information System (INIS)

    Sturge, D.W.; Meaden, G.W.

    1978-01-01

    In the method charges of micro-spherules of fuel element are designed to have two carbon layers, whereby a one aims to achieve a uniform granulation (standard measurement). Two drums are used for this purpose connected behind one another. The micro-spherules coated with the first layer (phenolformaldehyde resin coated graphite particles) leave the first drum and enter the second one. Following the coating with a second layer, the micro-spherules are introduced into a grain size separator. The spherules that are too small are directly recycled into the second drum and those ones that are too large are recycled into the first drum after removing the graphite layers. The method may also be applied to metal cladded particles to manufacture cermet fuels. (RW) [de

  7. 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)

  8. Producing synthetic solid fuel from Kansk-Achinsk coal

    Energy Technology Data Exchange (ETDEWEB)

    Zverev, D.P.; Krichko, A.A.; Smirnova, T.S.; Markina, T.I.

    1981-01-01

    Studies were conducted by the Soviet Institute of Fossil Fuels in order to develop a technology and equipment configuration for thermal processing of coals using gas heat carriers in swirl chambers. Characteristics of the starting Irsha-Borodinskii coal and those of the products of thermal processing at 290-600 C are given. Testing the method showed that the products of high-speed thermal processing (thermocoal, semicoke, drier products) can be used as raw materials in hydrogenation, combustion, gasification, thermal benefication, briquetting and a series of other processes in metallurgy. (10 refs.) (In Russian)

  9. Studies of neutron methods for process control and criticality surveillance of fissile material processing facilities

    International Nuclear Information System (INIS)

    Zoltowski, T.

    1988-01-01

    The development of radiochemical processes for fissile material processing and spent fuel handling need new control procedures enabling an improvement of plant throughput. This is strictly related to the implementation of continuous criticality control policy and developing reliable methods for monitoring the reactivity of radiochemical plant operations in presence of the process perturbations. Neutron methods seem to be applicable for fissile material control in some technological facilities. The measurement of epithermal neutron source multiplication with heuristic evaluation of measured data enables surveillance of anomalous reactivity enhancement leading to unsafe states. 80 refs., 47 figs., 33 tabs. (author)

  10. Hydropyrolysis of biomass to produce liquid hydrocarbon fuels. Final report. Biomass Alternative-Fuels Program

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, R K; Bodle, W W; Yuen, P C

    1982-10-01

    The ojective of the study is to provide a process design and cost estimates for a biomass hydropyrolysis plant and to establish its economic viability for commercial applications. A plant site, size, product slate, and the most probable feedstock or combination of feedstocks were determined. A base case design was made by adapting IGT's HYFLEX process to Hawaiian biomass feedstocks. The HYFLEX process was developed by IGT to produce liquid and/or gaseous fuels from carbonaceous materials. The essence of the process is the simultaneous extraction of valuable oil and gaseous products from cellulosic biomass feedstocks without forming a heavy hard-to-handle tar. By controlling rection time and temperature, the product slate can be varied according to feedstock and market demand. An optimum design and a final assessment of the applicability of the HYFLEX process to the conversion of Hawaiian biomass was made. In order to determine what feedstocks could be available in Hawaii to meet the demands of the proposed hydropyrolysis plant, various biomass sources were studied. These included sugarcane and pineapple wastes, indigenous and cultivated trees and indigenous and cultivated shrubs and grasses.

  11. Design and evaluation of an on-line fuel rod assay device for an HTGR fuel refabrication plant

    International Nuclear Information System (INIS)

    Rushton, J.E.; Allen, E.J.; Chiles, M.M.; Jenkins, J.D.

    1979-11-01

    Refabricated HTGR fuel rods will contain from approx. 0.15 to 0.5 g 233 U and/or 235 U. The fuel rods are approx. 16 mm in diameter and 62 mm long. A typical commercial fuel refabrication facility will have six fuel rod production lines, each producing approximately one fuel rod every 4 seconds at design capacity. One on-line assay device will be present for each two production lines. The relative standard deviation in an individual fuel rod fissile material measurement must be less than 3% to satisfy process and quality control requirements. Systematic errors must be kept less than approx. 0.3% for fissile material measured in fuel rods produced over two months to satisfy material accountability requirements. Several nondestructive assay (NDA) methods were investigated. Because the gamma-ray activity of the refabricated fuel is relatively high due to the presence of 232 U in the fuel and because the gamma-ray activity is not directly related to total or fissile uranium content, NDA methods employing gamma-ray detection did not appear practicable. A method using thermal neutron irradiation and fast-fission neutron detection was selected. An experimental assay device was fabricated based on this NDA method. Experiments were performed to determine the precision and accuracy of the measurements and to investigate potential interferences and systematic errors. Operating procedures were evaluated, and analysis procedures were identified

  12. A survey of processes for producing hydrogen fuel from different sources for automotive-propulsion fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Brown, L.F.

    1996-03-01

    Seven common fuels are compared for their utility as hydrogen sources for proton-exchange-membrane fuel cells used in automotive propulsion. Methanol, natural gas, gasoline, diesel fuel, aviation jet fuel, ethanol, and hydrogen are the fuels considered. Except for the steam reforming of methanol and using pure hydrogen, all processes for generating hydrogen from these fuels require temperatures over 1000 K at some point. With the same two exceptions, all processes require water-gas shift reactors of significant size. All processes require low-sulfur or zero-sulfur fuels, and this may add cost to some of them. Fuels produced by steam reforming contain {approximately}70-80% hydrogen, those by partial oxidation {approximately}35-45%. The lower percentages may adversely affect cell performance. Theoretical input energies do not differ markedly among the various processes for generating hydrogen from organic-chemical fuels. Pure hydrogen has severe distribution and storage problems. As a result, the steam reforming of methanol is the leading candidate process for on-board generation of hydrogen for automotive propulsion. If methanol unavailability or a high price demands an alternative process, steam reforming appears preferable to partial oxidation for this purpose.

  13. Underground autocatalytic-criticality potential and its implications to weapons fissile- material disposition

    International Nuclear Information System (INIS)

    Choi, J.-S.

    1998-01-01

    Several options for weapons fissile-material disposition, such as once-through mixed- oxide (MOX) fuel in reactors or immobilisation in waste glass, would result in end products requiring geologic disposal. The criticality potential of the fissile end products containing U-235 and Pu-239 and the associated consequences in a geologic setting are important considerations for the final disposal of these materials. The possibility of underground criticality, and especially autocatalytic criticality, is affected by (1) groundwater leaking into a failed waste container, (2) preferential leaching of neutron absorbers or of fissile material from a failed container, and (3) preferential deposition of fissile material in the surrounding rock. Bowman and Venneri have pointed out that fissile material mixed with varying compositions of water and silica can undergo a nuclear chain reaction. Some configurations can become autocatalytically supercritical resulting in considerable energy release, terminated finally by disassembly. Some reviews rejected the Bowman and Venneri warning as implausible because of low probabilities of scenarios that could lead to such configurations. Sanchez et al. reported possible supercritical conditions in systems of Pu-SiO 2 -H 2 O and Pu-tuff-H 2 O but concluded that the probability of forming such combinations is extremely low. Kastenberg et al. studied the potential for autocatalytic criticality of plutonium or highly enriched uranium in the proposed Yucca Mountain geologic repository. They concluded that plutonium or uranium could, theoretically, become supercritical, but that such criticality is unlikely given the hydrology, geology and geochemistry of the Yucca Mountain site. These studies are not definitive. The possibility of criticality exists. Detailed mechanisms have not been sufficiently studied for clear conclusions on the probabilities of occurrence. More technical analysis is needed to understand the potential for underground

  14. Source modulation-correlation measurement for fissile mass flow in gas or liquid fissile streams

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; March-Leuba, J.A.; Valentine, T.E.; Abston, R.A.; Mattingly, J.K.; Mullens, J.A.

    1996-01-01

    The method of monitoring fissile mass flow on all three legs of a blending point, where the input is high-enriched uranium (HEU) and low-enriched uranium (LEU) and the product is PEU, can yield the fissile stream velocity and, with calibration, the [sup235]U content. The product of velocity and content integrated over the pipe gives the fissile mass flow in each leg. Also, the ratio of fissile contents in each pipe: HEU/LEU, HEU/PEU, and PEU/LEU, are obtained. By modulating the source on the input HEU pipe differently from that on the output pipe, the HEU gas can be tracked through the blend point. This method can be useful for monitoring flow velocity, fissile content, and fissile mass flow in HEU blenddown of UF[sub 6] if the pressures are high enough to contain some of the induced fission products. This method can also be used to monitor transfer of fissile liquids and other gases and liquids that emit radiation delayed from particle capture. These preliminary experiments with the Oak Ridge apparatus show that the method will work and the modeling is adequate

  15. Simulation study of a PEM fuel cell system fed by hydrogen produced by partial oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Ozdogan, S [Marmara University, Faculty of Engineering, Istanbul (Turkey); Ersoz, A; Olgun, H [TUBITAK Marmara Research Center, Energy Systems and Environmental Research Institute, Kocaeli (Turkey)

    2003-09-01

    Within the frame of sustainable development, efficient and clean, if possible zero emission energy production technologies are of utmost importance in various sectors such as utilities, industry, households and transportation. Low-temperature fuel cell systems are suitable for powering transportation systems such as automobiles and trucks in an efficient and low-emitting manner. Proton exchange membrane (PEM) fuel cell systems constitute the most promising low temperature fuel cell option being developed globally. PEM fuel cells generate electric power from air and hydrogen or from a hydrogen rich gas via electrochemical reactions. Water and waste heat are the only by-products of PEM fuel cells. There is great interest in converting current hydrocarbon based common transportation fuels such as gasoline and diesel into hydrogen rich gases acceptable by PEM fuel cells. Hydrogen rich gases can be produced from conventional transportation fuels via various reforming technologies. Steam reforming, partial oxidation and auto-thermal reforming are the three major reforming technologies. In this paper, we discuss the results of a simulation study for a PEM fuel cell with partial oxidation. The Aspen HYSYS 3.1 code has been used for simulation purposes. Two liquid hydrocarbon fuels have been selected to investigate the effect of average molecular weights of hydrocarbons, on the fuel processing efficiency. The overall system efficiency depends on the fuel preparation and fuel cell efficiencies as well as on the heat integration within the system. It is desired to investigate the overall system efficiencies for net electrical power production at 100 kW considering bigger scale transport applications. Results indicate that fuel properties, fuel preparation system operating parameters and PEM fuel cell polarization curve characteristics all affect the overall system efficiency. (authors)

  16. High order statistical signatures from source-driven measurements of subcritical fissile systems

    International Nuclear Information System (INIS)

    Mattingly, J.K.

    1998-01-01

    This research focuses on the development and application of high order statistical analyses applied to measurements performed with subcritical fissile systems driven by an introduced neutron source. The signatures presented are derived from counting statistics of the introduced source and radiation detectors that observe the response of the fissile system. It is demonstrated that successively higher order counting statistics possess progressively higher sensitivity to reactivity. Consequently, these signatures are more sensitive to changes in the composition, fissile mass, and configuration of the fissile assembly. Furthermore, it is shown that these techniques are capable of distinguishing the response of the fissile system to the introduced source from its response to any internal or inherent sources. This ability combined with the enhanced sensitivity of higher order signatures indicates that these techniques will be of significant utility in a variety of applications. Potential applications include enhanced radiation signature identification of weapons components for nuclear disarmament and safeguards applications and augmented nondestructive analysis of spent nuclear fuel. In general, these techniques expand present capabilities in the analysis of subcritical measurements

  17. Reactor physics ideas to design novel reactors with faster fissile growth

    International Nuclear Information System (INIS)

    Jagannathan, V.; Pal, U.; Karthikeyan, R.; Raj, D.; Srivastava, A.; Khan, S. A.

    2007-01-01

    There are several types of fission reactors operating in the world adopting generally the open fuel cycle which considers the naturally available fissile nuclide, viz., 2 35U. The accumulated discharged fuel is considered as waste in some countries. However the discharged fuel contains the precious man-made fissile plutonium which would provide the sole means of harnessing the nuclear energy from either depleted uranium or the natural thorium in future. It must be emphasized that the present day power reactors use just about 0.5% of the mined uranium and it would be imprudent to discard the rest of the mass as waste. It is therefore necessary to explore ways and means of exploiting the fertile mass which has the potential of providing the energy without the green house effects for millennia to come. This has to be done by innovating means of large scale fertile to fissile conversion and then using the man-made fissile material for sustenance as well as growth of fission nuclear power. This paper attempts to give a broad picture of the available options and the challenges in realizing the theoretical possibilities

  18. Guidance on Biogas used to Produce CNG or LNG under the Renewable Fuel Standard Program

    Science.gov (United States)

    Provides EPA’s interpretation of biogas quality and RIN generation requirements that apply to renewable fuel production pathways involving the injection into a commercial pipeline of biogas for use in producing renewable CNG or renewable LNG.

  19. Performance and emission comparison of a supercharged dual-fuel engine fueled by producer gases with varying hydrogen content

    Energy Technology Data Exchange (ETDEWEB)

    Mohon Roy, Murari [Rajshahi University of Engineering and Technology (JSPS Research Fellow, Okayama University), Tsushima-Naka 3, Okayama 700-8530 (Japan); Department of Mechanical Engineering, Okayama University, Tsushima-Naka 3, Okayama 700-8530 (Japan); Tomita, Eiji; Kawahara, Nobuyuki; Harada, Yuji [Department of Mechanical Engineering, Okayama University, Tsushima-Naka 3, Okayama 700-8530 (Japan); Sakane, Atsushi (Mitsui Engineering and Shipbuilding Co. Ltd., 6-4 Tsukiji 5-chome, Chuo-ku, Tokyo)

    2009-09-15

    This study investigated the effect of hydrogen content in producer gas on the performance and exhaust emissions of a supercharged producer gas-diesel dual-fuel engine. Two types of producer gases were used in this study, one with low hydrogen content (H{sub 2} = 13.7%) and the other with high hydrogen content (H{sub 2} = 20%). The engine was tested for use as a co-generation engine, so power output while maintaining a reasonable thermal efficiency was important. Experiments were carried out at a constant injection pressure and injection quantity for different fuel-air equivalence ratios and at various injection timings. The experimental strategy was to optimize the injection timing to maximize engine power at different fuel-air equivalence ratios without knocking and within the limit of the maximum cylinder pressure. Two-stage combustion was obtained; this is an indicator of maximum power output conditions and a precursor of knocking combustion. Better combustion, engine performance, and exhaust emissions (except NO{sub x}) were obtained with the high H{sub 2}-content producer gas than with the low H{sub 2}-content producer gas, especially under leaner conditions. Moreover, a broader window of fuel-air equivalence ratio was found with highest thermal efficiencies for the high H{sub 2}-content producer gas. (author)

  20. Relation of fuel rod service parameters and design requirements to produced fuel rod and their components

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.

    1999-01-01

    Based on the presented material it is possible to state that there is a very close link between the fuel operational parameters and the requirements for its design and production process. The required performance and life-time of a fuel rod can be only assured by the correctly selected design and process solutions. The economical aspect of this problem is significantly depend on the commercial feasibility of a particular selected solution with the provision of an automated and comparative by inexpensive production of a fuel rod and its components. The operational conditions are also important for the life time of the fuel rods. If there are no special measures for the mitigation of the certain operation conditions the leakage of fuel elements can take place before the planned time. (authors)

  1. Electrochemically Produced Graphene for Microporous Layers in Fuel Cells.

    Science.gov (United States)

    Najafabadi, Amin Taheri; Leeuwner, Magrieta J; Wilkinson, David P; Gyenge, Előd L

    2016-07-07

    The microporous layer (MPL) is a key cathodic component in proton exchange membrane fuel cells owing to its beneficial influence on two-phase mass transfer. However, its performance is highly dependent on material properties such as morphology, porous structure, and electrical resistance. To improve water management and performance, electrochemically exfoliated graphene (EGN) microsheets are considered as an alternative to the conventional carbon black (CB) MPLs. The EGN-based MPLs decrease the kinetic overpotential and the Ohmic potential loss, whereas the addition of CB to form a composite EGN+CB MPL improves the mass-transport limiting current density drastically. This is reflected by increases of approximately 30 and 70 % in peak power densities at 100 % relative humidity (RH) compared with those for CB- and EGN-only MPLs, respectively. The composite EGN+CB MPL also retains the superior performance at a cathode RH of 20 %, whereas the CB MPL shows significant performance loss. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  2. Experience in producing LEU fuel elements for the RSG-GAS

    International Nuclear Information System (INIS)

    Suripto, A.; Soentono, S.

    1991-01-01

    To achieve a self-reliance in the operation of the 30 MW Multipurpose Research Reactor at Serpong (the RSG-GAS), a fuel element production facility has been constructed nearby. The main task of the facility is to produce MTR type fuel and control elements containing U 3 O 8 -Al dispersion LEU fuel for the RSG-GAS. The hot commissioning activity has started in early 1988 after completion of the cold commissioning using depleted uranium in 1987, marking the beginning of the real production activity. This paper briefly describes the main features of the fuel production facility, the production experience gained so far, and its current production activity. (orig.)

  3. Nuclear fuel pin

    International Nuclear Information System (INIS)

    Hartley, Kenneth; Moulding, T.L.J.; Rostron, Norman.

    1979-01-01

    Fuel pin for use in fast breeder nuclear reactors containing fissile and fertile areas of which the fissile and fertile materials do not mix. The fissile material takes the shape of large and small diameter microspheres (the small diameter microspheres can pass through the interstices between the large microspheres). The barrier layers being composed of microspheres with a diameter situated between those of the large and small microspheres ensure that the materials do not mix [fr

  4. Performance of commercially produced mixed-oxide fuels in EBR-II

    International Nuclear Information System (INIS)

    Hales, J.W.; Lawrence, L.A.

    1980-11-01

    Commercially produced fuels for the Fast Flux Test Facility (FFTF) were irradiated in EBR-II under conditions of high cladding temperature (approx. 700 0 C) and low power (approx. 200 W/cm) to verify that manufacturing processes did not introduce variables which significantly affect general fuel performance. Four interim examinations and a terminal examination were completed to a peak burnup of 5.2 at. % to provide irradiation data pertaining to fuel restructuring and dimensional stability at low fuel temperature, fuel-cladding reactions at high cladding temperature and general fuel behavior. The examinations indicate completely satisfactory irradiation performance for low heat rates and high cladding temperatures to 5.2 at. % burnup

  5. Feasibility of Technologies to Produce Coal-Based Fuels with Equal or Lower Greenhouse Gas Emissions than Petroleum Fuels

    Science.gov (United States)

    2014-12-22

    in operating pipeline compressors), and a negligible amount from coal; just under five percent was produced from biomass—mostly in the form of corn ...as is commonly reported for soy- and corn -based biofuels), and/or if biofuel production results in land use change causing deforestation (as has...produced via F-T synthesis are already approved for incorporation into commercial and military fuels, but other pathways (e.g., pyrolysis ) would

  6. Future reactors and their fuel cycle

    International Nuclear Information System (INIS)

    Rastoin, J.

    1990-01-01

    Known world reserves of oil and natural gas may only last another 50 years and therefore nuclear energy will become more important in the future. Industrialised countries should also be encouraged to conserve their oil reserves to make better use of them and share them with less developed countries. France already produces 30% or more of its primary energy from uranium in the form of nuclear generated electricity. France has therefore accumulated considerable expertise in all aspects of the nuclear fuel cycle. Each stage of the fuel cycle, extraction, enrichment, fuel fabrication, fissile material utilisation, reprocessing and waste storage is discussed. The utilisation of fissile material is the most important stage and this is considered in more detail under headings: increase in burn-up, spectral shift, plutonium utilisation including recycling in pressurized water reactors and fast reactors and utilisation of reprocessed uranium. It is concluded that nuclear power for electricity production will be widely used throughout the world in the future. (UK)

  7. Rotary kiln and batch pyrolysis of waste tire to produce gasoline and diesel like fuels

    International Nuclear Information System (INIS)

    Ayanoğlu, Abdulkadir; Yumrutaş, Recep

    2016-01-01

    Highlights: • Waste Tire Oil (WTO) is produced from waste tire at rotary kiln reactor. • Physical and chemical properties of WTO and fuel samples are analyzed. • Gasoline like fuel (GLF) and diesel like fuel (DLF) are produced from the WTO-10 wt% CaO mixture at fixed bed reactor. • Physical and chemical properties of the GLF and DLF are compared with the standard fuels. - Abstract: In this study, waste tire is pyrolyzed in a rotary kiln reactor to obtain more gas, light liquid, heavy liquid, wax products, and less carbon black at their maximum yields as, 20%, 12%, 25%, 8% and 35% of the total weight (4 tones), respectively. Then, the heavy and light oils are reacted with additives such as natural zeolite (NZ) and lime (CaO) at different mass ratio as 2, 6, and 10 wt%, respectively, in the batch reactor to produce liquids similar to standard petroleum fuels. The heavy and light oils mixture samples are distillated to observe their optimum graphics which are similar to gasoline and diesel like fuel. Consequently, the best results are obtained from the CaO sample with 10 wt% in comparison to the ones from the gasoline and diesel fuels. The 10 wt% CaO light liquid mixture resembles to gasoline named as gasoline like fuel (GLF) and the 10 wt% CaO heavy liquid mixture is similar to diesel called as diesel like fuel (DLF). The chemical and physical features of the waste tire, light oil, heavy oil, GLF, and DLF are analyzed by TG (thermogravimetric)/dTG (derivative thermogravimetric), proximate, ultimate, higher heating value (HHV), fourier transform-infrared spectroscopy (FT-IR), Brunauer–Emmett–Teller (BET), sulfur, density, viscosity, gas chromatography–mass spectroscopy (GC–MS), flash point, moisture, and distillation tests. The test results are turned out to be very close to the standard petroleum fuel.

  8. Combustion Chamber Deposits and PAH Formation in SI Engines Fueled by Producer Gas from Biomass Gasification

    DEFF Research Database (Denmark)

    Ahrenfeldt, Jesper; Henriksen, Ulrik Birk; Schramm, Jesper

    2003-01-01

    Investigations were made concerning the formation of combustion chamber deposits (CCD) in SI gas engines fueled by producer gas. The main objective was to determine and characterise CCD and PAH formation caused by the presence of the light tar compounds phenol and guaiacol in producer gas from an...

  9. Fuel cycle parameters for strategy studies

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1979-05-01

    This report summarizes seven fuel cycle parameters (efficiency, specific power, burnup, equilibrium net fissile feed, equilibrium net fissile surplus, first charge fissile content, and whether or not fuel reprocessing is required) to be used in long-term strategy analyses of fuel cycles based on natural UO 2 , low enriched uranium, mixed oxides, plutonium topped thorium, uranium topped thorium, and the fast breeder oxide cycle. (LL)

  10. Assessment of the U.S. regulations for fissile exemptions and fissile material general licenses

    International Nuclear Information System (INIS)

    Parks, C.V.; Hopper, C.M.; Lichtenwalter, J.J.; Easton, E.P.; Brochman, P.G.

    1998-05-01

    The paragraphs for general licenses for fissile material and exemptions (often termed exceptions in the international community) for fissile material have long been a part of the US Code of Federal Regulations (CFR) 10 CFR Part 71, Packaging and Transportation of Radioactive Material. More recently, the Nuclear Regulatory Commission (NRC) issued a final rule on Part 71 via emergency rule-making procedures in order to address an identified deficiency related to one of the fissile exemptions. To address the specified deficiency in a general fashion, the emergency rule adopted the approach of the 1996 Edition of the IAEA: Regulations for the Safe Transport of Radioactive Material (IAEA 1996), which places restrictions on certain moderating materials and limits the quantity of fissile material in a consignment. The public comments received by the NRC indicated general agreement with the need for restrictions on certain moderators (beryllium, deuterium, and graphite). The comments indicated concern relative to both the degree of restriction imposed (not more than 0.1% of fissile material mass) and the need to limit the fissile material mass of the consignment, particularly in light of the subsequent NRC staff position that the true intent was to provide control for limiting the fissile mass of the conveyance. The purpose of the review is to identify potential deficiencies that might be adverse to maintaining adequate subcriticality under normal conditions of transport and hypothetical accident conditions. In addition, ORNL has been asked to identify changes that would address any identified safety issues, enable inherently safe packages to continue to be unencumbered in transport, and seek to minimize the impact on current safe practices

  11. GLOBAL PROSPECTS OF SYNTHETIC DIESEL FUEL PRODUCED FROM HYDROCARBON RESOURCES IN OIL&GAS EXPORTING COUNTRIES

    OpenAIRE

    Kurevija, Tomislav; Kukulj, Nenad; Rajković, Damir

    2007-01-01

    Production of synthetic diesel fuel through Fischer-Tropsch process is a well known technology which dates from II World War, when Germany was producing transport fuel from coal. This process has been further improved in the South Africa due to period of international isolation. Today, with high crude oil market cost and increased demand of energy from China and India, as well as global ecological awareness and need to improve air quality in urban surroundings, many projects are being planned...

  12. Local tissue distribution of fissile nuclides

    International Nuclear Information System (INIS)

    Smith, J.M.

    1981-01-01

    Conventional tissue-section autoradiography of alpha-emitting actinide elements may require prohibitively long exposure times. Neutron-induced or fission-track autoradiography can be used for fissile nuclides such as 233 U, 235 U, and 239 Pu to circumvent this difficulty. The detection limit for these nuclides is about 4 x 10 -13 (weight fraction). This paper describes a specific technique for determining their microdistribution with histologically stained tissue sections

  13. Method and device for fabricating dispersion fuel comprising fission product collection spaces

    Science.gov (United States)

    Shaber, Eric L; Fielding, Randall S

    2015-05-05

    A method of fabricating a nuclear fuel comprising a fissile material, one or more hollow microballoons, a phenolic resin, and metal matrix. The fissile material, phenolic resin and the one or more hollow microballoons are combined. The combined fissile material, phenolic resin and the hollow microballoons are heated sufficiently to form at least some fissile material carbides creating a nuclear fuel particle. The resulting nuclear fuel particle comprises one or more fission product collection spaces. In a preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by forming the fissile material into microspheres. The fissile material microspheres are then overcoated with the phenolic resin and microballoon. In another preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by overcoating the microballoon with the fissile material, and phenolic resin.

  14. Computer-assisted nuclear fuel manufacture

    International Nuclear Information System (INIS)

    Maloney, J.P.; Schaumann, S.M.; Stone, E.

    1976-01-01

    At the ERDA Savannah River Plant, a process monitor, which incorporates an online digital computer, assists in manufacturing fuel elements used to produce nuclides such as plutonium, tritium, and californium in the plant's nuclear reactors. Also, inventory functions assist in safeguarding fissile material and protecting against accidental nuclear criticality. Terminals at strategic locations throughout the process area enable production operators to send and receive instructions and information on each manufacturing step

  15. Computer-assisted nuclear fuel manufacture

    International Nuclear Information System (INIS)

    Maloney, J.P.; Schaumann, C.M.; Stone, E.

    1976-06-01

    At the ERDA Savannah River Plant, a process monitor, which incorporates an online digital computer, assists in manufacturing fuel elements used to produce nuclides such as plutonium, tritium, and californium in the plant's nuclear reactors. Also, inventory functions assist in safeguarding fissile material and protecting against accidental nuclear criticality. Terminals at strategic locations throughout the process area enable production operators to send and receive instructions and information on each manufacturing step. 11 fig

  16. The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Peterson, P.F.; Ott, L.

    2004-01-01

    Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases

  17. Separation of silicon carbide-coated fertile and fissile particles by gas classification

    International Nuclear Information System (INIS)

    Vaughen, V.C.A.

    1976-07-01

    The separation of 235 U and 233 U in the reprocessing of HTGR fuels is a key feature of the feed-breed fuel cycle concept. This is attained in the Fort St. Vrain (FSV) reactor by coating the fissile (Th- 235 U) particles and the fertile (Th- 233 U) particles separately with silicon carbide (SiC) layers to contain the fission products and to protect the kernels from burning in the head-end reprocessing steps. Pneumatic (gas) classification based on size and density differences is the reference process for separating the SiC-coated particles into fissile and fertile streams for subsequent handling. Terminal velocities have been calculated for the +- 2 sigma ranges of particle sizes and densities for ''Fissile B''--''Fertile A'' particles used in the FSV reactor. Because of overlapping particle fractions, a continuous pneumatic separator appears infeasible; however, a batch separation process can be envisioned. Changing the gas from air to CO 2 and/or the temperature to 300 0 C results in less than 10 percent change in calculated terminal velocities. Recently reported work in gas classification is discussed in light of the theoretical calculations. The pneumatic separation of fissile and fertile particles needs more study, specifically with regard to (1) measuring the recoveries and separation efficiencies of actual fissile and fertile fractions in the tests of the pneumatic classifiers; and (2) improving the contactor design or flowsheet to avoid apparent flow separation or flooding problems at the feed point when using the feed rates required for the pilot plant

  18. Molybdenum-99-producing 37-element fuel bundle neutronically and thermal-hydraulically equivalent to a standard CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: Eleodor.Nichita@uoit.ca; Haroon, J., E-mail: Jawad.Haroon@uoit.ca

    2016-10-15

    Highlights: • A 37-element fuel bundle modified for {sup 99}Mo production in CANDU reactors is presented. • The modified bundle is neutronically and thermal-hydraulically equivalent to the standard bundle. • The modified bundle satisfies all safety criteria satisfied by the standard bundle. - Abstract: {sup 99m}Tc, the most commonly used radioisotope in diagnostic nuclear medicine, results from the radioactive decay of {sup 99}Mo which is currently being produced at various research reactors around the globe. In this study, the potential use of CANDU power reactors for the production of {sup 99}Mo is investigated. A modified 37-element fuel bundle, suitable for the production of {sup 99}Mo in existing CANDU-type reactors is proposed. The new bundle is specifically designed to be neutronically and thermal-hydraulically equivalent to the standard 37-element CANDU fuel bundle in normal, steady-state operation and, at the same time, be able to produce significant quantities of {sup 99}Mo when irradiated in a CANDU reactor. The proposed bundle design uses fuel pins consisting of a depleted-uranium centre surrounded by a thin layer of low-enriched uranium. The new molybdenum-producing bundle is analyzed using the lattice transport code DRAGON and the diffusion code DONJON. The proposed design is shown to produce 4081 six-day Curies of {sup 99}Mo activity per bundle when irradiated in the peak-power channel of a CANDU core, while maintaining the necessary reactivity and power rating limits. The calculated {sup 99}Mo yield corresponds to approximately one third of the world weekly demand. A production rate of ∼3 bundles per week can meet the global demand of {sup 99}Mo.

  19. GLOBAL PROSPECTS OF SYNTHETIC DIESEL FUEL PRODUCED FROM HYDROCARBON RESOURCES IN OIL&GAS EXPORTING COUNTRIES

    Directory of Open Access Journals (Sweden)

    Tomislav Kurevija

    2007-12-01

    Full Text Available Production of synthetic diesel fuel through Fischer-Tropsch process is a well known technology which dates from II World War, when Germany was producing transport fuel from coal. This process has been further improved in the South Africa due to period of international isolation. Today, with high crude oil market cost and increased demand of energy from China and India, as well as global ecological awareness and need to improve air quality in urban surroundings, many projects are being planned regarding production of synthetic diesel fuel, known as GTL (Gas To Liquid. Most of the future GTL plants are planned in oil exporting countries, such are Qatar and Nigeria, where natural gas as by-product of oil production is being flared, losing in that way precious energy and profit. In that way, otherwise flared natural gas, will be transformed into synthetic diesel fuel which can be directly used in all modern diesel engines. Furthermore, fossil fuel transportation and distribution technology grid can be used without any significant changes. According to lower emissions of harmful gasses during combustion than fossil diesel, this fuel could in the future play a significant part of EU efforts to reach 23% of alternative fuel share till 2020., which are now mostly relied on biodiesel, LPG (liquefied petroleum gas and CNG (compressed natural gas.

  20. A review of the prospects for fusion breeding of fissile material

    International Nuclear Information System (INIS)

    Geiger, J.S.; Bartholomew, G.A.

    1981-10-01

    This report is the result of an eight month study by the AECL Fusion Status Study Group. The objectives of this study were to review the current status of fusion research, to evaluate the neutronic performance of various fusion-breeder systems, and to assess the economic and technological outlook for the fusion breeder as a source of fissile material to support CANDU reactors operating on the thorium fuel cycle

  1. USAGE OF METHYL ESTER PRODUCED FROM WASTE GRAPE AND MN ADDITIVE AS ALTERNATIVE DIESEL FUEL

    Directory of Open Access Journals (Sweden)

    Hanbey Hazar

    2017-06-01

    Full Text Available In this study, methyl ester was produced from waste grape pulp sources. The produced methyl ester was mixed with diesel in different proportions, and was tested for engine performance and emission. It was found that with increasing biodiesel content, the specific fuel consumption and exhaust temperature have increased partially, while the CO, HC and smoke emissions decreased significantly. Additionally, in the scope of this study, dodecanol, propylene glycol and Mn based additives were added to fuel B50 to improve the emission and engine performance values. With the presence of additives, an increase in the exhaust temperature was observed, while a decrease in the specific fuel consumption, CO, HC, and smoke emissions were detected.

  2. The market for fuel pellets produced from biomass and waste in the Netherlands

    International Nuclear Information System (INIS)

    Koppejan, J.; Meulman, P.D.M.

    2001-12-01

    Several initiatives are currently being developed in the Netherlands for the production of fuel pellets from waste and biomass. This report presents an overview of the current producers and (potential) users of these pellets in the Netherlands. It also outlines the Dutch and European policies and legislations concerned. Furthermore, important barriers to market development of fuel pellets are defined and future expectations are summarized. The study covers fuel pellets made from different feedstock, varying from clean biomass to waste with traces of contaminants. In each project, pellets are produced that are unique as to their product specifications, as they are usually designed for a single application. It is therefore impossible to generalize characteristics and end use. 27 refs

  3. The environmental performance of three alcohol fuel plants producers of small, medium and big scale

    International Nuclear Information System (INIS)

    Borrero, Manuel Antonio Valdes; Pereira, Jose Tomaz Vieira; Miranda, Evaristo Eduardo de

    1999-01-01

    The article discusses the following issues of alcohol fuel plants producers: sizing; performance; natural resources; environmental aspects; and electric power generation. The environmental performance concept is introduced and a performance evaluation methodology are presented and applied. The results are also presented and criticized

  4. Thermodynamic properties of the DUPIC fuel and its performance

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kwang Heon; Kim, Hee Moon [Kyung Hee Univ., Seoul (Korea, Republic of)

    1997-07-01

    This study describes thermodynamic properties of DUPIC fuel and performance. In initial state, DUPIC fuel which contains fissile materials is different from general nuclear fuel. So this study analyzed oxygen potential, thermal conductivity and specific heat of the DUPIC fuel.

  5. Developing an energy efficient steam reforming process to produce hydrogen from sulfur-containing fuels

    Science.gov (United States)

    Simson, Amanda

    Hydrogen powered fuel cells have the potential to produce electricity with higher efficiency and lower emissions than conventional combustion technology. In order to realize the benefits of a hydrogen fuel cell an efficient method to produce hydrogen is needed. Currently, over 90% of hydrogen is produced from the steam reforming of natural gas. However, for many applications including fuel cell vehicles, the use of a liquid fuel rather than natural gas is desirable. This work investigates the feasibility of producing hydrogen efficiently by steam reforming E85 (85% ethanol/15% gasoline), a commercially available sulfur-containing transportation fuel. A Rh-Pt/SiO2-ZrO2 catalyst has demonstrated good activity for the E85 steam reforming reaction. An industrial steam reforming process is often run less efficiently, with more water and at higher temperatures, in order to prevent catalyst deactivation. Therefore, it is desirable to develop a process that can operate without catalyst deactivation at more energy efficient conditions. In this study, the steam reforming of a sulfur-containing fuel (E85) was studied at near stoichiometric steam/carbon ratios and at 650C, conditions at which catalyst deactivation is normally measured. At these conditions the catalyst was found to be stable steam reforming a sulfur-free E85. However, the addition of low concentrations of sulfur significantly deactivated the catalyst. The presence of sulfur in the fuel caused catalyst deactivation by promoting ethylene which generates surface carbon species (coke) that mask catalytic sites. The amount of coke increased during time on stream and became increasingly graphitic. However, the deactivation due to both sulfur adsorption and coke formation was reversible with air treatment at 650°C. However, regenerations were found to reduce the catalyst life. Air regenerations produce exotherms on the catalyst surface that cause structural changes to the catalyst. During regenerations the

  6. Self shielding in cylindrical fissile sources in the APNea system

    International Nuclear Information System (INIS)

    Hensley, D.

    1997-01-01

    In order for a source of fissile material to be useful as a calibration instrument, it is necessary to know not only how much fissile material is in the source but also what the effective fissile content is. Because uranium and plutonium absorb thermal neutrons so Efficiently, material in the center of a sample is shielded from the external thermal flux by the surface layers of the material. Differential dieaway measurements in the APNea System of five different sets of cylindrical fissile sources show the various self shielding effects that are routinely encountered. A method for calculating the self shielding effect is presented and its predictions are compared with the experimental results

  7. Revisited. Euratom's ownership of special fissile materials

    International Nuclear Information System (INIS)

    Pelzer, Norbert

    2015-01-01

    Among all Treaties on the Foundation of the European Community, seemingly, the Euratom Treaty ist the most unobtrusive one having even nearly been declared dead occasionally. For the opponents of nuclear energy the treaty is a thorn in their side because it aims for the peaceful exploitation of nuclear energy. Actually, the treaty likewise aims for the protection of dangers of nuclear energy and encloses a bundle of collective control instruments. The protective purpose provides the community with a strong position in numerous fields towards nuclear energy users including the right to intervene in the operations of nuclear facilities. The communitie's position is further strengthened by the communitie's ownership on special fissile materials. The EAEC Treaty determines: 'Special fissile materials are owned by the community'. The material content of Euratom's ownership is limited by Article 87 of the EAEC Treaty: Unlimited right of use and consumption is granted to the properly possessors unless obligations of the Euratom Treaty oppose. Inherently, the community does not have these rights. It was asked what would be left to the owner Euratom if the properly possessor is entitled to unlimited right of use and even right of consumption.

  8. General principles of the nuclear criticality safety for handling, processing and transportation fissile materials in the USSR

    International Nuclear Information System (INIS)

    Vnukov, V.S.; Rjazanov, B.G.; Sviridov, V.I.; Frolov, V.V.; Zubkov, Y.N.

    1991-01-01

    The paper describes the general principles of nuclear criticality safety for handling, processing, transportation and fissile materials storing. Measures to limit the consequences of critical accidents are discussed for the fuel processing plants and fissile materials storage. The system of scientific and technical measures on nuclear criticality safety as well as the system of control and state supervision based on the rules, limits and requirements are described. The criticality safety aspects for various stages of handling nuclear materials are considered. The paper gives descriptions of the methods and approaches for critical risk assessments for the processing facilities, plants and storages. (Author)

  9. Characterized hydrochar of algal biomass for producing solid fuel through hydrothermal carbonization.

    Science.gov (United States)

    Park, Ki Young; Lee, Kwanyong; Kim, Daegi

    2018-06-01

    The aim of this work was to study the characterized hydrochar of algal biomass to produce solid fuel though hydrothermal carbonization. Hydrothermal carbonization conducted at temperatures ranging from 180 to 270 °C with a 60 min reaction improved the upgrading of the fuel properties and the dewatering of wet-basis biomasses such as algae. The carbon content, carbon recovery, energy recovery, and atomic C/O and C/H ratios in all the hydrochars in this study were improved. These characteristic changes in hydrochar from algal biomass are similar to the coalification reactions due to dehydration and decarboxylation with an increase in the hydrothermal reaction temperature. The results of this study indicate that hydrothermal carbonization can be used as an effective means of generating highly energy-efficient renewable fuel resources using algal biomass. Copyright © 2018 Elsevier Ltd. All rights reserved.

  10. Monte Carlo assessment of the dose rates produced by spent fuel from CANDU reactors

    International Nuclear Information System (INIS)

    Pantazi, Doina; Mateescu, Silvia; Stanciu, Marcela

    2003-01-01

    One of the technical measures considered for biological protection is radiation shielding. The implementation process of a spent fuel intermediate storage system at Cernavoda NPP includes an evolution in computation methods related to shielding evaluation: from using simpler computer codes, like MicroShield and QAD, to systems of codes, like SCALE (which contains few independent modules) and the multipurpose and multi-particles transport code MCNP, based on Monte Carlo method. The Monte Carlo assessment of the dose rates produced by CANDU type spent fuel, during its handling for the intermediate storage, is the main objective of this paper. The work had two main features: -establishing of geometrical models according to description mode used in code MCNP, capable to account for the specific characteristics of CANDU nuclear fuel; - confirming the correctness of proposed models, by comparing MCNP results and the related results obtained with other computer codes for shielding evaluation and dose rates calculations. (authors)

  11. Directions and prospects of using low grade process fuel to produce alumina

    Directory of Open Access Journals (Sweden)

    О. А. Дубовиков

    2016-08-01

    Full Text Available Power consumption across the globe is constantly increasing for a variety of reasons: growing population, industrialization and fast economic growth. The most widespread gaseous fuel – natural gas – has the low production cost. It is 2-3 times cheaper than liquid fuel production and 6-12 times cheaper than coal production. When natural gas is transported to distances from 1.5 to 2.5 thousand km by the pipeline, its cost with account of transportation is 1.5-2 times less than the cost of coal and the fuel storage facilities are not needed. Plants powered by natural gas have the higher efficiency as compared to the plants operating on other types of fuel. They are easier and cheaper to maintain and are relatively simple in automation, thus enhancing safety and improving the production process flow, do not require complicated fuel feeding or ash handling systems. Gas is combusted with a minimum amount of polluting emissions, which adds to better sanitary conditions and environment protection. But due to depletion of major energy resources many experts see the future of the global energy industry in opportunities associated with the use of solid energy carriers. From the environmental perspective solid fuel gasification is a preferred technology. The use of synthetic gas was first offered and then put to mass scale by English mechanical engineer William Murdoch. He discovered a possibility to use gas for illumination by destructive distillation of bituminous coal. After invention of the gas burner by Robert Bunsen, the illumination gas began to be used as a household fuel. The invention of an industrial gas generator by Siemens brothers made it possible to produce a cheaper generator gas which became a fuel for industrial furnaces. As the calorific value of generator gas produced through gasification is relatively low compared to natural gas, the Mining University studied possibilities to use different types of low grade process fuel at the

  12. 49 CFR 173.477 - Approval of packagings containing greater than 0.1 kg of non-fissile or fissile-excepted uranium...

    Science.gov (United States)

    2010-10-01

    ... kg of non-fissile or fissile-excepted uranium hexafluoride. 173.477 Section 173.477 Transportation... non-fissile or fissile-excepted uranium hexafluoride. (a) Each offeror of a package containing more than 0.1 kg of uranium hexafluoride must maintain on file for at least one year after the latest...

  13. Properties of residual marine fuel produced by thermolysis from polypropylene waste

    Directory of Open Access Journals (Sweden)

    Linas Miknius

    2015-06-01

    Full Text Available Thermal degradation of waste plastics with the aim of producing liquid fuel is one of the alternative solutions to landfill disposal or incineration. The paper describes thermal conversion of polypropylene waste and analysis of produced liquid fuel that would satisfy ISO 8217-2012 requirements for a residual marine fuel. Single pass batch thermolysis processes were conducted at different own vapour pressures (20-80 barg that determined process temperature, residence time of intermediates what resulted in different yields of the liquid product. Obtained products were stabilized by rectification to achieve required standard flash point. Gas chromatography and 1H NMR spectrometry show aliphatic nature of the liquid product where majority of the compounds are isoalkanes and isoalkenes. Only lightest fractions boiling up to a temperature of 72 oC have significant amount of n-pentane. Distribution of aromatic hydrocarbons is not even along the boiling range. The fractions boiling at a temperature of 128 oC and 160 oC have the highest content of monocyclic arenes – 3.16 % and 4.09 % respectively. The obtained final liquid residual product meets all but one requirements of ISO 8217-2012 for residual marine fuels.DOI: http://dx.doi.org/10.5755/j01.ms.21.2.6105

  14. Transportation of fissile materials and the danger of criticity

    International Nuclear Information System (INIS)

    Haon, D.; Leclerc, J.; Maubert, L.

    1981-01-01

    The authors examine the risk of criticity that can arise during the transportation of fissile matter. They then outline the regulations and studies made in the field of criticity-safety and the computation methods used. They discuss the applications that are reflected in the concept and design of fissile material packagings [fr

  15. 77 FR 8254 - Notice of Data Availability Concerning Renewable Fuels Produced From Palm Oil Under the RFS...

    Science.gov (United States)

    2012-02-14

    ... Concerning Renewable Fuels Produced From Palm Oil Under the RFS Program; Extension of Comment Period AGENCY... of Data Availability Concerning Renewable Fuels Produced From Palm Oil Under the RFS Program'' (the notice is herein referred to as the ``palm oil NODA''). EPA published a NODA, which included a request...

  16. 77 FR 19663 - Notice of Data Availability Concerning Renewable Fuels Produced from Palm Oil Under the RFS...

    Science.gov (United States)

    2012-04-02

    ... Concerning Renewable Fuels Produced from Palm Oil Under the RFS Program; Extension of Comment Period AGENCY... of Data Availability Concerning Renewable Fuels Produced from Palm Oil under the RFS Program'' (the notice is herein referred to as the ``palm oil NODA''). EPA published a NODA, which included a request...

  17. Producing fuel alcohol by extractive distillation: Simulating the process with glycerol

    OpenAIRE

    Ana María Uyazán; Iván Dario Gil; Jaime Aguilar; Gerardo Rodríguez Niño; Luis A Caicedo Mesa

    2006-01-01

    Downstream separation processes in biotechnology form part of the stages having most impact on a product’s final cost. The tendency throughout the world today is to replace fossil fuels with those having a renewable origin such as ethanol; this, in turn, produces a demand for the same and the need for optimising fermentation, treating vinazas and dehydration processes. The present work approaches the problem of dehydration through simulating azeotropic ethanol extractive distillation using gl...

  18. Process of producing fuels from slates or bituminous shales. [distillation at incandescent heat

    Energy Technology Data Exchange (ETDEWEB)

    Huppenbauer, M

    1902-07-31

    A process of producing a fuel from slates or bituminous shales by saturating or impregnating them after preliminary distillation with the vapors of tars, resins, oils, etc., is given. The process is characterized by the bituminous shale being submitted in the form of fragments to distillation at incandescent heat to make the shale porous and able to absorb the vapors of the substances already mentioned.

  19. Fissile materials in solution concentration measured by active neutron interrogation; Mesure de concentration en matiere fissile dans les liquides par interrogation neutronique active

    Energy Technology Data Exchange (ETDEWEB)

    Romeyer Dherbey, J.; Passard, Ch.; Cloue, J.; Bignan, G.

    1993-12-31

    The use of the active neutron interrogation to measure the concentration of plutonium contained in flow solutions is particularly interesting for fuel reprocessing plants. Indeed, this method gives a signal which is in a direct relation with the fissile materials concentration. Moreover, it is less sensitive to the gamma dose rate than the other nondestructive methods. Two measure methods have been evolved in CEA. Their principles are given into details in this work. The first one consists to detect fission delayed neutrons induced by a {sup 252} Cf source. In the second one fission prompt neutrons induced by a neutron generator of 14 MeV are detected. (O.M.). 6 refs.

  20. Thermal Cracking of Jatropha Oil with Hydrogen to Produce Bio-Fuel Oil

    Directory of Open Access Journals (Sweden)

    Yi-Yu Wang

    2016-11-01

    Full Text Available This study used thermal cracking with hydrogen (HTC to produce bio-fuel oil (BFO from jatropha oil (JO and to improve its quality. We conducted HTC with different hydrogen pressures (PH2; 0–2.07 MPa or 0–300 psig, retention times (tr; 40–780 min, and set temperatures (TC; 623–683 K. By applying HTC, the oil molecules can be hydrogenated and broken down into smaller molecules. The acid value (AV, iodine value, kinematic viscosity (KV, density, and heating value (HV of the BFO produced were measured and compared with the prevailing standards for oil to assess its suitability as a substitute for fossil fuels or biofuels. The results indicate that an increase in PH2 tends to increase the AV and KV while decreasing the HV of the BFO. The BFO yield (YBFO increases with PH2 and tr. The above properties decrease with increasing TC. Upon HTC at 0.69 MPa (100 psig H2 pressure, 60 min time, and 683 K temperature, the YBFO was found to be 86 wt%. The resulting BFO possesses simulated distillation characteristics superior to those of boat oil and heavy oil while being similar to those of diesel oil. The BFO contains 15.48% light naphtha, 35.73% heavy naphtha, 21.79% light gas oil, and 27% heavy gas oil and vacuum residue. These constituents can be further refined to produce gasoline, diesel, lubricants, and other fuel products.

  1. Combustion Chamber Deposits and PAH Formation in SI Engines Fueled by Producer Gas from Biomass Gasification

    DEFF Research Database (Denmark)

    Ahrenfeldt, Jesper; Henriksen, Ulrik Birk; Schramm, Jesper

    2003-01-01

    Investigations were made concerning the formation of combustion chamber deposits (CCD) in SI gas engines fueled by producer gas. The main objective was to determine and characterise CCD and PAH formation caused by the presence of the light tar compounds phenol and guaiacol in producer gas from...... on filters and a sorbent was used for collection of vapour phase aromatic compounds. The filters and sorbent were analysed for polycyclic aromatic hydrocarbons (PAH) formed during combustion. The measurements showed that there was no significant increase in particulate PAH emissions due to the tar compounds...

  2. Experimental investigation of solid oxide fuel cells using biomass gasification producer gases

    Energy Technology Data Exchange (ETDEWEB)

    Norheim, Arnstein

    2005-07-01

    The main objective of this thesis is theoretical and experimental investigations related to utilisation of biomass gasification producer gases as fuel for Solid Oxide Fuel Cells (SOFC). Initial fundamental steps towards a future system of combined heat and power production based on biomass gasification and SOFC are performed and include: 1) Theoretical modeling of the composition of biomass gasification producer gases. 2) Experimental investigation of SOFC performance using biomass gasification producer gas as fuel. 3) Experimental investigation of SOFC performance using biomass gasification producer gas containing high sulphur concentration. The modeling of the composition of gasifier producer gas was performed using the program FactSage. The main objective was to investigate the amount and speciation of trace species in the producer gases as several parameters were varied. Thus, the composition at thermodynamic equilibrium of sulphur, chlorine, potassium, sodium and compounds of these were established. This was done for varying content of the trace species in the biomass material at different temperatures and fuel utilisation i.e. varying oxygen content in the producer gas. The temperature interval investigated was in the range of normal SOFC operation. It was found that sulphur is expected to be found as H2S irrespective of temperature and amount of sulphur. Only at very high fuel utilisation some S02 is formed. Important potassium containing compounds in the gas are gaseous KOH and K. When chlorine is present, the amount of KOH and K will decrease due to the formation of KCI. The level of sodium investigated here was low, but some Na, NaOH and NaCl is expected to be formed. Below a certain temperature, condensation of alkali rich carbonates may occur. The temperature at which condensation begins is mainly depending on the amount of potassium present; the condensation temperature increases with increasing potassium content. In the first experimental work

  3. Interim report on core physics and fuel cycle analysis of the pebble bed reactor power plant concept

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1977-12-01

    Calculations were made to predict the performance of a pebble bed reactor operated in a mode to produce fissile fuel (high conversion or breeding). Both a one pebble design and a design involving large primary feed pebbles and small fertile pebbles were considered. A relatively short residence time of the primary pebbles loaded with 233 U fuel was found to be necessary to achieve a high breeding ratio, but this leads to relatively high fuel costs. A high fissile inventory is associated with a low C/Th ratio and a high thorium loading, causing the doubling time to be long, even though the breeding ratio is high, and the fuel cost of electrical product to be high. Production of 233 U fuel from 235 U feed was studied and performances of the converter and breeder reactor concepts were examined varying the key parameters

  4. Measurement of inventories with mixed fissile materials

    International Nuclear Information System (INIS)

    Rinard, P.M.; Krick, M.S.; Kelley, T.; Schneider, C.M.

    1997-01-01

    An inventory with a large number of diverse items containing mixtures of uranium and plutonium has been measured with two nondestructive assay (NDA) instruments used in four modes. A segmented gamma scanner (SGS) was used to find the number of cans and the positions of the fissile materials by scanning each item in front of a transmissions source; at each position, uranium and plutonium isotopics were measured with the passive gamma rays emitted. A shuffler was then used in both the passive and active modes to measure the masses of the two elements. The measured masses for the inventory items were generally in agreement with the declared values, but anomalies were identified for a small fraction of the inventory

  5. Fissile material disposition program: Screening of alternate immobilization candidates for disposition of surplus fissile materials

    International Nuclear Information System (INIS)

    Gray, L.W.

    1996-01-01

    With the end of the Cold War, the world faces for the first time the need to dismantle vast numbers of ''excess'' nuclear weapons and dispose of the fissile materials they contain, together with fissile residues in the weapons production complex left over from the production of these weapons. If recently agreed US and Russian reductions are fully implemented, tens of thousands of nuclear weapons, containing a hundred tons or more of plutonium and hundreds of tonnes* of highly enriched uranium (HEU), will no longer be needed worldwide for military purposes. These two materials are the essential ingredients of nuclear weapons, and limits on access to them are the primary technical barrier to prospective proliferants who might desire to acquire a nuclear weapons capability. Theoretically, several kilograms of plutonium, or several times that amount of HEU, is sufficient to make a nuclear explosive device. Therefore, these materials will continue to be a potential threat to humanity for as long as they exist

  6. 40 CFR 80.530 - Under what conditions can 500 ppm motor vehicle diesel fuel be produced or imported after May 31...

    Science.gov (United States)

    2010-07-01

    ... motor vehicle diesel fuel be produced or imported after May 31, 2006? 80.530 Section 80.530 Protection... FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel Temporary Compliance Option § 80.530 Under what conditions can 500 ppm motor vehicle diesel...

  7. Chemistry and the development of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Amphlett, C.B.

    1991-01-01

    This chapter traces the chemical industry's involvement in the development of the nuclear industry from wartime projects to provide fissile material for bombs to the challenge of producing nuclear power competitively in the post-war period. Skills in the chemical industry have led to the production of new fuels by simpler methods, improvements in reprocessing and advances in the management and storage of radioactive wastes. (UK)

  8. High density fuels using dispersion and monolithic fuel

    International Nuclear Information System (INIS)

    Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia; Universidade de São Paulo

    2017-01-01

    Fuel plates used in high-performance research reactors need to be converted to low-enrichment uranium fuel; the fuel option based on a monolithic formulation requires alloys to contain 6 - 10 wt% Mo. In this case, the fuel plates are composed of the metallic alloy U-10Mo surrounded by a thin zirconium layer encapsulated in aluminum cladding. This study reviewed the physical properties of monolithic forms. The constraints produced during the manufacturing process were analyzed and compared to those of dispersed fuel. The bonding process used for dispersion fuels differs from the techniques applied to foil bonding used for pure alloys. The quality of monolithic plates depends on the fabrication method, which usually involves hot isostatic pressing and the thermal annealing effect of residual stress, which degrades the uranium cubic phase. The preservation of the metastable phase has considerable influence on fuel performance. The physical properties of the foil fuel under irradiation are superior to those of aluminum-dispersed fuels. The fuel meat, using zirconium as the diffusion barrier, prevents the interaction layer from becoming excessively thick. The problem with dispersed fuel is breakaway swelling with a medium fission rate. It has been observed that the fuel dispersed in aluminum was minimized in monolithic forms. The pure alloys exhibited a suitable response from a rate at least twice as much as the fission rate of dispersions. The foils can support fissile material concentration combined with a reduced swelling rate. (author)

  9. High density fuels using dispersion and monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br, E-mail: alfredo@ctmsp.mar.mil.br, E-mail: rafael.orm@gmail.com, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Departamento de Engenharia Naval e Oceânica

    2017-07-01

    Fuel plates used in high-performance research reactors need to be converted to low-enrichment uranium fuel; the fuel option based on a monolithic formulation requires alloys to contain 6 - 10 wt% Mo. In this case, the fuel plates are composed of the metallic alloy U-10Mo surrounded by a thin zirconium layer encapsulated in aluminum cladding. This study reviewed the physical properties of monolithic forms. The constraints produced during the manufacturing process were analyzed and compared to those of dispersed fuel. The bonding process used for dispersion fuels differs from the techniques applied to foil bonding used for pure alloys. The quality of monolithic plates depends on the fabrication method, which usually involves hot isostatic pressing and the thermal annealing effect of residual stress, which degrades the uranium cubic phase. The preservation of the metastable phase has considerable influence on fuel performance. The physical properties of the foil fuel under irradiation are superior to those of aluminum-dispersed fuels. The fuel meat, using zirconium as the diffusion barrier, prevents the interaction layer from becoming excessively thick. The problem with dispersed fuel is breakaway swelling with a medium fission rate. It has been observed that the fuel dispersed in aluminum was minimized in monolithic forms. The pure alloys exhibited a suitable response from a rate at least twice as much as the fission rate of dispersions. The foils can support fissile material concentration combined with a reduced swelling rate. (author)

  10. Drinking water purification by electrosynthesis of hydrogen peroxide in a power-producing PEM fuel cell.

    Science.gov (United States)

    Li, Winton; Bonakdarpour, Arman; Gyenge, Előd; Wilkinson, David P

    2013-11-01

    The industrial anthraquinone auto-oxidation process produces most of the world's supply of hydrogen peroxide. For applications that require small amounts of H2 O2 or have economically difficult transportation means, an alternate, on-site H2 O2 production method is needed. Advanced drinking water purification technologies use neutral-pH H2 O2 in combination with UV treatment to reach the desired water purity targets. To produce neutral H2 O2 on-site and on-demand for drinking water purification, the electroreduction of oxygen at the cathode of a proton exchange membrane (PEM) fuel cell operated in either electrolysis (power consuming) or fuel cell (power generating) mode could be a possible solution. The work presented here focuses on the H2 /O2 fuel cell mode to produce H2 O2 . The fuel cell reactor is operated with a continuous flow of carrier water through the cathode to remove the product H2 O2 . The impact of the cobalt-carbon composite cathode catalyst loading, Teflon content in the cathode gas diffusion layer, and cathode carrier water flowrate on the production of H2 O2 are examined. H2 O2 production rates of up to 200 μmol h(-1)  cmgeometric (-2) are achieved using a continuous flow of carrier water operating at 30 % current efficiency. Operation times of more than 24 h have shown consistent H2 O2 and power production, with no degradation of the cobalt catalyst. Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  11. APEX nuclear fuel cycle for production of LWR fuel and elimination of radioactive waste

    International Nuclear Information System (INIS)

    Steinberg, M.; Powell, J.R.

    1981-08-01

    The development of a nuclear fission fuel cycle is proposed which eliminates all the radioactive fission product waste effluent and the need for geological-age high level waste storage and provides a long term supply of fissile fuel for an LWR power reactor economy. The fuel cycle consists of reprocessing LWR spent fuel (1 to 2 years old) to remove the stable nonradioactive (NRFP, e.g. lanthanides, etc.) and short-lived fission products SLFP e.g. half-lives of (1 to 2 years) and returning, in dilute form, the long-lived fission products, ((LLFPs, e.g. 30 y half-life Cs, Sr, and 10 y Kr, and 16 x 10 6 y I) and the transuranics (TUs, e.g. Pu, Am, Cm, and Np) to be refabricated into fresh fuel elements. Makeup fertile and fissile fuel are to be supplied through the use of a Spallator (linear accelerator spallation-target fuel-producer). The reprocessing of LWR fuel elements is to be performed by means of the Chelox process which consists of Airox treatment (air oxidation and hydrogen reduction) followed by chelation with an organic reagent (β-diketonate) and vapor distillation of the organometallic compounds for separation and partitioning of the fission products

  12. Costs of electronuclear fuel production

    International Nuclear Information System (INIS)

    Flaim, T.; Loose, V.

    1978-07-01

    The Los Alamos Scientific Laboratory (LASL) proposes to study the electronuclear fuel producer (EFP) as a means of producing fissile fuel to generate electricity. The main advantage of the EFP is that it may reduce the risks of nuclear proliferation by breeding 233 U from thorium, thereby avoiding plutonium separation. A report on the costs of electronuclear fuel production based upon two designs considered by LASL is presented. The findings indicate that the EFP design variations considered are not likely to result in electricity generation costs as low as the uranium fuel cycle used in the US today. At current estimates of annual fuel output (500 kg 233 U per EFP), the costs of electricity generation using fuel produced by the EFP are more than three times higher than generating costs using the traditional fuel cycle. Sensitivity analysis indicates that electronuclear fuel production would become cost competitive with the traditional uranium fuel cycle when U 3 O 8 (yellowcake) prices approach $1000 per pound

  13. Technical Appraisal of Continuous Destilator Type as Alternative Fuel Producer from Basic Materials of Arak Bali

    Directory of Open Access Journals (Sweden)

    Sukadana -

    2012-11-01

    Full Text Available Arrack Bali which is produced from traditional process has low quality (<40%. With controlling of operational variable such as evaporation temperature, will improve arrack Bali quality. Arrack Bali with quality more than 80 % has octane number more than 108,6, higher then petroleum octane number (80 until 90, easy burning and evaporation, very good to be alternative fuel to engine. In order to product height quality any operational variables like temperature, step, and sprayer models should be noticed. This experiment is to obtain operational variables of distillatory to product arrack Bali as an alternative fuel and it is tested in motor cycle engine at speed and compression ratio variables toward performance like emission. The higher evaporation temperature is the higher capacity of product to be obtained, on the other hand, the lower quality to be reached. Generally, comparing with petroleum, arrack Bali yields lower emission.

  14. Acid leaching of coal: to produce clean fuels from Turkish lignite

    Energy Technology Data Exchange (ETDEWEB)

    Seferinoglu, Meryem [Mineral Research and Exploration Directorate (Turkey)], email: meryem_seferinoglu66@yahoo.com; Duzenli, Derya [Ankara Central Laboratory (Turkey)

    2011-07-01

    With the increasing concerns about the environment, energy producers and governments are looking at developing clean energy sources. However, Turkey has limited clean energy resources and is using low grade coal which has high sulphur content as an alternative energy source. The aim of this paper is to study the possibility of generating clean fuel from Edirne Lignite and to get a better understanding of chemical mechanisms involved in coal leaching with hydrofluoric acid (HF) solutions. Leaching was conducted on Edirne Lignite with HF solution at ambient temperature and the effects of parameters such as reaction time and concentration of acid solutions on the process were evaluated. The optimum conditions were found and it was shown that ash levels can be reduced from 28.9% to 10.5% and the calorific value increased by 500kcal/kg with the HF leaching method. This study demonstrated that the production of clean fuel from high sulphur lignite is possible.

  15. Update on Monitoring Technologies for International Safeguards and Fissile Material Verification

    International Nuclear Information System (INIS)

    Croessmann, C. Dennis; Glidewell, Don D.; Mangan, Dennis L.; Smathers, Douglas C.

    1999-01-01

    Monitoring technologies are playing an increasingly important part in international safeguards and fissile material verification. The developments reduce the time an inspector must spend at a site while assuring continuity of knowledge. Monitoring technologies' continued development has produced new seal systems and integrated video surveillance advances under consideration for Trilateral Initiative use. This paper will present recent developments for monitoring systems at Embalse, Argentina, VNHEF, Sarov, Russian, and Savannah River Site, Aiken, South Carolina

  16. Performance of nickel-based oxygen carrier produced using renewable fuel aloe vera

    Science.gov (United States)

    Afandi, NF; Devaraj, D.; Manap, A.; Ibrahim, N.

    2017-04-01

    Consuming and burning of fuel mainly fossil fuel has gradually increased in this upcoming era due to high-energy demand and causes the global warming. One of the most effective ways to reduce the greenhouse gases is by capturing carbon dioxide (CO2) during the combustion process. Chemical looping combustion (CLC) is one of the most effective methods to capture the CO2 without the need of an energy intensive air separation unit. This method uses oxygen carrier to provide O2 that can react with fuel to form CO2 and H2O. This research focuses on synthesizing NiO/NiAl2O4 as an oxygen carrier due to its properties that can withstand high temperature during CLC application. The NiO/NiAl2O4 powder was synthesized using solution combustion method with plant extract renewable fuel, aloe vera as the fuel. In order to optimize the performance of the particles that can be used in CLC application, various calcination temperatures were varied at 600°C, 800°C, 1050°C and 1300°C. The phase and morphology of obtained powders were characterized using X-ray diffraction (XRD) and Field Emission Microscopy (FESEM) respectively together with the powder elements. In CLC application, high reactivity can be achieved by using smaller particle size of oxygen carrier. This research succeeded in producing nano-structured powder with high crystalline structure at temperature 1050°C which is suitable to be used in CLC application.

  17. Fuel arrangement for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Tobin, J.M.

    1978-01-01

    Disclosed is a fuel arrangement for a high temperature gas cooled reactor including fuel assemblies with separate directly cooled fissile and fertile fuel elements removably inserted in an elongated moderator block also having a passageway for control elements

  18. Verification of a Fissile Material Cut-off Treaty (FMCT): The Potential Role of the IAEA

    International Nuclear Information System (INIS)

    Chung, Jin Ho

    2016-01-01

    The objective of a future verification of a FMCT(Fissile Material Cut-off Treaty) is to deter and detect non-compliance with treaty obligations in a timely and non-discriminatory manner with regard to banning the production of fissile material for nuclear weapons or other nuclear devices. Since the International Atomic Energy Agency (IAEA) has already established the IAEA safeguards as a verification system mainly for Non -Nuclear Weapon States (NNWSs), it is expected that the IAEA's experience and expertise in this field will make a significant contribution to setting up a future treaty's verification regime. This paper is designed to explore the potential role of the IAEA in verifying the future treaty by analyzing verification abilities of the Agency in terms of treaty verification and expected challenges. Furthermore, the concept of multilateral verification that could be facilitated by the IAEA will be examined as a measure of providing a credible assurance of compliance with a future treaty. In this circumstance, it is necessary for the IAEA to be prepared for playing a leading role in FMCT verifications as a form of multilateral verification by taking advantage of its existing verification concepts, methods, and tools. Also, several challenges that the Agency faces today need to be overcome, including dealing with sensitive and proliferative information, attribution of fissile materials, lack of verification experience in military fuel cycle facilities, and different attitude and culture towards verification between NWSs and NNWSs

  19. Verification of a Fissile Material Cut-off Treaty (FMCT): The Potential Role of the IAEA

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Jin Ho [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of a future verification of a FMCT(Fissile Material Cut-off Treaty) is to deter and detect non-compliance with treaty obligations in a timely and non-discriminatory manner with regard to banning the production of fissile material for nuclear weapons or other nuclear devices. Since the International Atomic Energy Agency (IAEA) has already established the IAEA safeguards as a verification system mainly for Non -Nuclear Weapon States (NNWSs), it is expected that the IAEA's experience and expertise in this field will make a significant contribution to setting up a future treaty's verification regime. This paper is designed to explore the potential role of the IAEA in verifying the future treaty by analyzing verification abilities of the Agency in terms of treaty verification and expected challenges. Furthermore, the concept of multilateral verification that could be facilitated by the IAEA will be examined as a measure of providing a credible assurance of compliance with a future treaty. In this circumstance, it is necessary for the IAEA to be prepared for playing a leading role in FMCT verifications as a form of multilateral verification by taking advantage of its existing verification concepts, methods, and tools. Also, several challenges that the Agency faces today need to be overcome, including dealing with sensitive and proliferative information, attribution of fissile materials, lack of verification experience in military fuel cycle facilities, and different attitude and culture towards verification between NWSs and NNWSs.

  20. Evaluation of simulated-LOCA tests that produced large fuel cladding ballooning

    International Nuclear Information System (INIS)

    Powers, D.A.; Meyer, R.O.

    1979-02-01

    A description is given of the NRC review and evaluation of simulated-LOCA tests that produced large axially extended ballooing in Zircaloy fuel cladding. Technical summaries are presented on the likelihood of the transient that was used in the tests, the effects of temperature variations on strain localization, and the results of other similar experiments. It is concluded that (a) the large axially extended deformations were an artifact of the experimental technique, (b) current NRC licensing positions are not invalidated by this new information, and (c) no new research programs are needed to study this phenomenon

  1. System and process for producing fuel with a methane thermochemical cycle

    Science.gov (United States)

    Diver, Richard B.

    2015-12-15

    A thermochemical process and system for producing fuel are provided. The thermochemical process includes reducing an oxygenated-hydrocarbon to form an alkane and using the alkane in a reforming reaction as a reducing agent for water, a reducing agent for carbon dioxide, or a combination thereof. Another thermochemical process includes reducing a metal oxide to form a reduced metal oxide, reducing an oxygenated-hydrocarbon with the reduced metal oxide to form an alkane, and using the alkane in a reforming reaction as a reducing agent for water, a reducing agent for carbon dioxide, or a combination thereof. The system includes a reformer configured to perform a thermochemical process.

  2. Evaluation of DD and DT fusion fuel cycles for different fusion-fission energy systems

    International Nuclear Information System (INIS)

    Gohar, Y.

    1980-01-01

    A study has been carried out in order to investigate the characteristics of an energy system to produce a new source of fissile fuel for existing fission reactors. The denatured fuel cycles were used because it gives additional proliferation resistance compared to other fuel cycles. DT and DD fusion drivers were examined in this study with a thorium or uranium blanket for each fusion driver. Various fuel cycles were studied for light-water and heavy-water reactors. The cost of electricity for each energy system was calculated

  3. Design of an extrusion screw and solid fuel produced from coconut shell

    Directory of Open Access Journals (Sweden)

    Madhiyanon, T

    2006-03-01

    Full Text Available The objectives were to design an extrusion screw to produce biomass solid fuel in a cold extrusion process, and investigate the effects of molasses used as a selected adhesive on the physical properties of extruded products. The material employed consisted of crushed coconut shell char and coconut fiber char mixed at a ratio of 40:60. The ratios of molasses in the mixture were 10:100, 15:100 and 20:100 (by weight and the extrusion die angles were 1.0, 1.1, 1.2, and 1.3 degrees gradation per experiment. The experimental results showed that the newly designed screw could function properly in the output range 0.75-0.90 kg/min, which is close to the design value. Regarding the molasses's effect on solid fuel properties, increasing the share of molasses was positive for both output and strength of the resulting briquettes, whereas the results of increasing die angle showed decreases in both output and strength. The compressive strength varied between 2.49-2.87 MPa in all circumstances, which was considerably higher than acceptable industrial level. Furthermore, the extruded solid fuel showed excellent resistance to impact force. Regarding energy consumption, the amount of electrical energy used in the extrusion process was insignificant, ranging between 0.040-0.079 kWh/kg.

  4. Nitrogen Isotope Composition of Thermally Produced NOx from Various Fossil-Fuel Combustion Sources.

    Science.gov (United States)

    Walters, Wendell W; Tharp, Bruce D; Fang, Huan; Kozak, Brian J; Michalski, Greg

    2015-10-06

    The nitrogen stable isotope composition of NOx (δ(15)N-NOx) may be a useful indicator for NOx source partitioning, which would help constrain NOx source contributions in nitrogen deposition studies. However, there is large uncertainty in the δ(15)N-NOx values for anthropogenic sources other than on-road vehicles and coal-fired energy generating units. To this end, this study presents a broad analysis of δ(15)N-NOx from several fossil-fuel combustion sources that includes: airplanes, gasoline-powered vehicles not equipped with a three-way catalytic converter, lawn equipment, utility vehicles, urban buses, semitrucks, residential gas furnaces, and natural-gas-fired power plants. A relatively large range of δ(15)N-NOx values was measured from -28.1‰ to 8.5‰ for individual exhaust/flue samples that generally tended to be negative due to the kinetic isotope effect associated with thermal NOx production. A negative correlation between NOx concentrations and δ(15)N-NOx for fossil-fuel combustion sources equipped with selective catalytic reducers was observed, suggesting that the catalytic reduction of NOx increases δ(15)N-NOx values relative to the NOx produced through fossil-fuel combustion processes. Combining the δ(15)N-NOx measured in this study with previous published values, a δ(15)N-NOx regional and seasonal isoscape was constructed for the contiguous U.S., which demonstrates seasonal and regional importance of various NOx sources.

  5. Nonintrusive verification attributes for excess fissile materials

    International Nuclear Information System (INIS)

    Nicholas, N.J.; Eccleston, G.W.; Fearey, B.L.

    1997-10-01

    Under US initiatives, over two hundred metric tons of fissile materials have been declared to be excess to national defense needs. These excess materials are in both classified and unclassified forms. The US has expressed the intent to place these materials under international inspections as soon as practicable. To support these commitments, members of the US technical community are examining a variety of nonintrusive approaches (i.e., those that would not reveal classified or sensitive information) for verification of a range of potential declarations for these classified and unclassified materials. The most troublesome and potentially difficult issues involve approaches for international inspection of classified materials. The primary focus of the work to date has been on the measurement of signatures of relevant materials attributes (e.g., element, identification number, isotopic ratios, etc.), especially those related to classified materials and items. The authors are examining potential attributes and related measurement technologies in the context of possible verification approaches. The paper will discuss the current status of these activities, including their development, assessment, and benchmarking status

  6. Linear accelerator fuel enricher regenerator (LAFER) and fission product transmutor (APEX)

    International Nuclear Information System (INIS)

    Steinberg, M.; Powell, J.R.; Takahashi, H.; Grand, P.; Kouts, H.J.C.

    1979-01-01

    In addition to safety, two other major problems face the nuclear industry today; first is the long-term supply of fissle material and second is the disposal of long-lived fission product waste. The higher energy proton linear accelerator can assist in the solution of each of these problems. High energy protons from the linear accelerator interact with a molten lead target to produce spallation and evaporation neutrons. The neutrons are absorbed in a surrounding blanket of light water power reactor (LWR) fuel elements to produce fissile Pu-239 or U-233 fuel from natural fertile U-238 or Th-232 contained in the elements. The fissile enriched fuel element is used in the LWR power reactor until its reactivity is reduced after which the element is regenerated in the linear accelerator target/blanket assembly and then the element is once again burned (fissioned) in the power LWR. In this manner the natural uranium fuel resource can supply an expanding nuclear power reactor economy without the need for fuel reprocessing, thus satisfying the US policy of non-proliferation. In addition, the quantity of spent fuel elements for long-term disposal is reduced in proportion to the number of fuel regeneration cycles through the accelerator. The limiting factor for in-situ regeneration is the burnup damage to the fuel cladding material. A 300 ma-1.5 GeV (450 MW) proton linear accelerator can produce approximately one ton of fissile (Pu-239) material annually which is enough to supply fuel to three 1000 MW(e) LWR power reactors. With two cycles of enriching and regenerating, the nuclear fuel natural resource can be stretched by a factor of 3.6 compared to present fuel cycle practice without the need for reprocessing. Furthermore, the need for isotopic enrichment facilities is drastically reduced

  7. 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.

  8. 40 CFR 80.620 - What are the additional requirements for diesel fuel or distillates produced by foreign...

    Science.gov (United States)

    2010-07-01

    ... audits of the foreign refinery. (i) Inspections and audits may be either announced in advance by EPA, or... diesel fuel or distillate was produced, assurance that the diesel fuel or distillate remained segregated...: (i) Be approved in advance by EPA, based on a demonstration of ability to perform the procedures...

  9. Fissile materials principles of criticality safety in handling and processing

    International Nuclear Information System (INIS)

    1976-01-01

    This Swedish Standard consists of the English version of the International Standard ISO 1709-1975-Nuclear energy. Fissile materials. Principles of criticality safety in handling and processing. (author)

  10. Bioelectrochemical Haber-Bosch Process: An Ammonia-Producing H2 /N2 Fuel Cell.

    Science.gov (United States)

    Milton, Ross D; Cai, Rong; Abdellaoui, Sofiene; Leech, Dónal; De Lacey, Antonio L; Pita, Marcos; Minteer, Shelley D

    2017-03-01

    Nitrogenases are the only enzymes known to reduce molecular nitrogen (N 2 ) to ammonia (NH 3 ). By using methyl viologen (N,N'-dimethyl-4,4'-bipyridinium) to shuttle electrons to nitrogenase, N 2 reduction to NH 3 can be mediated at an electrode surface. The coupling of this nitrogenase cathode with a bioanode that utilizes the enzyme hydrogenase to oxidize molecular hydrogen (H 2 ) results in an enzymatic fuel cell (EFC) that is able to produce NH 3 from H 2 and N 2 while simultaneously producing an electrical current. To demonstrate this, a charge of 60 mC was passed across H 2  /N 2 EFCs, which resulted in the formation of 286 nmol NH 3  mg -1 MoFe protein, corresponding to a Faradaic efficiency of 26.4 %. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  11. Economics of producing hydrogen as transportation fuel using offshore wind energy systems

    International Nuclear Information System (INIS)

    Mathur, Jyotirmay; Agarwal, Nalin; Swaroop, Rakesh; Shah, Nikhar

    2008-01-01

    Over the past few years, hydrogen has been recognized as a suitable substitute for present vehicular fuels. This paper covers the economic analysis of one of the most promising hydrogen production methods-using wind energy for producing hydrogen through electrolysis of seawater-with a concentration on the Indian transport sector. The analysis provides insights about several questions such as the advantages of offshore plants over coastal installations, economics of large wind-machine clusters, and comparison of cost of producing hydrogen with competing gasoline. Robustness of results has been checked by developing several scenarios such as fast/slow learning rates for wind systems for determining future trends. Results of this analysis show that use of hydrogen for transportation is not likely to be attractive before 2012, and that too with considerable learning in wind, electrolyzer and hydrogen storage technology

  12. Multi-unit Inertial Fusion Energy (IFE) plants producing hydrogen fuel

    International Nuclear Information System (INIS)

    Logan, B.G.

    1993-12-01

    A quantitative energy pathway comparison is made between a modern oil refinery and genetic fusion hydrogen plant supporting hybrid-electric cars powered by gasoline and hydrogen-optimized internal combustion engines, respectively, both meeting President Clinton's goal for advanced car goal of 80 mpg gasoline equivalent. The comparison shows that a fusion electric plant producing hydrogen by water electrolysis at 80% efficiency must have an electric capacity of 10 GWe to support as many hydrogen-powered hybrid cars as one modern 200,000 bbl/day-capacity oil refinery could support in gasoline-powered hybrid cars. A 10 GWe fusion electric plant capital cost is limited to 12.5 B$ to produce electricity at 2.3 cents/kWehr, and hydrogen production by electrolysis at 8 $/GJ, for equal consumer fuel cost per passenger mile as in the oil-gasoline-hybrid pathway

  13. Methods of refining natural oils, and methods of producing fuel compositions

    Science.gov (United States)

    Firth, Bruce E.; Kirk, Sharon E.

    2015-10-27

    A method of refining a natural oil includes: (a) providing a feedstock that includes a natural oil; (b) reacting the feedstock in the presence of a metathesis catalyst to form a metathesized product that includes olefins and esters; (c) passivating residual metathesis catalyst with an agent that comprises nitric acid; (d) separating the olefins in the metathesized product from the esters in the metathesized product; and (e) transesterifying the esters in the presence of an alcohol to form a transesterified product and/or hydrogenating the olefins to form a fully or partially saturated hydrogenated product. Methods for suppressing isomerization of olefin metathesis products produced in a metathesis reaction, and methods of producing fuel compositions are described.

  14. Lagooning microbial fuel cells: A first approach by coupling electricity-producing microorganisms and algae

    International Nuclear Information System (INIS)

    Lobato, Justo; González del Campo, Araceli; Fernández, Francisco J.; Cañizares, Pablo; Rodrigo, Manuel A.

    2013-01-01

    Highlights: • An algae cathode of a MFC has been used without artificial mediators or catalysts. • To perform a lagooning wastewater treatment coupled with energy-producing MFC. • The producing electricity operates under day/night irradiation cycles, is shown. - Abstract: The paper focused on the start-up and performance characterisation of a new type of microbial fuel cell (MFC), in which an algae culture was seeded in the cathodic chamber to produce the oxygen required to complete the electrochemical reactions of the MFC, thus circumventing the need for a mechanical aerator. The system did not use mediators or high cost catalysts and it can be started-up easily using a straightforward three-stage procedure. The start-up consists of the separate production of the electricity-producing microorganisms and the algae cultures (stage I), replacement of the mechanical aeration system by the algae culture (stage II) and a change in the light dosage from a continuous input to a dynamic day/night profile. The MFC was operated under a regime of 12 h light and 12 h dark and was also operated in batch and continuous substrate-feeding modes. The same cell voltage was achieved when the cathode compartment was operated with air supplied by aerators, which means that this configuration can perform as well as the traditional one. The results also show the influence of both the organic load and light irradiation on electricity production and demonstrate that this type MFC is a robust and promising technology that can be considered as a first approach to perform a lagooning wastewater treatment with microbial fuel cells

  15. 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

  16. Enhanced safety in the storage of fissile materials

    International Nuclear Information System (INIS)

    Williams, G.E.; Alvares, N.J.

    1979-01-01

    A ''plastic-like'' supporting material impregnated with a neutron-absorbing agent that is suitable for ''lining'' the inner surfaces of fissile-material storage containers was fabricated. The material consists, by weight, of 50% food-grade borax, 25% coal tar, and 25% epoxy resin. It costs much less than commercially available materials, can absorb enough neutrons to isolate units of fissile material, and possesses such structural qualities as flexibility and machinability. Properties and performance of the material are discussed

  17. Scenarios for multi-unit inertial fusion energy plants producing hydrogen fuel

    International Nuclear Information System (INIS)

    Logan, B.G.

    1993-12-01

    This work describes: (a) the motivation for considering fusion in general, and Inertial Fusion Energy (IFE) in particular, to produce hydrogen fuel powering low-emission vehicles; (b) the general requirements for any fusion electric plant to produce hydrogen by water electrolysis at costs competitive with present consumer gasoline fuel costs per passenger mile, for advanced car architectures meeting President Clinton's 80 mpg advanced car goal, and (c) a comparative economic analysis for the potential cost of electricity (CoE) and corresponding cost of hydrogen (CoH) from a variety of multi-unit IFE plants with one to eight target chambers sharing a common driver and target fab facility. Cases with either heavy-ion or diode-pumped, solid-state laser drivers are considered, with ''conventional'' indirect drive target gains versus ''advanced, e.g. Fast Ignitor'' direct drive gain assumptions, and with conventional steam balance-of-plant (BoP) versus advanced MHD plus steam combined cycle BoP, to contrast the potential economics under ''conventional'' and ''advanced'' IFE assumptions, respectively

  18. Methodology for interpretation of fissile mass flow measurements

    International Nuclear Information System (INIS)

    March-Leuba, J.; Mattingly, J.K.; Mullens, J.A.

    1997-01-01

    This paper describes a non-intrusive measurement technique to monitor the mass flow rate of fissile material in gaseous or liquid streams. This fissile mass flow monitoring system determines the fissile mass flow rate by relying on two independent measurements: (1) a time delay along a given length of pipe, which is inversely proportional to the fissile material flow velocity, and (2) an amplitude measurement, which is proportional to the fissile concentration (e.g., grams of 235 U per length of pipe). The development of this flow monitor was first funded by DOE/NE in September 95, and initial experimental demonstration by ORNL was described in the 37th INMM meeting held in July 1996. This methodology was chosen by DOE/NE for implementation in November 1996; it has been implemented in hardware/software and is ready for installation. This paper describes the methodology used to interpret the data measured by the fissile mass flow monitoring system and the models used to simulate the transport of fission fragments from the source location to the detectors

  19. Status and prospects of advanced fissile fuel breeders

    International Nuclear Information System (INIS)

    Kostoff, R.N.

    1979-01-01

    Fusion--fission hybrid systems, fast breeder systems, and accelerator breeder systems were compared on a common basis using a simple economic model. Electricity prices based on system capital costs only were computed, and were plotted as functions of five key breeder system parameters. Nominally, hybrid system electricity costs were about twenty-five percent lower than fast breeder system electricity costs, and fast breeder system electricity costs were about forty percent lower than accelerator breeder system electricity costs. In addition, hybrid system electricity costs were very insensitive to key parameter variations on the average, fast breeder system electricity costs were moderately sensitive to key parameter variations on the average, and accelerator breeder system electricity costs were the most sensitive to key parameter variations on the average

  20. Requirements for the transport of surplus fissile materials in the United States

    International Nuclear Information System (INIS)

    Wilson, R.K.

    1995-01-01

    This paper discusses the requirements and issues associated with the transportation of surplus fissile materials in the United States. The paper describes the materials that will be transported, the permissible modes of transport for these materials, and the safety and security requirements for each mode of transport. The paper also identifies transportation issues associated with these requirements, including the differences in requirements corresponding to who owns the material and whether the transport is on-site or off-site. Finally, the paper provides a discussion that suggests that by adopting the spent fuel standard and stored weapon standard proposed by the National Academy of Sciences, the requirements for transportation become straightforward

  1. IAEA technical meeting on fissile material strategies for sustainable nuclear energy

    International Nuclear Information System (INIS)

    Ganguly, Chaitanyamoy; Koyama, Kazutoshi

    2005-01-01

    A Technical Meeting (TM) on 'Fissile Material Management Strategies for Sustainable Nuclear Energy' was organized by the International Atomic Energy Agency (IAEA) in Vienna from 12 to 15 September 2005. Prior to the TM, three Working Groups (WG) composed of experts from 10 countries prepared Key Issues papers on: 1) Uranium Demand and Supply through 2050; 2) Back-end Fuel Cycle Options; and 3) Sustainable Nuclear Energy beyond 2050: Cross-cutting Issues. Some 36 papers, including 3 key issue papers, were presented during the TM in 3 different sessions. The present paper summarizes the deliberations of the TM. (author)

  2. Electric breeding of fissile materials with low Q, non-mainline fusion drivers

    International Nuclear Information System (INIS)

    Benford, J.; Bailey, V.; Oliver, D.; DiCapua, M.; Cooper, R.; Lopez, O.; Lindsey, H.

    1977-10-01

    The application of two novel fusion reactor concepts to the production of fissile fuel for existing and planned fission reactors has been shown to be technically feasible and potentially economically competitive. The performance required of fusion based breeders has been derived in terms of the fusion gain, blanket neutron and energy multiplication, and the performance and economic parameters of the fission reactors. Electron beam heated, linear solenoid confined plasmas were one concept which showed the most promise. A shock heated, wall confined reactor also appeared attractive for breeding

  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. Investigation on the performance and emission parameters of dual fuel diesel engine with mixture combination of hydrogen and producer gas as secondary fuel

    Directory of Open Access Journals (Sweden)

    A. E. Dhole

    2016-06-01

    Full Text Available This study presents experimental investigation in to the effects of using mixture of producer gas and hydrogen in five different proportions as a secondary fuel with diesel as pilot fuel at wide range of load conditions in dual fuel operation of a 4 cylinder turbocharged and intercooled 62.5 kW gen-set diesel engine at constant speed of 1500 RPM. Secondary fuel Substitution is in different percentage of diesel at each load. To generate producer gas, the rice husk was used as source in the downdraft gasifier. The performance and emission characteristics of the dual fuel engine are compared with that of diesel engine at different load conditions. It was found that of all the combinations tested, mixture combination of PG:H2=(60:40% is the most suited one at which the brake thermal efficiency is in good comparison to that of diesel operation. Decreased NOx emissions and increased CO emissions were observed for dual fuel mode for all the fuel combinations compared to diesel fuel operation.

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Wataumi, Kazutoshi; Tajiri, Hiroshi.

    1992-01-01

    In a fuel assembly of a BWR type reactor, a pellet to be loaded comprises an external layer of fissile materials containing burnable poisons and an internal layer of fissile materials not containing burnable poison. For example, there is provided a dual type pellet comprising an external layer made of UO 2 incorporated with Gd 2 O 3 at a predetermined concentration as the burnable poisons and an internal layer made of UO 2 not containing Gd 2 O 3 . The amount of the burnable poisons required for predetermined places is controlled by the thickness of the ring of the external layer. This can dissipate an unnecessary poisoning effect at the final stage of the combustion cycle. Further, since only one or a few kinds of powder mixture of the burnable poisons and the fissile materials is necessary, production and product control can be facilitated. (I.N.)

  6. Spallator and APEX nuclear fuel cycle: a new option for nuclear power

    International Nuclear Information System (INIS)

    Steinberg, M.

    1982-01-01

    A new nuclear fuel cycle is described which provides a long term supply of nuclear fuel for the thermal LWR nuclear power reactors and eliminates the need for long-term storage of radioactive waste. Fissile fuel is produced by the Spallator which depends on the production of spallation neutrons by the interaction of high-energy (1 to 2 GeV) protons on a heavy-metal target. The neutrons are absorbed in a surrounding natural-uranium or thorium blanket in which fissile Pu-239 to U-233 is produced. Advances in linear accelerator technology makes it possible to design and construct a high-beam-current continuous-wave proton linac for production purposes. The target is similar to a sub-critical reactor and produces heat which is converted to electricity for supplying the linac. The Spallator is a self-sufficient fuel producer, which can compete with the fast breeder. The APEX fuel cycle depends on recycling the transuranics and long-lived fission products while extracting the stable and short-lived fission products when reprocessing the fuel. Transmutation and decay within the fuel cycle and decay of short-lived fission products external to the fuel cycle eliminates the need for long-term geological age shortage of fission-product waste

  7. Spallator and APEX nuclear fuel cycle: a new option for nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Steinberg, M.

    1982-01-01

    A new nuclear fuel cycle is described which provides a long term supply of nuclear fuel for the thermal LWR nuclear power reactors and eliminates the need for long-term storage of radioactive waste. Fissile fuel is produced by the Spallator which depends on the production of spallation neutrons by the interaction of high-energy (1 to 2 GeV) protons on a heavy-metal target. The neutrons are absorbed in a surrounding natural-uranium or thorium blanket in which fissile Pu-239 to U-233 is produced. Advances in linear accelerator technology makes it possible to design and construct a high-beam-current continuous-wave proton linac for production purposes. The target is similar to a sub-critical reactor and produces heat which is converted to electricity for supplying the linac. The Spallator is a self-sufficient fuel producer, which can compete with the fast breeder. The APEX fuel cycle depends on recycling the transuranics and long-lived fission products while extracting the stable and short-lived fission products when reprocessing the fuel. Transmutation and decay within the fuel cycle and decay of short-lived fission products external to the fuel cycle eliminates the need for long-term geological age shortage of fission-product waste.

  8. Study of DD versus DT fusion fuel cycles for different fusion-fission hybrid energy systems

    International Nuclear Information System (INIS)

    Gohar, Y.; Baker, C.C.

    1981-01-01

    A study was performed to investigate the characteristics of an energy system to produce fissile fuel for fission reactors. DD and DT fusion reactors were examined in this study with either a thorium or uranium blanket for each fusion reactor. Various fuel cycles were examined for light-water reactors including the denatured fuel cycles (which may offer proliferation resistance compared to other fuel cycles); these fuel cycles include a uranium fuel cycle with 239 Pu makeup, a thorium fuel cycle with 239 Pu makeup, a denatured uranium fuel cycle with 233 U makeup, and a denatured thorium fuel cycle with 233 U makeup. Four different blankets were considered for this study. The first two blankets have a tritium breeding capability for DT reactors. Lithium oxide (Li 2 O) was used for tritium breeding due to its high lithium density and high temperature capability; however, the use of Li 2 O may result in higher tritium inventories compared to other solid breeders

  9. Generation of low-Btu fuel gas from agricultural residues experiments with a laboratory scale gas producer

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R O

    1977-01-01

    Two successive laboratory-scale, downdraft gas producers were fabricated and tested. Agricultural and food processing residues including walnut shells, corn cobs, tree prunings, and cotton gin waste, were converted to a low Btu producer gas. The performance of 2 spark ignition engines, when running on producer gas, was highly satisfactory. The ability of the producer to maintain a continuous supply of good quality gas was determined largely by firebox configuration. Fuel handling and fuel flow control problems tended to be specific to individual types of residues. During each test run, air input, firebox temperature, fuel consumption rate, and pressure differential across the producer were monitored. An overall conversion efficiency of 65% was achieved.

  10. Mass, energy and material balances of SRF production process. Part 3: solid recovered fuel produced from municipal solid waste.

    Science.gov (United States)

    Nasrullah, Muhammad; Vainikka, Pasi; Hannula, Janne; Hurme, Markku; Kärki, Janne

    2015-02-01

    This is the third and final part of the three-part article written to describe the mass, energy and material balances of the solid recovered fuel production process produced from various types of waste streams through mechanical treatment. This article focused the production of solid recovered fuel from municipal solid waste. The stream of municipal solid waste used here as an input waste material to produce solid recovered fuel is energy waste collected from households of municipality. This article presents the mass, energy and material balances of the solid recovered fuel production process. These balances are based on the proximate as well as the ultimate analysis and the composition determination of various streams of material produced in a solid recovered fuel production plant. All the process streams are sampled and treated according to CEN standard methods for solid recovered fuel. The results of the mass balance of the solid recovered fuel production process showed that 72% of the input waste material was recovered in the form of solid recovered fuel; 2.6% as ferrous metal, 0.4% as non-ferrous metal, 11% was sorted as rejects material, 12% as fine faction and 2% as heavy fraction. The energy balance of the solid recovered fuel production process showed that 86% of the total input energy content of input waste material was recovered in the form of solid recovered fuel. The remaining percentage (14%) of the input energy was split into the streams of reject material, fine fraction and heavy fraction. The material balances of this process showed that mass fraction of paper and cardboard, plastic (soft) and wood recovered in the solid recovered fuel stream was 88%, 85% and 90%, respectively, of their input mass. A high mass fraction of rubber material, plastic (PVC-plastic) and inert (stone/rock and glass particles) was found in the reject material stream. © The Author(s) 2014.

  11. Calculated nuclide production yields in relativistic collisions of fissile nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Benlliure, J.; Schmidt, K.H. [Gesellschaft fuer Schwerionenforschung mbH, Darmstadt (Germany); Grewe, A.; Jong, M. de [Technische Univ. Darmstadt (Germany). Inst. fuer Kernphysik; Zhdanov, S. [AN Kazakhskoj SSR, Alma-Ata (USSR). Inst. Yadernoj Fiziki

    1997-11-01

    A model calculation is presented which predicts the complex nuclide distribution resulting from peripheral relativistic heavy-ion collisions involving fissile nuclei. The model is based on a modern version of the abrasion-ablation model which describes the formation of excited prefragments due to the nuclear collisions and their consecutive decay. The competition between the evaporation of different light particles and fission is computed with an evaporation code which takes dissipative effects and the emission of intermediate-mass fragments into account. The nuclide distribution resulting from fission processes is treated by a semiempirical description which includes the excitation-energy dependent influence of nuclear shell effects and pairing correlations. The calculations of collisions between {sup 238}U and different reaction partners reveal that a huge number of isotopes of all elements up to uranium is produced. The complex nuclide distribution shows the characteristics of fragmentation, mass-asymmetric low-energy fission and mass-symmetric high-energy fission. The yields of the different components for different reaction partners are studied. Consequences for technical applications are discussed. (orig.)

  12. Characterisation of ashes produced by co-combustion of recovered fuels and peat

    Energy Technology Data Exchange (ETDEWEB)

    Frankenhaeuser, M. [Borealis Polymers Oy, Porvoo (Finland)

    1997-10-01

    The current project focuses on eventual changes in ash characteristics during co-combustion of refuse derived fuel with coal, peat, wood or bark, which could lead to slagging, fouling and corrosion in the boiler. Ashes were produced at fluidised bed (FB) combustion conditions in the 15 kW reactor at VTT Energy, Jyvaeskylae, the fly ash captured by the cyclone was further analysed by XRF at Outokumpu Geotechnical Laboratory, Outokumpu. The sintering behaviour of these ashes was investigated using a test procedure developed at the Combustion Chemistry Research Group at Aabo Akademi University. The current extended programme includes a Danish refuse-derived fuel (RDF), co-combusted with bark/coal (5 tests) and wood/coal (2 tests), a RF from Jyvaskyla (2 tests with peat/coal) and de-inking sludges co- combusted at full-scale with wood waste or paper mill sludge (4 ashes provided by IVO Power). Ash pellets were thermally treated in nitrogen in order to avoid residual carbon combustion. The results obtained show no sintering tendencies below 600 deg C, significant changes in sintering are seen with pellets treated at 1000 deg C. Ash from 100 % RDF combustion does not sinter, 25 % RDF co-combustion with wood and peat, respectively, gives an insignificant effect. The most severe sintering occurs during co-combustion of RDF with bark. Contrary to the earlier hypothesis a 25 % coal addition seems to have a negative effect on all fuel blends. Analysis of the sintering results versus ash chemical composition shows, that (again), in general, an increased level of alkali chlorides and sulphates gives increased sintering. Finally, some results on sintering tendency measurements on ashes from full-scale CFB co-combustion of deinking sludge with wood waste and paper mill sludge are given. This shows that these ashes show very little, if any, sintering tendency, which can be explained from ash chemistry

  13. Potential for producing bio-fuel in the Amazon deforested areas

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Ricardo Cunha da [Banco Nacional de Desenvolvimento Economico e Social (BNDES), Rio de Janeiro, RJ (Brazil)

    2004-05-01

    This paper analyzes the possibility of producing bio-fuel in the Amazon degraded lands. The aim here is to combine environmental concerns with an improvement of local people well-being. Firstly, a historical analysis is conducted in order to figure out the major deforestation driving forces in Amazon and to help to arrive at a feasible energy choice. Secondly, the geographical area is chosen. It is the spatial boundaries of Carajas Iron Ore Program in the southeastern Amazon where most of the deforestation has taken place in the last few decades. For this specific context, palm oil is chosen as a technological energy alternative due to its social production structure, its environmental benefits and its productivity . A quantified analysis is realized in terms of income generation (2000-3000 US dollars/family/yr), job creation (200,000-300,000 families settled), land required and restored (2-3.2 million ha), and carbon emission from fossil fuel avoided (13.1 Mt C). Some recommendations related to institutional and economic barriers are proposed in order to encourage the technology penetration in the market. (Author)

  14. A review of catalysts for the electroreduction of carbon dioxide to produce low-carbon fuels.

    Science.gov (United States)

    Qiao, Jinli; Liu, Yuyu; Hong, Feng; Zhang, Jiujun

    2014-01-21

    This paper reviews recent progress made in identifying electrocatalysts for carbon dioxide (CO2) reduction to produce low-carbon fuels, including CO, HCOOH/HCOO(-), CH2O, CH4, H2C2O4/HC2O4(-), C2H4, CH3OH, CH3CH2OH and others. The electrocatalysts are classified into several categories, including metals, metal alloys, metal oxides, metal complexes, polymers/clusters, enzymes and organic molecules. The catalyts' activity, product selectivity, Faradaic efficiency, catalytic stability and reduction mechanisms during CO2 electroreduction have received detailed treatment. In particular, we review the effects of electrode potential, solution-electrolyte type and composition, temperature, pressure, and other conditions on these catalyst properties. The challenges in achieving highly active and stable CO2 reduction electrocatalysts are analyzed, and several research directions for practical applications are proposed, with the aim of mitigating performance degradation, overcoming additional challenges, and facilitating research and development in this area.

  15. Characterisation of ashes produced by co-combustion of recovered fuels and peat

    Energy Technology Data Exchange (ETDEWEB)

    Frankenhaeuser, M.; Zevenhoven, R. [Borealis Polymers Oy, Porvoo (Finland); Skrifvars, B.J. [Aabo Akademi, Turku (Finland); Orjala, M. [VTT Energy, Espoo (Finland); Peltola, K. [Foster Wheeler Energy (Finland)

    1996-12-01

    Source separation of combustible materials from household or municipal solid waste yields a raw material for the production of Packaging Derived Fuel (PDF). This fuel can substitute the traditional fuels in heat and power generation and is also called recycled fuel. Co-combustion of these types of fuels with coal has been studied in several LIEKKI-projects and the results have been both technically and environmentally favourable. (author)

  16. Bacillus spp. produce antibacterial activities against lactic acid bacteria that contaminate fuel ethanol plants.

    Science.gov (United States)

    Manitchotpisit, Pennapa; Bischoff, Kenneth M; Price, Neil P J; Leathers, Timothy D

    2013-05-01

    Lactic acid bacteria (LAB) frequently contaminate commercial fuel ethanol fermentations, reducing yields and decreasing profitability of biofuel production. Microorganisms from environmental sources in different geographic regions of Thailand were tested for antibacterial activity against LAB. Four bacterial strains, designated as ALT3A, ALT3B, ALT17, and MR1, produced inhibitory effects on growth of LAB. Sequencing of rRNA identified these strains as species of Bacillus subtilis (ALT3A and ALT3B) and B. cereus (ALT17 and MR1). Cell mass from colonies and agar samples from inhibition zones were analyzed by matrix-assisted laser desorption/ionization-time of flight mass spectrometry. The spectra of ALT3A and ALT3B showed a strong signal at m/z 1,060, similar in mass to the surfactin family of antimicrobial lipopeptides. ALT3A and ALT3B were analyzed by zymogram analysis using SDS-PAGE gels placed on agar plates inoculated with LAB. Cell lysates possessed an inhibitory protein of less than 10 kDa, consistent with the production of an antibacterial lipopeptide. Mass spectra of ALT17 and MR1 had notable signals at m/z 908 and 930 in the whole cell extracts and at m/z 687 in agar, but these masses do not correlate with those of previously reported antibacterial lipopeptides, and no antibacterial activity was detected by zymogram. The antibacterial activities produced by these strains may have application in the fuel ethanol industry as an alternative to antibiotics for prevention and control of bacterial contamination.

  17. Microscopic Fuel Particles Produced by Self-Assembly of Actinide Nanoclusters on Carbon Nanomaterials

    Energy Technology Data Exchange (ETDEWEB)

    Na, Chongzheng [Univ. of Notre Dame, IN (United States)

    2016-10-17

    Many consider further development of nuclear power to be essential for sustained development of society; however, the fuel forms currently used are expensive to recycle. In this project, we sought to create the knowledge and knowhow that are needed to produce nanocomposite materials by directly depositing uranium nanoclusters on networks of carbon-­ based nanomaterials. The objectives of the proposed work were to (1) determine the control of uranium nanocluster surface chemistry on nanocomposite formation, (2) determine the control of carbon nanomaterial surface chemistry on nanocomposite formation, and (3) develop protocols for synthesizing uranium-­carbon nanomaterials. After examining a wide variety of synthetic methods, we show that synthesizing graphene-­supported UO2 nanocrystals in polar ethylene glycol compounds by polyol reduction under boiling reflux can enable the use of an inexpensive graphene precursor graphene oxide in the production of uranium-carbon nanocomposites in a one-­pot process. We further show that triethylene glycol is the most suitable solvent for producing nanometer-­sized UO2 crystals compared to monoethylene glycol, diethylene glycol, and polyethylene glycol. Graphene-­supported UO2 nanocrystals synthesized with triethylene glycol show evidence of heteroepitaxy, which can be beneficial for facilitating heat transfer in nuclear fuel particles. Furthermore, we show that graphene-supported UO2 nanocrystals synthesized by polyol reduction can be readily stored in alcohols, preventing oxidation from the prevalent oxygen in air. Together, these methods provide a facile approach for preparing and storing graphene-supported UO nanocrystals for further investigation and development under ambient conditions.

  18. Thorium fuel-cycle development through plutonium incineration by THORIMS-NES (Thorium Molten-Salt nuclear energy synergetics)

    International Nuclear Information System (INIS)

    Furukawa, K.; Furuhashi, A.; Chigrinov, S.E.

    1996-01-01

    Thorium fuel-cycle has benefit on not-only trans-U element reduction but also their incineration. The disadvantage of high gamma activity of fuel, which is useful for improving the resistance to nuclear proliferation and terrorism, can overcome by molten fluorides fuel, and practically by THORIMS-NES, symbiotically coupled with fission Molten-Salt Reactor (FUJI) and fissile-producing Accelerator Molten-Salt Breeder (AMSB). This will have wide excellent advantages in global application, and will be deployed by incinerating Pu and Producing 233 U. Some details of this strategy including time schedule are presented. 14 refs, 2 figs, 4 tabs

  19. Characterization of biomass producer gas as fuel for stationary gas engines in combined heat and power production

    DEFF Research Database (Denmark)

    Ahrenfeldt, Jesper

    2008-01-01

    The aim of this project has been the characterization of biomass producer gas as a fuel for stationary gas engines in heat and power production. More than 3200 hours of gas engine operation, with producer gas as fuel, has been conducted at the biomass gasification combined heat and power (CHP...... different measuring methods. Likewise, no particles were detected in the gas. Considerable amounts of NH3 were measured in the produced gas.An analysis of engine operation at varying load has been carried out. Standard emissions, load and efficiency have been measured at varying operating conditions ranging...... from 50% to 90% load. Biomass producer gas is an excellent lean burn engine fuel: Operation of a natural aspirated engine has been achieved for 1.2...

  20. Research reactors fuel cycle problems and dilemma

    International Nuclear Information System (INIS)

    Romano, R.

    2004-01-01

    During last 10 years, some problems appeared in different steps of research reactors fuel cycle. Actually the majority of these reactors have been built in the 60s and these reactors were operated during all this long period in a cycle with steps which were dedicated to this activity. Progressively and for reasons often economical, certain steps of the cycle became more and more difficult to manage due to closing of some specialised workshops in the activities of scraps recycling, irradiated fuel reprocessing, even fuel fabrication. Other steps of the cycle meet or will meet difficulties, in particular supplying of fissile raw material LEU or HEU because this material was mostly produced in enrichment units existing mainly for military reason. Rarefaction of fissile material lead to use more and more enriched uraniums said 'of technical quality', that is to say which come from mixing of varied qualities of enriched material, containing products resulting from reprocessing. Actually, problems of end of fuel cycle are increased, either consisting of intermediary storage on the site of reactor or on specialised sites, or consisting of reprocessing. This brief summary shows most difficulties which are met today by a major part of industrials of the fuel cycle in the exercise of their activities

  1. Studying the effect of compression ratio on an engine fueled with waste oil produced biodiesel/diesel fuel

    Directory of Open Access Journals (Sweden)

    Mohammed EL_Kassaby

    2013-03-01

    Full Text Available Wasted cooking oil from restaurants was used to produce neat (pure biodiesel through transesterification, and then used to prepare biodiesel/diesel blends. The effect of blending ratio and compression ratio on a diesel engine performance has been investigated. Emission and combustion characteristics was studded when the engine operated using the different blends (B10, B20, B30, and B50 and normal diesel fuel (B0 as well as when varying the compression ratio from 14 to 16 to 18. The result shows that the engine torque for all blends increases as the compression ratio increases. The bsfc for all blends decreases as the compression ratio increases and at all compression ratios bsfc remains higher for the higher blends as the biodiesel percent increase. The change of compression ratio from 14 to 18 resulted in, 18.39%, 27.48%, 18.5%, and 19.82% increase in brake thermal efficiency in case of B10, B20, B30, and B50 respectively. On an average, the CO2 emission increased by 14.28%, the HC emission reduced by 52%, CO emission reduced by 37.5% and NOx emission increased by 36.84% when compression ratio was increased from 14 to 18. In spite of the slightly higher viscosity and lower volatility of biodiesel, the ignition delay seems to be lower for biodiesel than for diesel. On average, the delay period decreased by 13.95% when compression ratio was increased from 14 to 18. From this study, increasing the compression ratio had more benefits with biodiesel than that with pure diesel.

  2. Hot impact densification (HID) - a new method of producing ceramic nuclear fuel pellets with tight dimensional tolerances

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.; Muehling, G.; Vollath, D.; Zimmermann, H.

    1984-01-01

    The hot impact densification (HID) is a new powerful method for producing ceramic fuel pellets for nuclear reactors. Green ceramic bodies are directly processed to pellets by high speed shaping in the plastic temperature region of ceramic material. Opposed to the well established press sintering procedure it can be heated, densified, and cooled by orders of magnitude faster. Therefore, at high throughputs, small equipment dimensions become possible. The fuel pellets produced meet all requirements, particular the dimensional tolerances achieved are very closed, consequently circular grinding is omitted. Furthermore, the relatively high temperature level of the impact pressing favors the mixed crystal formation of uranium and plutonium oxide. This improves the solubility of the fuel in nitric acid, an essential point at reprocessing. A prototype facility is designed so that automatic fabrication in continuous operation will be possible. The target working cycle for a fuel pellet is in the range of some seconds. (orig.)

  3. Plant for producing an oxygen-containing additive as an ecologically beneficial component for liquid motor fuels

    Science.gov (United States)

    Siryk, Yury Paul; Balytski, Ivan Peter; Korolyov, Volodymyr George; Klishyn, Olexiy Nick; Lnianiy, Vitaly Nick; Lyakh, Yury Alex; Rogulin, Victor Valery

    2013-04-30

    A plant for producing an oxygen-containing additive for liquid motor fuels comprises an anaerobic fermentation vessel, a gasholder, a system for removal of sulphuretted hydrogen, and a hotwell. The plant further comprises an aerobic fermentation vessel, a device for liquid substance pumping, a device for liquid aeration with an oxygen-containing gas, a removal system of solid mass residue after fermentation, a gas distribution device; a device for heavy gases utilization; a device for ammonia adsorption by water; a liquid-gas mixer; a cavity mixer, a system that serves superficial active and dispersant matters and a cooler; all of these being connected to each other by pipelines. The technical result being the implementation of a process for producing an oxygen containing additive, which after being added to liquid motor fuels, provides an ecologically beneficial component for motor fuels by ensuring the stability of composition fuel properties during long-term storage.

  4. Fuel properties of biodiesel produced from the crude fish oil from the soapstock of marine fish

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Cherng-Yuan; Li, Rong-Ji [Department of Marine Engineering, National Taiwan Ocean, University, Keelung 20224 (China)

    2009-01-15

    The soapstock of a mixture of marine fish was used as the raw material to produce the biodiesel in this study. The soapstock was collected from discarded fish products. Crude fish oil was squeezed from the soapstock of the fish and refined by a series of processes. The refined fish oil was transesterified to produce biodiesel. The fuel properties of the biodiesel were analyzed. The experimental results showed that oleic acid (C18:1) and palmitic acid (C16:0) were the two major components of the marine fish-oil biodiesel. The biodiesel from the mixed marine fish oil contained a significantly greater amount of polyunsaturated fatty acids than did the biodiesel from waste cooking oil. In addition, the marine fish-oil biodiesel contained as high as 37.07 wt.% saturated fatty acids and 37.3 wt.% long chain fatty acids in the range between C20 and C22. Moreover, the marine fish-oil biodiesel appeared to have a larger acid number, a greater increase in the rate of peroxidization with the increase in the time that it was stored, greater kinematic viscosity, higher heating value, higher cetane index, more carbon residue, and a lower peroxide value, flash point, and distillation temperature than those of waste cooking-oil biodiesel. (author)

  5. Micro Fine Sized Palm Oil Fuel Ash Produced Using a Wind Tunnel Production System

    Directory of Open Access Journals (Sweden)

    R. Ahmadi

    2016-01-01

    Full Text Available Micro fine sized palm oil fuel ash (POFA is a new supplementary cementitious material that can increase the strength, durability, and workability of concrete. However, production of this material incurs high cost and is not practical for the construction industry. This paper investigates a simple methodology of producing micro fine sized POFA by means of a laboratory scale wind tunnel system. The raw POFA obtained from an oil palm factory is first calcined to remove carbon residue and then grinded in Los Angeles abrasion machine. The grinded POFA is then blown in the fabricated wind tunnel system for separation into different ranges of particle sizes. The physical, morphological, and chemical properties of the micro fine sized POFA were then investigated using Laser Particle Size Analyser (PSA, nitrogen sorption, and Scanning Electron Microscopy with Energy Dispersive X-Ray (SEM-EDX. A total of 32.1% micro fine sized POFA were collected from each sample blown, with the size range of 1–10 micrometers. The devised laboratory scale of wind tunnel production system is successful in producing micro fine sized POFA and, with modifications, this system is envisaged applicable to be used to commercialize micro fine sized POFA production for the construction industry.

  6. Method of producing exfoliated graphite composite compositions for fuel cell flow field plates

    Energy Technology Data Exchange (ETDEWEB)

    Zhamu, Aruna; Shi, Jinjun; Guo, Jiusheng; Jang, Bor Z

    2014-04-08

    A method of producing an electrically conductive composite composition, which is particularly useful for fuel cell bipolar plate applications. The method comprises: (a) providing a supply of expandable graphite powder; (b) providing a supply of a non-expandable powder component comprising a binder or matrix material; (c) blending the expandable graphite with the non-expandable powder component to form a powder mixture wherein the non-expandable powder component is in the amount of between 3% and 60% by weight based on the total weight of the powder mixture; (d) exposing the powder mixture to a temperature sufficient for exfoliating the expandable graphite to obtain a compressible mixture comprising expanded graphite worms and the non-expandable component; (e) compressing the compressible mixture at a pressure within the range of from about 5 psi to about 50,000 psi in predetermined directions into predetermined forms of cohered graphite composite compact; and (f) treating the so-formed cohered graphite composite to activate the binder or matrix material thereby promoting adhesion within the compact to produce the desired composite composition. Preferably, the non-expandable powder component further comprises an isotropy-promoting agent such as non-expandable graphite particles. Further preferably, step (e) comprises compressing the mixture in at least two directions. The method leads to composite plates with exceptionally high thickness-direction electrical conductivity.

  7. Biomass utilization for green environment: Co-combustion of diesel fuel and producer gas in thermal application

    International Nuclear Information System (INIS)

    Hussain, A.; Ani, F.N.; Mehamed, A.F.

    2007-01-01

    Study of co-combustion of diesel oil and producer gas from a gasifier, individually as well as combined, in an experimental combustion chamber revealed that the producer gas can be co-combusted with liquid fuel. The process produced more CO, NO/sub x/, SO/sub 2/ and CO/sub 2/ as compared to the combustion of diesel oil alone; the exhaust temperature for the process was higher than the diesel combustion alone. (author)

  8. Design, Fabrication, and Operation of Innovative Microalgae Culture Experiments for the Purpose of Producing Fuels: Final Report, Phase I

    Energy Technology Data Exchange (ETDEWEB)

    1985-01-01

    A conceptual design was developed for a 1000-acre (water surface) algae culture facility for the production of fuels. The system is modeled after the shallow raceway system with mixing foils that is now being operated at the University of Hawaii. A computer economic model was created to calculate the discounted breakeven price of algae or fuels produced by the culture facility. A sensitivity analysis was done to estimate the impact of changes in important biological, engineering, and financial parameters on product price.

  9. Requirements for materials of dispersion fuel elements

    International Nuclear Information System (INIS)

    Samojlov, A.G.; Kashtanov, A.I.; Volkov, V.S.

    1982-01-01

    Requirements for materials of dispersion fuel elements are considered. The necessity of structural and fissile materials compatibility at maximum permissible operation temperatures and temperatures arising in a fuel element during manufacture is pointed out. The fuel element structural material must be ductile, possess high mechanical strength minimum neutron absorption cross section, sufficient heat conductivity, good corrosion resistance in a coolant and radiation resistance. The fissile material must have high fissile isotope concentration, radiation resistance, high thermal conductivity, certain porosity high melting temperature must not change the composition under irradiation

  10. Signatures of Extended Storage of Used Nuclear Fuel Comprehensive Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-21

    This report serves as a comprehensive overview of the Extended Storage of Used Nuclear Fuel work performed for the Material Protection, Accounting and Control Technologies campaign under the Department of Energy Office of Nuclear Energy. This paper describes a signature based on the source and fissile material distribution found within a population of used fuel assemblies combined with the neutron absorbers found within cask design that is unique to a specific cask with its specific arrangement of fuel. The paper describes all the steps used in producing and analyzing this signature from the beginning to the project end.

  11. Enhanced safety in the storage of fissile materials

    International Nuclear Information System (INIS)

    Williams, G.E.; Alvares, N.J.

    1978-01-01

    An inexpensive boron-loaded liner of epoxy resin for fissile-material storage containers was developed that can be easily fabricated of readily available, low-cost materials. Computer calculations indicate reactivity will be reduced substantially if this neutron-absorbing liner is added to containers in a typical storage array. These calculations compare favorably with neutron-attenuation experiments with thermal and fission neutron spectra, and tests at the Fire Test Facility indicate the epoxy resin will survive extreme environmental and accident conditions. The fire-resistant and insulating properties of the epoxy-resin liner further augment its ability to protect fissile materials. Boron-loaded epoxy resin is adaptable to many tasks but is particularly useful for providing enhanced criticality safety in the packaging and storage of fissile materials

  12. Accelerating fissile material detection with a neutron source

    Science.gov (United States)

    Rowland, Mark S.; Snyderman, Neal J.

    2018-01-30

    A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly to count neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source. The system includes a Poisson neutron generator for in-beam interrogation of a possible fissile neutron source and a DC power supply that exhibits electrical ripple on the order of less than one part per million. Certain voltage multiplier circuits, such as Cockroft-Walton voltage multipliers, are used to enhance the effective of series resistor-inductor circuits components to reduce the ripple associated with traditional AC rectified, high voltage DC power supplies.

  13. A line of defense approach to fissile material control

    International Nuclear Information System (INIS)

    Holloway, S.P.; Holloway, N.J.

    1995-01-01

    A crucial element of the safety policy of the UK Atomic Weapons Establishment (AWE) is that concerned with the safe control of fissile material in order to minimize the potential for unplanned criticality. The principles by which AWE controls fissile material advocate a simple Line of Defense (LOD) approach to assessing criticality-safety related aspects of fissile operations. An LOD assessment provides a measure of the depth of defense available to prevent general types of criticality accident and can be used to demonstrate compliance with the risk-based Basic Safety Limits (BSLs) and Objectives (BSOs) used by the UK Nuclear Installations Inspectorate (NII) to judge the safety of operations in accordance with its Safety Assessment Principles (SAPs) for Nuclear Plants. This paper discusses the LOD concept, the basis of LOD assessment and describes LODs specific to criticality control

  14. Application of proton exchange membrane fuel cells for the monitoring and direct usage of biohydrogen produced by Chlamydomonas reinhardtii

    Energy Technology Data Exchange (ETDEWEB)

    Oncel, S.; Vardar-Sukan, F. [Department of Bioengineering, Faculty of Engineering, Ege University, 35100 Bornova, Izmir (Turkey)

    2011-01-01

    Photo-biologically produced hydrogen by Chlamydomonas reinhardtii is integrated with a proton exchange (PEM) fuel cell for online electricity generation. To investigate the fuel cell efficiency, the effect of hydrogen production on the open circuit fuel cell voltage is monitored during 27 days of batch culture. Values of volumetric hydrogen production, monitored by the help of the calibrated water columns, are related with the open circuit voltage changes of the fuel cell. From the analysis of this relation a dead end configuration is selected to use the fuel cell in its best potential. After the open circuit experiments external loads are tested for their effects on the fuel cell voltage and current generation. According to the results two external loads are selected for the direct usage of the fuel cell incorporating with the photobioreactors (PBR). Experiments with the PEM fuel cell generate a current density of 1.81 mA cm{sup -2} for about 50 h with 10 {omega} load and 0.23 mA cm{sup -2} for about 80 h with 100 {omega} load. (author)

  15. Producer gas production of Indonesian biomass in fixed-bed downdraft gasifier as an alternative fuels for internal combustion engines

    Science.gov (United States)

    Simanjuntak, J. P.; Lisyanto; Daryanto, E.; Tambunan, B. H.

    2018-03-01

    downdraft biomass gasification reactors, coupled with reciprocating internal combustion engines (ICE) are a viable technology for small scale heat and power generation. The direct use of producer gas as fuel subtitution in an ICE could be of great interest since Indonesia has significant land area in different forest types that could be used to produce bioenergy and convert forest materials to bioenergy for use in energy production and the versatility of this engine. This paper will look into the aspect of biomass energie as a contributor to energy mix in Indonesia. This work also contains information gathered from numerous previews study on the downdraft gasifier based on experimental or simulation study on the ability of producer gas as fuels for internal combustion engines aplication. All data will be used to complement the preliminary work on biomass gasification using downdraft to produce producer gas and its application to engines.

  16. Systems analysis and simulation of fissile materials disposition alternatives

    International Nuclear Information System (INIS)

    Farish, T.J.; Farmen, R.F.; Boerigter, S.T.; DeMuth, N.S.

    1996-01-01

    A detailed process flow model has been developed for use in the Fissile Materials Disposition program. The model calculates fissile material flows and inventories among the various processing and storage facilities over the life of the disposition program. Given existing inventories and schedules for processing, we can estimate the required size of processing and storage facilities, including equipment requirements, plant floorspace, approximate costs, and surge capacities. The model was designed to allow rapid prototyping, parallel and team development of facility and sub-facility models, consistent levels of detail and the use of a library of generic objects representing unit process operations

  17. Manufacturing method for fuel assembly

    International Nuclear Information System (INIS)

    Yamaguchi, Takashi.

    1997-01-01

    In an FBR type reactor, uranium/plutonium mixed oxide fuels (MOX fuels) are used. Nuclear fuel materials containing uranium and plutonium are filled to a portion or all of a plurality of fuel rods. In this case, an equivalent fissile coefficient (B) based on a combustion guarantee method defined by the formula: (B) = (M) · (F) is determined. (M) is a combustion matrix constituted based on the solution of equation of combustion which is a differential equation representing change with time of each of nuclear fuel materials during combustion. (F) is an equivalent fissile coefficient based on a reactivity keeping method which is a coefficient representing a reactivity worth equivalent with plutonium-239. The content of each of the nuclear fuel materials is determined so that the effective multiplication factor at the final stage of the operation cycle is substantially constant by using the equivalent fissile coefficient (B) based on the combustion guarantee method. (I.N.)

  18. Non-proliferation issues for the disposition of fissile materials using reactor alternatives

    International Nuclear Information System (INIS)

    Jaeger, C.D.; Duggan, R.A.; Tolk, K.M.

    1996-01-01

    The Department of Energy (DOE) is analyzing long-term storage on options for excess weapons-usable fissile materials. A number of the disposition alternatives are being considered which involve the use of reactors. The various reactor alternatives are all very similar and include front-end processes that could convert plutonium to a usable form for fuel fabrication, a MOX fuel fab facility, reactors to bum the MOX fuel and ultimate disposal of spent fuel in some geologic repository. They include existing, partially completed, advanced or evolutionary light water reactors and Canadian deuterium uranium (CANDU) reactors. In addition to the differences in the type of reactors, other variants on these alternatives are being evaluated to include the location and number of the reactors, the location of the mixed oxide (MOX) fabrication facility, the ownership of the facilities (private or government) and the colocation and/or separation of these facilities. All of these alternatives and their variants must be evaluated with respect to non-proliferation resistance. Both domestic and international safeguards support are being provided to DOE's Fissile Materials Disposition Program (FMDP) and includes such areas as physical protection, nuclear materials accountability and material containment and surveillance. This paper will focus on how the non-proliferation objective of reducing security risks and strengthening arms reduction will be accomplished and what some of the nonproliferation issues are for the reactor alternatives. Proliferation risk has been defined in terms of material form, physical environment, and the level of security and safeguards that is applied to the material. Metrics have been developed for each of these factors. The reactor alternatives will be evaluated with respect to these proliferation risk factors at each of the unit process locations in the alternative

  19. Non-proliferation issues for the disposition of fissile materials using reactor alternatives

    International Nuclear Information System (INIS)

    Jaeger, C.D.; Duggan, R.A.; Tolk, K.M.

    1996-01-01

    The Department of Energy (DOE) is analyzing long-term storage imposition options for excess weapons-usable fissile materials. A number of the disposition alternatives are being considered which involve the use of reactors. The various reactor alternatives are all very similar and include front-end processes that could convert plutonium to a usable form for fuel fabrication, a MOX fuel fab facility, reactors to burn the MOX fuel and ultimate disposal of spent fuel in some geologic repository. They include existing, partially completed, advanced or evolutionary light water reactors and Canadian deuterium uranium (CANDU) reactors. In addition to the differences in the type of reactors, other variants on these alternatives are being evaluated to include the location and number of the reactors, the location of the mixed oxide (MOX) fabrication facility, the ownership of the facilities (private or government) and the colocation and/or separation of these facilities. All of these alternatives and their variants must be evaluated with respect to non-proliferation resistance. Both domestic and international safeguards support are being provided to DOE's Fissile Materials Disposition Program (FMDP) and includes such areas as physical protection, nuclear materials accountability and material containment and surveillance. This paper will focus on how the non-proliferation objective of reducing security risks and strengthening arms reduction will be accomplished and what some of the non-proliferation issues are for the reactor alternatives. Proliferation risk has been defined in terms of material form, physical environment, and the level of security and safeguards that is applied to the material. Metrics have been developed for each of these factors. The reactor alternatives will be evaluated with respect to these proliferation risk factors at each of the unit process locations in the alternative

  20. Elemental balance of SRF production process: solid recovered fuel produced from municipal solid waste.

    Science.gov (United States)

    Nasrullah, Muhammad; Vainikka, Pasi; Hannula, Janne; Hurme, Markku; Oinas, Pekka

    2016-01-01

    In the production of solid recovered fuel (SRF), certain waste components have excessive influence on the quality of product. The proportion of rubber, plastic (hard) and certain textiles was found to be critical as to the elemental quality of SRF. The mass flow of rubber, plastic (hard) and textiles (to certain extent, especially synthetic textile) components from input waste stream into the output streams of SRF production was found to play the decisive role in defining the elemental quality of SRF. This paper presents the mass flow of polluting and potentially toxic elements (PTEs) in SRF production. The SRF was produced from municipal solid waste (MSW) through mechanical treatment (MT). The results showed that of the total input chlorine content to process, 55% was found in the SRF and 30% in reject material. Of the total input arsenic content, 30% was found in the SRF and 45% in fine fraction. In case of cadmium, lead and mercury, of their total input content to the process, 62%, 38% and 30%, respectively, was found in the SRF. Among the components of MSW, rubber material was identified as potential source of chlorine, containing 8.0 wt.% of chlorine. Plastic (hard) and textile components contained 1.6 and 1.1. wt.% of chlorine, respectively. Plastic (hard) contained higher lead and cadmium content compared with other waste components, i.e. 500 mg kg(-1) and 9.0 mg kg(-1), respectively. © The Author(s) 2015.

  1. Producing fuel alcohol by extractive distillation: Simulating the process with glycerol

    Directory of Open Access Journals (Sweden)

    Ana María Uyazán

    2006-01-01

    Full Text Available Downstream separation processes in biotechnology form part of the stages having most impact on a product’s final cost. The tendency throughout the world today is to replace fossil fuels with those having a renewable origin such as ethanol; this, in turn, produces a demand for the same and the need for optimising fermentation, treating vinazas and dehydration processes. The present work approaches the problem of dehydration through simulating azeotropic ethanol extractive distillation using glycerol as separation agent. Simulations were done on an Aspen Plus process simulator (Aspen Tech version 11.1. The simulated process involves two distillation columns, a dehydrator and a glycerol recuperation column. Simulation restrictions were ethanol’s molar composition in dehydrator column distillate and the process’s energy consumption. The effect of molar reflux ratio, solvent-feed ratio, solvent entry and feed stage and solvent entry temperature were evaluated on the chosen restrictions. The results showed that the ethanol-water mixture dehydration with glycerol as separation agent is efficient from the energy point of view.

  2. Mass-produced multi-walled carbon nanotubes as catalyst supports for direct methanol fuel cells.

    Science.gov (United States)

    Jang, In Young; Park, Ki Chul; Jung, Yong Chae; Lee, Sun Hyung; Song, Sung Moo; Muramatsu, Hiroyuki; Kim, Yong Jung; Endo, Morinobu

    2011-01-01

    Commercially mass-produced multi-walled carbon nanotubes, i.e., VGNF (Showa Denko Co.), were applied to support materials for platinum-ruthenium (PtRu) nanoparticles as anode catalysts for direct methanol fuel cells. The original VGNFs are composed of high-crystalline graphitic shells, which hinder the favorable surface deposition of the PtRu nanoparticles that are formed via borohydride reduction. The chemical treatment of VGNFs with potassium hydroxide (KOH), however, enables highly dispersed and dense deposition of PtRu nanoparticles on the VGNF surface. This capability becomes more remarkable depending on the KOH amount. The electrochemical evaluation of the PtRu-deposited VGNF catalysts showed enhanced active surface areas and methanol oxidation, due to the high dispersion and dense deposition of the PtRu nanoparticles. The improvement of the surface deposition states of the PtRu nanoparticles was significantly due to the high surface area and mesorporous surface structure of the KOH-activated VGNFs.

  3. Prompt neutron fission spectrum mean energies for the fissile nuclides and 252Cf

    International Nuclear Information System (INIS)

    Holden, N.E.

    1985-01-01

    The international standard for a neutron spectrum is that produced from the spontaneous fission of 252 Cf, while the thermal neutron induced fission neutron spectra for the four fissile nuclides, 233 U, 235 U, 239 Pu, and 241 Pu are of interest from the standpoint of nuclear reactors. The average neutron energies of these spectra are tabulated. The individual measurements are recorded with the neutron energy range measured, the method of detection as well as the average neutron energy for each author. Also tabulated are the measurements of the ratio of mean energies for pairs of fission neutron spectra. 75 refs., 9 tabs

  4. Apparatus and method for quantitatively evaluating total fissile and total fertile nuclide content in samples

    International Nuclear Information System (INIS)

    Caldwell, J.T.; Cates, M.R.; Franks, L.A.; Kunz, W.E.

    1985-01-01

    Simultaneous photon and neutron interrogation of samples for the quantitative determination of total fissile nuclide and total fertile nuclide material present is made possible by the use of an electron accelerator. Prompt and delayed neutrons produced from resulting induced fissions are counted using a single detection system and allow the resolution of the contributions from each interrogating flux leading in turn to the quantitative determination sought. Detection limits for 239 Pu are estimated to be about 3 mg using prompt fission neutrons and about 6 mg using delayed delayed neutrons

  5. Analyse of the potential of the high temperature reactor with respect to the use of fissile materials; Analyse des capacites des reacteurs a haute temperature sous l'aspect de l'utilisation des matieres fissiles

    Energy Technology Data Exchange (ETDEWEB)

    Damian, F

    2001-07-01

    The high temperature reactors fuel is made of micro-particles dispersed in a graphite matrix. This configuration makes it possible to reach high burnup, higher than 700 GWj/t. Thanks to the decoupling between the thermal and the neutronic behaviors in the core many types of fuels can be used. These characteristics give to HTR reactor very good capacities to burn fissile materials. This work was done in the frame of the evaluation of HTR capacities to enhance the value of the plutonium stocks. These stocks are currently composed of the irradiated fuels discharged from classical PWR or the dismantling of the nuclear weapons and represent a significant energy potential. These studies concluded that high cycles length can be reached whatever the plutonium quality is (from 50 % to 94 % of fissile plutonium). In addition, it was demonstrated that the moderator temperature coefficient becomes locally positive for highly burn fuel while the core global moderator temperature coefficient remained negative in the operation range of the reactor. A significant share of this work was first devoted to the setting of a modeling of the fuel element but also of the reactor's core with the codes of system SAPHYR. The whole of modeling was validated by reference calculations. This work of code assessment is justified by a preliminary work that showed that the classical calculation scheme used for PWR could not be transposed directly to HTR core. (author)

  6. Analyse of the potential of the high temperature reactor with respect to the use of fissile materials; Analyse des capacites des reacteurs a haute temperature sous l'aspect de l'utilisation des matieres fissiles

    Energy Technology Data Exchange (ETDEWEB)

    Damian, F

    2001-07-01

    The high temperature reactors fuel is made of micro-particles dispersed in a graphite matrix. This configuration makes it possible to reach high burnup, higher than 700 GWj/t. Thanks to the decoupling between the thermal and the neutronic behaviors in the core many types of fuels can be used. These characteristics give to HTR reactor very good capacities to burn fissile materials. This work was done in the frame of the evaluation of HTR capacities to enhance the value of the plutonium stocks. These stocks are currently composed of the irradiated fuels discharged from classical PWR or the dismantling of the nuclear weapons and represent a significant energy potential. These studies concluded that high cycles length can be reached whatever the plutonium quality is (from 50 % to 94 % of fissile plutonium). In addition, it was demonstrated that the moderator temperature coefficient becomes locally positive for highly burn fuel while the core global moderator temperature coefficient remained negative in the operation range of the reactor. A significant share of this work was first devoted to the setting of a modeling of the fuel element but also of the reactor's core with the codes of system SAPHYR. The whole of modeling was validated by reference calculations. This work of code assessment is justified by a preliminary work that showed that the classical calculation scheme used for PWR could not be transposed directly to HTR core. (author)

  7. Interim development report: engineering-scale HTGR fuel particle crusher

    International Nuclear Information System (INIS)

    Baer, J.W.; Strand, J.B.

    1978-09-01

    During the reprocessing of HTGR fuel, a double-roll crusher is used to fracture the silicon carbide coatings on the fuel particles. This report describes the development of the roll crusher used for crushing Fort-St.Vrain type fissile and fertile fuel particles, and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. Recommendations are made for design improvements and further testing

  8. Proposed plan for critical experiments supporting thorium fuel cycle development

    International Nuclear Information System (INIS)

    Gore, B.F.

    1978-09-01

    A preliminary plan is proposed for critical experiments to provide data needed for the recycle of thorium based nuclear fuels. The sequence of experimentation starts with well moderated solutions followed by highly concentrated low moderated solutions. It then progresses through lattices moderated by water, by water plus soluble poisons, and by fissile solutions, to solutions poisoned by raschig rings and soluble poisons. Final experiments would treat lattices moderated by poisoned fissile solution, and arrays of stored fissile units

  9. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  10. Ternary fission of spontaneously fissile uranium isomers excited by neutrons

    International Nuclear Information System (INIS)

    Makarenko, V.E.; Molchanov, Y.D.; Otroshchenko, G.A.; Yan'kov, G.B.

    1989-01-01

    Spontaneously fissile isomers (SFI) of uranium were excited in the reactions 236,238 U(n,n') at an average neutron energy 4.5 MeV. A pulsed electrostatic accelerator and time analysis of the fission events were used. Fission fragments were detected by the scintillation method, and long-range particles from fission were detected by an ionization method. The relative probability of fission of nuclei through a spontaneously fissile isomeric state was measured: (1.30±0.01)·10 -4 ( 236 U) and (1.48±0.02)·10 -4 ( 238 U). Half-lives of the isomers were determined: 121±2 nsec (the SFI 236 U) and 267±13 nsec (the SFI 238 U). In study of the ternary fission of spontaneously fissile isotopes of uranium it was established that the probability of the process amounts to one ternary fission per 163±44 binary fissions of the SFI 236 U and one ternary fission per 49±14 binary fissions of the SFI 238 U. The substantial increase of the probability of ternary fission of SFI of uranium in comparison with the case of ternary fission of nuclei which are not in an isomeric state may be related to a special nucleon configuration of the fissile isomers of uranium

  11. Fissile mass estimation by pulsed neutron source interrogation

    Energy Technology Data Exchange (ETDEWEB)

    Israelashvili, I., E-mail: israelashvili@gmail.com [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel); Dubi, C.; Ettedgui, H.; Ocherashvili, A. [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel); Pedersen, B. [Nuclear Security Unit, Institute for Transuranium Elements, Joint Research Centre, Via E. Fermi, 2749, 21027 Ispra (Italy); Beck, A. [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel); Roesgen, E.; Crochmore, J.M. [Nuclear Security Unit, Institute for Transuranium Elements, Joint Research Centre, Via E. Fermi, 2749, 21027 Ispra (Italy); Ridnik, T.; Yaar, I. [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel)

    2015-06-11

    Passive methods for detecting correlated neutrons from spontaneous fissions (e.g. multiplicity and SVM) are widely used for fissile mass estimations. These methods can be used for fissile materials that emit a significant amount of fission neutrons (like plutonium). Active interrogation, in which fissions are induced in the tested material by an external continuous source or by a pulsed neutron source, has the potential advantages of fast measurement, alongside independence of the spontaneous fissions of the tested fissile material, thus enabling uranium measurement. Until recently, using the multiplicity method, for uranium mass estimation, was possible only for active interrogation made with continues neutron source. Pulsed active neutron interrogation measurements were analyzed with techniques, e.g. differential die away analysis (DDA), which ignore or implicitly include the multiplicity effect (self-induced fission chains). Recently, both, the multiplicity and the SVM techniques, were theoretically extended for analyzing active fissile mass measurements, made by a pulsed neutron source. In this study the SVM technique for pulsed neutron source is experimentally examined, for the first time. The measurements were conducted at the PUNITA facility of the Joint Research Centre in Ispra, Italy. First promising results, of mass estimation by the SVM technique using a pulsed neutron source, are presented.

  12. Safeguards and security issues for the disposition of fissile materials

    International Nuclear Information System (INIS)

    Jaeger, C.D.; Moya, R.W.; Duggan, R.A.; Mangan, D.L.; Tolk, K.M.; Rutherford, D.; Fearey, B.; Moore, L.

    1995-01-01

    The Department of Energy's Office of Fissile Material Disposition (FMD) is analyzing long-term storage and disposition options for surplus weapons-usable fissile materials, preparing a programmatic environmental impact statement (PEIS), preparing for a record of decision (ROD) regarding this material and conducting other activities. The primary security objectives of this program are to reduce major security risks and strengthen arms reduction and nonproliferation (NP). To help achieve these objectives, a safeguards and security (S ampersand S) team consisting of participants from Sandia, Los Alamos, and Lawrence Livermore National Laboratories was established. The S ampersand S activity for this program is a cross-cutting task which addresses all of the FMD program options. It includes both domestic and international safeguards and includes areas such as physical protection, nuclear materials accountability and material containment and surveillance. This paper will discuss the activities of the Fissile Materials Disposition Program (FMDP) S ampersand S team as well as some specific S ampersand S issues associated with various FMDP options/facilities. Some of the items to be discussed include the threat, S ampersand S requirements, S ampersand S criteria for assessing risk, S ampersand S issues concerning fissile material processing/facilities, and international and domestic safeguards

  13. Multilevel parametrization of fissile nuclei resonance cross sections

    International Nuclear Information System (INIS)

    Lukyanov, A.A.; Kolesov, V.V.; Janeva, N.

    1987-01-01

    Because the resonance interference has an important influence on the resonance structure of neutron cross sections energy dependence at lowest energies, multilevel scheme of the cross section parametrization which take into account the resonance interference is used for the description with the same provisions in the regions of the interferential maximum and minimum of the resonance cross sections of the fissile nuclei

  14. Gold nanoparticles produced in situ mediate bioelectricity and hydrogen production in a microbial fuel cell by quantized capacitance charging.

    Science.gov (United States)

    Kalathil, Shafeer; Lee, Jintae; Cho, Moo Hwan

    2013-02-01

    Oppan quantized style: By adding a gold precursor at its cathode, a microbial fuel cell (MFC) is demonstrated to form gold nanoparticles that can be used to simultaneously produce bioelectricity and hydrogen. By exploiting the quantized capacitance charging effect, the gold nanoparticles mediate the production of hydrogen without requiring an external power supply, while the MFC produces a stable power density. Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  15. First Industrial Tests of a Drum Monitor Matrix Correction for the Fissile Mass Measurement in Large Volume Historic Metallic Residues with the Differential Die-away Technique

    Energy Technology Data Exchange (ETDEWEB)

    Antoni, R.; Passard, C.; Perot, B.; Batifol, M.; Vandamme, J.C. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 St Paul-lez-Durance, (France); Grassi, G. [AREVA NC, 1 place Jean-Millier, 92084 Paris-La-Defense cedex (France)

    2015-07-01

    The fissile mass in radioactive waste drums filled with compacted metallic residues (spent fuel hulls and nozzles) produced at AREVA La Hague reprocessing plant is measured by neutron interrogation with the Differential Die-away measurement Technique (DDT. In the next years, old hulls and nozzles mixed with Ion-Exchange Resins will be measured. The ion-exchange resins increase neutron moderation in the matrix, compared to the waste measured in the current process. In this context, the Nuclear Measurement Laboratory (NML) of CEA Cadarache has studied a matrix effect correction method, based on a drum monitor ({sup 3}He proportional counter inside the measurement cavity). A previous study performed with the NML R and D measurement cell PROMETHEE 6 has shown the feasibility of method, and the capability of MCNP simulations to correctly reproduce experimental data and to assess the performances of the proposed correction. A next step of the study has focused on the performance assessment of the method on the industrial station using numerical simulation. A correlation between the prompt calibration coefficient of the {sup 239}Pu signal and the drum monitor signal was established using the MCNPX computer code and a fractional factorial experimental design composed of matrix parameters representative of the variation range of historical waste. Calculations have showed that the method allows the assay of the fissile mass with an uncertainty within a factor of 2, while the matrix effect without correction ranges on 2 decades. In this paper, we present and discuss the first experimental tests on the industrial ACC measurement system. A calculation vs. experiment benchmark has been achieved by performing dedicated calibration measurement with a representative drum and {sup 235}U samples. The preliminary comparison between calculation and experiment shows a satisfactory agreement for the drum monitor. The final objective of this work is to confirm the reliability of the

  16. First Industrial Tests of a Drum Monitor Matrix Correction for the Fissile Mass Measurement in Large Volume Historic Metallic Residues with the Differential Die-away Technique

    International Nuclear Information System (INIS)

    Antoni, R.; Passard, C.; Perot, B.; Batifol, M.; Vandamme, J.C.; Grassi, G.

    2015-01-01

    The fissile mass in radioactive waste drums filled with compacted metallic residues (spent fuel hulls and nozzles) produced at AREVA La Hague reprocessing plant is measured by neutron interrogation with the Differential Die-away measurement Technique (DDT. In the next years, old hulls and nozzles mixed with Ion-Exchange Resins will be measured. The ion-exchange resins increase neutron moderation in the matrix, compared to the waste measured in the current process. In this context, the Nuclear Measurement Laboratory (NML) of CEA Cadarache has studied a matrix effect correction method, based on a drum monitor ( 3 He proportional counter inside the measurement cavity). A previous study performed with the NML R and D measurement cell PROMETHEE 6 has shown the feasibility of method, and the capability of MCNP simulations to correctly reproduce experimental data and to assess the performances of the proposed correction. A next step of the study has focused on the performance assessment of the method on the industrial station using numerical simulation. A correlation between the prompt calibration coefficient of the 239 Pu signal and the drum monitor signal was established using the MCNPX computer code and a fractional factorial experimental design composed of matrix parameters representative of the variation range of historical waste. Calculations have showed that the method allows the assay of the fissile mass with an uncertainty within a factor of 2, while the matrix effect without correction ranges on 2 decades. In this paper, we present and discuss the first experimental tests on the industrial ACC measurement system. A calculation vs. experiment benchmark has been achieved by performing dedicated calibration measurement with a representative drum and 235 U samples. The preliminary comparison between calculation and experiment shows a satisfactory agreement for the drum monitor. The final objective of this work is to confirm the reliability of the modeling approach

  17. Analyse of the potential of the high temperature reactor with respect to the use of fissile materials

    International Nuclear Information System (INIS)

    Damian, F.

    2001-01-01

    The high temperature reactors fuel is made of micro-particles dispersed in a graphite matrix. This configuration makes it possible to reach high burnup, higher than 700 GWj/t. Thanks to the decoupling between the thermal and the neutronic behaviors in the core many types of fuels can be used. These characteristics give to HTR reactor very good capacities to burn fissile materials. This work was done in the frame of the evaluation of HTR capacities to enhance the value of the plutonium stocks. These stocks are currently composed of the irradiated fuels discharged from classical PWR or the dismantling of the nuclear weapons and represent a significant energy potential. These studies concluded that high cycles length can be reached whatever the plutonium quality is (from 50 % to 94 % of fissile plutonium). In addition, it was demonstrated that the moderator temperature coefficient becomes locally positive for highly burn fuel while the core global moderator temperature coefficient remained negative in the operation range of the reactor. A significant share of this work was first devoted to the setting of a modeling of the fuel element but also of the reactor's core with the codes of system SAPHYR. The whole of modeling was validated by reference calculations. This work of code assessment is justified by a preliminary work that showed that the classical calculation scheme used for PWR could not be transposed directly to HTR core. (author)

  18. Sustainably produced ethanol. A premium fuel component; Nachhaltig produziertes Ethanol. Eine Premium Kraftstoffkomponente

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, Joerg [Suedzucker AG, Obrigheim/Pfalz (Germany)

    2012-07-01

    Ethanol is the most used biofuel in the world. It is part of the European biofuel strategy, which is intended to preserve finite fossil resources, reduce greenhouse gas emissions and strengthen European agriculture. In addition to its traditional use in E5 fuel, ethanol most recently features in new fuels for petrol engines in Europe: as E10 as an expansion of the already existing concept of ethanol blends, such as in E5, or as ethanol fuel E85, a blend made up primarily of ethanol. There is already extensive international experience for both types of fuel for example in the USA or Brazil. The use of ethanol as a biofuel is linked to sustainability criteria in Europe which must be proven through a certification scheme. In addition to ethanol, the integrated production process also provides vegetable protein which is used in food as well as in animal feed and therefore provides the quality products of processed plants used for sustainable energy and in animal and human food. Ethanol has an effect on the vapour pressure, boiling behaviour and octane number of the fuel blend. Adjusting the blend stock petrol to fulfil the quality requirements of the final fuel is therefore necessary. Increasing the antiknock properties, increasing the heat of evaporation of the fuel using ethanol and the positive effects this has on the combustion efficiency of the petrol engine are particularly important. Investigations on cars or engines that were specifically designed for fuel with a higher ethanol content show significant improvements in using the energy from the fuel and the potential to reduce carbon dioxide emissions if fuels containing ethanol are used. The perspective based purely on an energy equivalent replacement of fossil fuels with ethanol is therefore misleading. Ethanol can also contribute to increasing the energy efficiency of petrol engines as well as being a replacement source of energy. (orig.)

  19. Assesment of the energy quality of the synthesis gas produced from biomass derived fuels conversion: Part I: Liquid Fuels, Ethanol

    International Nuclear Information System (INIS)

    Arteaga Perez, Luis E; Casas, Yannay; Peralta, Luis M; Granda, Daikenel; Prieto, Julio O

    2011-01-01

    The use of biofuels plays an important role to increase the efficiency and energetic safety of the energy processes in the world. The main goal of the present research is to study from the thermodynamics and kinetics the effect of the operational variables on the thermo-conversion processes of biomass derived fuels focused on ethanol reforming. Several models are developed to assess the technological proposals. The minimization of Gibbs free energy is the criterion applied to evaluate the performance of the different alternatives considering the equilibrium constraints. All the models where validated on an experimental data base. The gas composition, HHV and the ratio H2/CO are used as measures for the process efficiency. The operational parameters are studied in a wide range (reactants molar ratio, temperature and oxygen/fuel ratio). (author)

  20. Covariance Spectroscopy for Fissile Material Detection

    International Nuclear Information System (INIS)

    Trainham, Rusty; Tinsley, Jim; Hurley, Paul; Keegan, Ray

    2009-01-01

    Nuclear fission produces multiple prompt neutrons and gammas at each fission event. The resulting daughter nuclei continue to emit delayed radiation as neutrons boil off, beta decay occurs, etc. All of the radiations are causally connected, and therefore correlated. The correlations are generally positive, but when different decay channels compete, so that some radiations tend to exclude others, negative correlations could also be observed. A similar problem of reduced complexity is that of cascades radiation, whereby a simple radioactive decay produces two or more correlated gamma rays at each decay. Covariance is the usual means for measuring correlation, and techniques of covariance mapping may be useful to produce distinct signatures of special nuclear materials (SNM). A covariance measurement can also be used to filter data streams because uncorrelated signals are largely rejected. The technique is generally more effective than a coincidence measurement. In this poster, we concentrate on cascades and the covariance filtering problem

  1. Study Of Thorium As A Nuclear Fuel.

    Directory of Open Access Journals (Sweden)

    Prakash Humane

    2017-10-01

    Full Text Available Conventional fuel sources for power generation are to be replacing by nuclear power sources like nuclear fuel Uranium. But Uranium-235 is the only fissile fuel which is in 0.72 found in nature as an isotope of Uranium-238. U-238 is abundant in nature which is not fissile while U-239 by alpha decay naturally converted to Uranium- 235. For accompanying this nuclear fuel there is another nuclear fuel Thorium is present in nature is abundant can be used as nuclear fuel and is as much as safe and portable like U-235.

  2. Fuel burn-up distribution and transuranic nuclide contents produced at the first cycle operation of AP1000

    International Nuclear Information System (INIS)

    Jati Susilo; Jupiter Sitorus Pane

    2016-01-01

    AP1000 reactor core was designed with nominal power of 1154 MWe (3415 MWth), operated within life time of 60 years and cycle length of 18 months. For the first cycle, the AP1000 core uses three kinds of UO 2 enrichment, they are 2.35 w/o, 3.40 w/o and 4.45 w/o. Absorber materials such as ZrB 2 , Pyrex and Boron solution are used to compensate the excess reactivity at the beginning of cycle. In the core, U-235 fuels are burned by fission reaction and produce energy, fission products and new neutron. Because of the U-238 neutron absorption reaction, the high level radioactive waste of heavy nuclide transuranic such as Pu, Am, Cm and Np are also generated. They have a very long half life. The purpose of this study is to evaluate the result of fuel burn-up distribution and heavy nuclide transuranic contents produced by AP1000 at the end of first cycle operation (EOFC). Calculation of ¼ part of the AP1000 core in the 2 dimensional model has been done using SRAC2006 code with the module of COREBN/HIST. The input data called the table of macroscopic cross section, is calculated using module of PIJ. The result shows that the maximum fuel assembly (FA) burn-up is 27.04 GWD/MTU, that is still lower than allowed maximum burn-up of 62 GWD/MTU. Fuel loading position at the center/middle of the core will produce bigger burn-up and transuranic nuclide than one at the edges the of the core. The use of IFBA fuel just give a small effect to lessen the fuel burn-up and transuranic nuclide production. (author)

  3. The nuclear fuel elements' world market and the position of the Argentine Republic as producer

    International Nuclear Information System (INIS)

    Biondo, C.D.

    1983-01-01

    The development of the nuclear fuel elements' industry is analyzed, both in the present and projected world market, up to the year 2000, in the light of the situation affecting the nucleoelectric industry. By means of the offer/demand function, an analysis is made of the behaviour of the fuel elements' market throughout the fuel cycle structure. The regional unbalances between availability and demand of uranium resources are considered, as well as the factors having an unfavorable incidence on the fuel cycle's economic equation. The economic structure to be used for the calculation of the nucleoelectric generating cost is presented, in order to situate, within said nuclear economy, the component corresponding to the fuel cycle cost. Emphasis is placed on the 'front end' stages of the fuel cycle, but also considering those stages belonging to the 'back end'. Argentina's fuel elements market and its present and projected nucleoelectric park are analyzed, indicating their relative position in the world market. (R.J.S.) [es

  4. Characterization and Performance Test of Palm Oil Based Bio-Fuel Produced Via Ni/Zeolite-Catalyzed Cracking Process

    Directory of Open Access Journals (Sweden)

    Sri Kadarwati

    2015-02-01

    Full Text Available Catalytic cracking process of palm oil into bio-fuel using Ni/zeolite catalysts (2-10% wt. Ni at various reaction temperatures (400-500oC in a flow-fixed bed reactor system has been carried out. Palm oil was pre-treated to produce methyl ester of palm oil as feedstock in the catalytic cracking reactions. The Ni/zeolite catalysts were prepared by wetness impregnation method using Ni(NO32.6H2O as the precursor. The products were collected and analysed using GC, GC-MS, and calorimeter. The effects of process temperatures and Ni content in Ni/zeolite have been studied. The results showed that Ni-2/zeolite could give a yield of 99.0% at 500oC but only produced gasoline fraction of 18.35%. The physical properties of bio-fuel produced in this condition in terms of density, viscosity, flash point, and specific gravity were less than but similar to commercial fuel. The results of performance test in a 4-strike engine showed that the mixture of commercial gasoline (petrol and bio-fuel with a ratio of 9:1 gave similar performance to fossil-based gasoline with much lower CO and O2 emissions and more efficient combustion

  5. Future developments and technological and economic assessment of methods for producing synthetic liquid fuel from coal

    Energy Technology Data Exchange (ETDEWEB)

    Shlikhter, E B; Khor' kov, A V; Zhorov, Yu M

    1980-11-01

    Promising methods for obtaining synthetic liquid fuel from coal are surveyed and described: thermal dissolution of coal by means of a hydrogen donor solution: hydrogenation; gasification with subsequent synthesis and pyrolysis. A technological and economic assessment of the above processes is given. Emphasis is placed on methods employing catalytic conversion of methanol into hydrocarbon fuels. On the basis of thermodynamic calculations of the process for obtaining high-calorific liquid fuel from methanol the possibility of obtaining diesel fractions as well as gasoline is demonstrated. (12 refs.) (In Russian)

  6. Neutronic study of heavy nucleus produced in nuclear reactor fuel cycle

    International Nuclear Information System (INIS)

    Giacometti, A.

    1978-01-01

    Importance of minor actinides (U, Np, Pu, Am and Cm isotopes) PWR and fast neutron reactors and their associated fuel cycle is examined in this thesis. The amount of actinides formed in the various types of fuels or reactors are given. The different ways of formation and their importance are described. Modifications of the core reactivity due to actinides are shown. After a review of the fuel cycle (enrichment, fabrication, reprocessing, transport) actinide evolution outside the core is described and main problems concerning radioactivity in the different steps of the cycle or long term storage are underlined [fr

  7. Recycling used palm oil and used engine oil to produce white bio oil, bio petroleum diesel and heavy fuel

    Science.gov (United States)

    Al-abbas, Mustafa Hamid; Ibrahim, Wan Aini Wan; Sanagi, Mohd. Marsin

    2012-09-01

    Recycling waste materials produced in our daily life is considered as an additional resource of a wide range of materials and it conserves the environment. Used engine oil and used cooking oil are two oils disposed off in large quantities as a by-product of our daily life. This study aims at providing white bio oil, bio petroleum diesel and heavy fuel from the disposed oils. Toxic organic materials suspected to be present in the used engine oil were separated using vacuum column chromatography to reduce the time needed for the separation process and to avoid solvent usage. The compounds separated were detected by gas chromatography-mass spectrometry (GC-MS) and found to contain toxic aromatic carboxylic acids. Used cooking oils (thermally cracked from usage) were collected and separated by vacuum column chromatography. White bio oil produced was examined by GC-MS. The white bio oil consists of non-toxic hydrocarbons and is found to be a good alternative to white mineral oil which is significantly used in food industry, cosmetics and drugs with the risk of containing polycyclic aromatic compounds which are carcinogenic and toxic. Different portions of the used cooking oil and used engine were mixed to produce several blends for use as heavy oil fuels. White bio oil was used to produce bio petroleum diesel by blending it with petroleum diesel and kerosene. The bio petroleum diesel produced passed the PETRONAS flash point and viscosity specification test. The heat of combustion of the two blends of heavy fuel produced was measured and one of the blends was burned to demonstrate its burning ability. Higher heat of combustion was obtained from the blend containing greater proportion of used engine oil. This study has provided a successful recycled alternative for white bio oil, bio petroleum fuel and diesel which can be an energy source.

  8. Reduction of the uncertainty due to fissile clusters in radioactive waste characterization with the Differential Die-away Technique

    Science.gov (United States)

    Antoni, R.; Passard, C.; Perot, B.; Guillaumin, F.; Mazy, C.; Batifol, M.; Grassi, G.

    2018-07-01

    AREVA NC is preparing to process, characterize and compact old used fuel metallic waste stored at La Hague reprocessing plant in view of their future storage ("Haute Activité Oxyde" HAO project). For a large part of these historical wastes, the packaging is planned in CSD-C canisters ("Colis Standard de Déchets Compacté s") in the ACC hulls and nozzles compaction facility ("Atelier de Compactage des Coques et embouts"). . This paper presents a new method to take into account the possible presence of fissile material clusters, which may have a significant impact in the active neutron interrogation (Differential Die-away Technique) measurement of the CSD-C canisters, in the industrial neutron measurement station "P2-2". A matrix effect correction has already been investigated to predict the prompt fission neutron calibration coefficient (which provides the fissile mass) from an internal "drum flux monitor" signal provided during the active measurement by a boron-coated proportional counter located in the measurement cavity, and from a "drum transmission signal" recorded in passive mode by the detection blocks, in presence of an AmBe point source in the measurement cell. Up to now, the relationship between the calibration coefficient and these signals was obtained from a factorial design that did not consider the potential for occurrence of fissile material clusters. The interrogative neutron self-shielding in these clusters was treated separately and resulted in a penalty coefficient larger than 20% to prevent an underestimation of the fissile mass within the drum. In this work, we have shown that the incorporation of a new parameter in the factorial design, representing the fissile mass fraction in these clusters, provides an alternative to the penalty coefficient. This new approach finally does not degrade the uncertainty of the original prediction, which was calculated without taking into consideration the possible presence of clusters. Consequently, the

  9. The role of accelerators in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Takahashi, Hiroshi.

    1990-01-01

    The use of neutrons produced by the medium energy proton accelerator (1 GeV--3 GeV) has considerable potential in reconstructing the nuclear fuel cycle. About 1.5 ∼ 2.5 ton of fissile material can be produced annually by injecting a 450 MW proton beam directly into fertile materials. A source of neutrons, produced by a proton beam, to supply subcritical reactors could alleviate many of the safety problems associated with critical assemblies, such as positive reactivity coefficients due to coolant voiding. The transient power of the target can be swiftly controlled by controlling the power of the proton beam. Also, the use of a proton beam would allow more flexibility in the choice of fuel and structural materials which otherwise might reduce the reactivity of reactors. This paper discusses the rate of accelerators in the transmutation of radioactive wastes of the nuclear fuel cycles. 34 refs., 17 figs., 9 tabs

  10. Supplying the six. [Supplies of nuclear fuels and ores to the European Community

    Energy Technology Data Exchange (ETDEWEB)

    Oboussier, F

    1975-07-01

    Under the Euratom Treaty, the European Community must ensure that all users in the Community receive a regular and equitable supply of ores and nuclear fuels. Supply to users in the Community of ores, source materials, and special fissile materials is based on the principle of equal access of the users to the supply sources. To ensure such equal access, the Treaty prohibits all practices designed to secure a privileged position for certain users. In addition, an agency has been set up with two essential rights--that of an option on all ores, source materials, and special fissile materials produced in the territories of the Member States; and the exclusive right to conclude all contracts relating to the supply of ores, source materials, and special fissile materials coming from inside the Community or from outside. Dealings of the Agency with outside agencies, especially the former US AEC, are described. The uranium market and its economics and the availability of special fissile materials are summarized. (MCW)

  11. Contribution of civilian industry to the management of military fissile materials

    International Nuclear Information System (INIS)

    Montalembert de, J.A.

    2001-01-01

    The situation about using of highly enriched uranium (HEU) and weapon grade plutonium (WgPu) for nuclear fuel preparation in U.S.A. and Russian Federation is reviewed. A few remarks were concluded: (1) We stand at the onset of a process that will be lengthy and which is unlikely to stop with the elimination of the 700 t of HEU and 2 x 34.5 t of WgPu concerned so far. If the announced negotiation of the third START treaty concludes favorably, additional tonnages will have to be recycled, particularly on the Russian side whose estimated inventory is larger. (2) The time scales necessitated by the management of these materials should be no surprise. On the one hand, the aim is to reduce an arsenal built up during 45 years of a Cold War. And this return to civilian life of materials of military origin must be achieved in conditions of safety and bilateral or international safeguards (IAEA), which obviously did not constitute the primary concern of the powers who produced them. Besides, insofar as it enlists the services of civilian industry, this return must be carried out with due respect for the equilibrium of markets that are severely mauled today, in other words, in an orderly and progressive manner. (3) Finally, it is important to recognize that without the contribution of the nuclear power industry, the elimination of military fissile materials would raise problems at another scale and would inevitably lead to regrettable waste. It is to be hoped that this will jog the minds of those who urge a rapid end to nuclear energy, when all the evidence demonstrates that the best way to eliminate surplus weapon grade materials is to recycle them in a reactor, in other words, to destroy them or to denature them while generating electricity. (4) The civilian nuclear industry is happy to contribute concretely and significantly to the solution of a problem of surplus nuclear weaponry, while at the same time utilizing technologies successfully developed for power generation

  12. Exploiting Fission Chain Reaction Dynamics to Image Fissile Materials

    Science.gov (United States)

    Chapman, Peter Henry

    Radiation imaging is one potential method to verify nuclear weapons dismantlement. The neutron coded aperture imager (NCAI), jointly developed by Oak Ridge National Laboratory (ORNL) and Sandia National Laboratories (SNL), is capable of imaging sources of fast (e.g., fission spectrum) neutrons using an array of organic scintillators. This work presents a method developed to discriminate between non-multiplying (i.e., non-fissile) neutron sources and multiplying (i.e., fissile) neutron sources using the NCAI. This method exploits the dynamics of fission chain-reactions; it applies time-correlated pulse-height (TCPH) analysis to identify neutrons in fission chain reactions. TCPH analyzes the neutron energy deposited in the organic scintillator vs. the apparent neutron time-of-flight. Energy deposition is estimated from light output, and time-of-flight is estimated from the time between the neutron interaction and the immediately preceding gamma interaction. Neutrons that deposit more energy than can be accounted for by their apparent time-of-flight are identified as fission chain-reaction neutrons, and the image is reconstructed using only these neutron detection events. This analysis was applied to measurements of weapons-grade plutonium (WGPu) metal and 252Cf performed at the Nevada National Security Site (NNSS) Device Assembly Facility (DAF) in July 2015. The results demonstrate it is possible to eliminate the non-fissile 252Cf source from the image while preserving the fissileWGPu source. TCPH analysis was also applied to additional scenes in which theWGPu and 252Cf sources were measured individually. The results of these separate measurements further demonstrate the ability to remove the non-fissile 252Cf source and retain the fissileWGPu source. Simulations performed using MCNPX-PoliMi indicate that in a one hour measurement, solid spheres ofWGPu are retained at a 1sigma level for neutron multiplications M -˜ 3.0 and above, while hollowWGPu spheres are

  13. Performance and emissions of an engine fuelled with a biodiesel fuel produced from animal fats

    Directory of Open Access Journals (Sweden)

    Taymaz Imdat

    2013-01-01

    Full Text Available Oil reserves which are located around the world are declining day by day, so new alternative energy sources must be invented for engines of internal combustion and compression ignition, so biodiesel that is an alternative fuel source for diesel engines and it is a renewable energy resource. Biodiesel is a fuel made from vegetable oils, animals’ fats and waste oils. In this study, physical and chemical properties of biodiesel were analyzed and matched to the diesel fuel. In the experimental study, biodiesel was made from animal fats and compared to diesel fuel. Its effects on engine performance and emissions are studied. A single-cylinder, four-stroke, direct injected diesel engine with air cooling system are used as test equipment in different cycles. After the experimental study, it is concluded that the reduction of the emissions of CO and HC as biodiesel has the advantage of emission output. Environmentalist property of biodiesel is the most important characteristic of it. But the sight of engine performance diesel fuel has more advantage to biodiesel fuel.

  14. Safety analysis report: packages. Argonne National Laboratory SLSF test train shipping container, P-1 shipment. Fissile material. Final report

    International Nuclear Information System (INIS)

    Meyer, C.A.

    1975-06-01

    The package is used to ship an instrumented test fuel bundle (test train) containing fissile material. The package assembly is Argonne National Laboratory (ANL) Model R1010-0032. The shipment is fissile class III. The packaging consists of an outer carbon steel container into which an inner container is placed; the inner container is separated from the outer container by urethane foam cushioning material. The test train is supported in the inner container by a series of transverse supports spaced along the length of the test train. Both the inner and outer containers are closed with bolted covers. The covers do not seal the containers in a leaktight manner. The gross weight of the shipment is about 8350 lb. The unirradiated fissile material content is less than 3 kg of UO 2 of up to 93.2 percent enrichment. This is a Type A quantity (transport group III and less than 3 curies) of radioactive material which does not require shielding, cooling or heating, or neutron absorption or moderation functions in its packaging. The maximum exterior dimensions of the container are 37 ft 11 in. long, 24 1 / 2 in. wide, and 19 3 / 4 in. high

  15. Vitrification of HLW produced by uranium/molybdenum fuel reprocessing in cogema's cold crucible melter

    International Nuclear Information System (INIS)

    Quang, R. Do; Petitjean, V.; Hollebeque, F.; Pinet, O.; Flament, T.; Prodhomme, A.; Dalcorso, J. P.

    2003-01-01

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R and D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12% in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed

  16. Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA's Cold Crucible Melter

    International Nuclear Information System (INIS)

    Do Quang, R.; Petitjean, V.; Hollebecque, F.; Pinet, O.; Flament, T.; Prod'homme, A.

    2003-01-01

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R and D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed

  17. Calculation of multiplication factors regarding criticality aiming at the storage of fissile material

    International Nuclear Information System (INIS)

    Lima Barros, M. de.

    1982-04-01

    The multiplication factors of several systems with low enrichment, 3,5% and 3,2% in the isotope 235 U, aiming at the storage of fuel of ANGRA-I and ANGRA II, through the method of Monte Carlo, by the computacional code KENO-IV and the library of section of cross Hansen - Roach with 16 groups of energy. The method of Monte Carlo is specially suitable to the calculation of the factor of multiplication, because it is one of the most acurate models of solution and allows the description of complex tridimensional systems. Various tests of sensibility of this method have been done in order to present the most convenient way of working with KENO-IV code. The safety on criticality of stores of fissile material of the 'Fabrica de Elementos Combustiveis ', has been analyzed through the method of Monte Carlo. (Author) [pt

  18. Experimental spectrum of reactor antineutrinos and spectra of main fissile isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Sinev, V. V., E-mail: vsinev@pcbai10.inr.ruhep.ru [Russian Academy of Sciences, Institute for Nuclear Research (Russian Federation)

    2013-05-15

    Within the period between the years 1988 and 1990, the spectrum of positrons from the inverse-beta-decay reaction on a proton was measured at the Rovno atomic power plant in the course of experiments conducted there. The measured spectrum has the vastest statistics in relation to other neutrino experiments at nuclear reactors and the lowest threshold for positron detection. An experimental reactor-antineutrino spectrum was obtained on the basis of this positron spectrum and was recommended as a reference spectrum. The spectra of individual fissile isotopes were singled out from the measured antineutrino spectrum. These spectra can be used to analyze neutrino experiments performed at nuclear reactors for various compositions of the fuel in the reactor core.

  19. Improved resonance formulas for cross sections of fissile elements

    International Nuclear Information System (INIS)

    Segev, M.

    1978-01-01

    The Adler--Adler cross-section formalism with energy-dependent parameters is a practical approximation to the R-matrix formalism, on the basis of the smallness of the s-wave neutron width in fissile elements. Attempts were made to represent experimental cross sections by the Adler--Adler formulas through an initial representation by the Reich--Moore approximation of R-matrix and a subsequent conversion of the Reich--Moore formulas to the Adler--Adler formulas. Adler and Adler foresaw difficulties in associating their formulas with approximate R-matrix theories such as those of Reich and Moore. Indeed, it is shown that, due to the nonunitarity of the Adler--Adler formalism on the one hand and the unitarity, by definition, of the Reich--Moore formalism on the other hand, the conversion from the latter to the former is ambiguous. Examples are shown to demonstrate that this ambiguity results in numerical inaccuracies, sometimes very large ones, for neutron widths that are not extremely small. Improved Adler--Adler-type formulas have been derived from the R-matrix formalism. In these formulas, the multipliers of the Breit--Wigner resonance lines exhibit more explicit energy dependence than their original counterparts, mainly in the form of additional terms in the formula for the total cross section. The conversion from Reich--Moore cross sections to the improved resonance formulas is shown to be much less ambiguous and to produce very accurate cross sections. In particular, the inaccuracies encountered with the Reich--Moore to Adler--Adler conversion are eliminated. A computer code, PEDRA, was written to perform the conversion from a given set of Reich--Moore parameters to the parameters required in the improved formulas. The numerical algorithm of this code is based on an adaptation with modifications of the numerical approach of de Saussure--Perez in the POLLA code, which converts Reich--Moore parameters to Adler--Adler parameters. 7 figures, 1 table

  20. Modeling of fissile material diversion in solvent extraction cascades

    International Nuclear Information System (INIS)

    Schneider, A.; Carlson, R.W.

    1980-01-01

    Changes were calculated for measurable parameters of a solvent extraction section of a reprocessing plant resulting from postulated fissile material diversion actions. The computer program SEPHIS was modified to calculate the time-dependent concentrations of uranium and plutonium in each stage of a cascade. The calculation of the inventories of uranium and plutonium in each contactor was also included. The concentration and inventory histories were computed for a group of four sequential columns during start-up and for postulated diversion conditions within this group of columns. Monitoring of column exit streams or of integrated column inventories for fissile materials could provide qualitative indications of attempted diversions. However, the time delays and resulting changes are complex and do not correlate quantitatively with the magnitude of the initiating event

  1. Gamma ray absorption of cylindrical fissile material with dual shields

    International Nuclear Information System (INIS)

    Wu Chenyan; Cheng Yiying; Huang Yongyi; Lu Fuquan; Yang Fujia

    2005-01-01

    This work analyzed the gamma ray attenuation effect from the self-absorption and shield attenuation perspectively. An exact mathematical equation was given for the geometric factor of the cylindrical fissile material with dual shields. In addition, several approximation approaches suitable for real situation were discussed, especially in the radial and axial directions of the cylinders, since the G-factors have simple forms. Then the space distribution patterns of the G-factor were analyzed based on numerical result and effective ways to solved the geometric information of the cylindrical fissile material, the radii and the heights, were deduced. This method was checked and verified by numerical calculation. Because of the efficiency of the method, it is ideal for application in real situations, such as nuclear safeguards, which demands speed of detection and accuracy of geometric analysis. (authors)

  2. User manual of FUNF code for fissile material data calculation

    International Nuclear Information System (INIS)

    Zhang, Jingshang

    2006-03-01

    The FUNF code (2005 version) is used to calculate fast neutron reaction data of fissile materials with incident energies from about 1 keV up to 20 MeV. The first version of the FUNF code was completed in 1994. the code has been developed continually since that time and has often been used as an evaluation tool for setting up CENDL and for analyzing the measurements of fissile materials. During these years many improvements have been made. In this manual, the format of the input parameter files and the output files, as well as the functions of flag used in FUNF code, are introduced in detail, and the examples of the format of input parameters files are given. FUNF code consists of the spherical optical model, the Hauser-Feshbach model, and the unified Hauser-Feshbach and exciton model. (authors)

  3. Calibration measurements using the ORNL fissile mass flow monitor

    International Nuclear Information System (INIS)

    March-Leuba, J.; Uckan, T.; Sumner, J.; Mattingly, J.; Mihalczo, J.

    1998-01-01

    This paper presents a demonstration of fissile-mass-flow measurements using the Oak Ridge National Laboratory (ORNL) Fissile Mass Flow Monitor in the Paducah Gaseous Diffusion Plant (PGDP). This Flow Monitor is part of a Blend Down Monitoring System (BDMS) that will be installed in at least two Russian Federation (R.F.) blending facilities. The key objectives of the demonstration of the ORNL Flow Monitor are two: (a) demonstrate that the ORNL Flow Monitor equipment is capable of reliably monitoring the mass flow rate of 235 UF 6 gas, and (b) provide a demonstration of ORNL Flow Monitor system in operation with UF 6 flow for a visiting R.F. delegation. These two objectives have been met by the PGDP demonstration, as presented in this paper

  4. IAEA safeguards for the Fissile Materials Disposition Project

    International Nuclear Information System (INIS)

    Close, D.A.

    1995-06-01

    This document is an overview of International Atomic Energy Agency (IAEA) safeguards and the basic requirements or elements of an IAEA safeguards regime. The primary objective of IAEA safeguards is the timely detection of the diversion of a significant quantity of material and the timely detection of undeclared activities. The two important components of IAEA safeguards to accomplish their primary objective are nuclear material accountancy and containment and surveillance. This overview provides guidance to the Fissile Materials Disposition Project for IAEA inspection requirements. IAEA requirements, DOE Orders, and Nuclear Regulatory Commission regulations will be used as the basis for designing a safeguards and security system for the facilities recommended by the Fissile Materials Disposition Project

  5. Fissile material ban: global and non-discriminatory?

    International Nuclear Information System (INIS)

    Datt, Savita

    1995-01-01

    With the indefinite and unconditional extension of the nuclear Non-Proliferation Treaty (NPT) now out of the way, the next issue on the non-proliferation agenda is that of the existing stocks and further production of plutonium and weapons grade uranium. More than the existing stocks and the surplus fissile materials made available through arms control and disarmament measures, it is the further production of such materials which is sought to be tackled urgently. Of prime concern are the nuclear programmes of threshold countries like India, Pakistan and Israel (countries out of the NPT fold) which need to be capped at all costs. The best method of achieving this, it is believed can be through a global ban on the production of fissile materials. 15 refs

  6. Combustion of biodiesel fuel produced from hazelnut soapstock/waste sunflower oil mixture in a Diesel engine

    International Nuclear Information System (INIS)

    Usta, N.; Oeztuerk, E.; Can, Oe.; Conkur, E.S.; Nas, S.; Con, A.H.; Can, A.C.; Topcu, M.

    2005-01-01

    Biodiesel is considered as an alternative fuel to Diesel fuel No. 2, which can be generally produced from different kinds of vegetable oils. Since the prices of edible vegetable oils are higher than that of Diesel fuel No. 2, waste vegetable oils and non-edible crude vegetable oils are preferred as potential low priced biodiesel sources. In addition, it is possible to use soapstock, a by-product of edible oil production, for cheap biodiesel production. In this study, a methyl ester biodiesel was produced from a hazelnut soapstock/waste sunflower oil mixture using methanol, sulphuric acid and sodium hydroxide in a two stage process. The effects of the methyl ester addition to Diesel No. 2 on the performance and emissions of a four cycle, four cylinder, turbocharged indirect injection (IDI) Diesel engine were examined at both full and partial loads. Experimental results showed that the hazelnut soapstock/waste sunflower oil methyl ester can be partially substituted for the Diesel fuel at most operating conditions in terms of the performance parameters and emissions without any engine modification and preheating of the blends

  7. Engine performance and emission characteristics of plastic oil produced from waste polyethylene and its blends with diesel fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yin, Sudong; Tan, Zhongchao [Department of Mechanical and Mechatronics Engineering, University of Waterloo (Canada)], Email: tanz@uwaterloo.ca

    2011-07-01

    This paper describes an experiment to determine the possibility of transforming waste plastics into a potential source of diesel fuel. Experiments were done on the use of various blends of plastic oil produced from waste polyethylene (WPE) with diesel fuel (D) at different volumetric ratios and the results were reviewed. WPE was thermally degraded with catalysis of sodium aluminum silicate at optimum conditions (414-480 degree celsius range and 1 h reaction time) and the collected oil was fractionated at various temperatures. The properties of the fuel blends at different volumetric ratios were measured in this study. It was shown that these blends can be used as fuel in compression ignition engines without any modification. With respect to engine performance and exhaust emission, it was found that using a 5% WPE-D (WPE5) blend instead of diesel fuel reduced carbon monoxide (CO) emission. However, the results of experiment showed that carbon dioxide (CO2) emission and oxides of nitrogen (NOx) emission rose.

  8. Risk-constrained self-scheduling of a fuel and emission constrained power producer using rolling window procedure

    International Nuclear Information System (INIS)

    Kazempour, S. Jalal; Moghaddam, Mohsen Parsa

    2011-01-01

    This work addresses a relevant methodology for self-scheduling of a price-taker fuel and emission constrained power producer in day-ahead correlated energy, spinning reserve and fuel markets to achieve a trade-off between the expected profit and the risk versus different risk levels based on Markowitz's seminal work in the area of portfolio selection. Here, a set of uncertainties including price forecasting errors and available fuel uncertainty are considered. The latter uncertainty arises because of uncertainties in being called for reserve deployment in the spinning reserve market and availability of power plant. To tackle the price forecasting errors, variances of energy, spinning reserve and fuel prices along with their covariances which are due to markets correlation are taken into account using relevant historical data. In order to tackle available fuel uncertainty, a framework for self-scheduling referred to as rolling window is proposed. This risk-constrained self-scheduling framework is therefore formulated and solved as a mixed-integer non-linear programming problem. Furthermore, numerical results for a case study are discussed. (author)

  9. Fissile interrogation using gamma rays from oxygen

    Science.gov (United States)

    Smith, Donald; Micklich, Bradley J.; Fessler, Andreas

    2004-04-20

    The subject apparatus provides a means to identify the presence of fissionable material or other nuclear material contained within an item to be tested. The system employs a portable accelerator to accelerate and direct protons to a fluorine-compound target. The interaction of the protons with the fluorine-compound target produces gamma rays which are directed at the item to be tested. If the item to be tested contains either a fissionable material or other nuclear material the interaction of the gamma rays with the material contained within the test item with result in the production of neutrons. A system of neutron detectors is positioned to intercept any neutrons generated by the test item. The results from the neutron detectors are analyzed to determine the presence of a fissionable material or other nuclear material.

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

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1998-01-01

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

  11. Development of AGNES, a kinetics code for fissile solutions, 1

    International Nuclear Information System (INIS)

    Nakajima, Ken; Ohnishi, Nobuaki

    1986-01-01

    A kinetics code for fissile solutions, AGNES (Accidentally Generated Nuclear Excursion Simulation code), has been developed. This code calculates the radiolytic gas void effect as a reactivity feedback. Physical and calculative models of the radiolytic gas void are summarized and the usage of AGNES is described. In addition, some benchmark calculations were performed and results of calculations show good agreement with those of experiments. (author)

  12. Rationale for continuing R&D in direct coal conversion to produce high quality transportation fuels

    Energy Technology Data Exchange (ETDEWEB)

    Srivastava, R.D.; McIlvried, H.G. [Burns and Roe Services Corp., Pittsburgh, PA (United States); Gray, D. [Mitre Corp, McLean, VA (United States)] [and others

    1995-12-31

    For the foreseeable future, liquid hydrocarbon fuels will play a significant role in the transportation sector of both the United States and the world. Factors favoring these fuels include convenience, high energy density, and the vast existing infrastructure for their production and use. At present the U.S. consumes about 26% of the world supply of petroleum, but this situation is expected to change because of declining domestic production and increasing competition for imports from countries with developing economies. A scenario and time frame are developed in which declining world resources will generate a shortfall in petroleum supply that can be allieviated in part by utilizing the abundant domestic coal resource base. One option is direct coal conversion to liquid transportation fuels. Continued R&D in coal conversion technology will results in improved technical readiness that can significantly reduce costs so that synfuels can compete economically in a time frame to address the shortfall.

  13. Hardware implementation of the ORNL fissile mass flow monitor

    International Nuclear Information System (INIS)

    McEvers, J.; Sumner, J.; Jones, R.; Ferrell, R.; Martin, C.; Uckan, T.; March-Leuba, J.

    1998-01-01

    This paper provides an overall description of the implementation of the Oak Ridge National Laboratory (ORNL) Fissile Mass Flow Monitor, which is part of a Blend Down Monitoring System (BDMS) developed by the US Department of Energy (DOE). The Fissile Mass Flow Monitor is designed to measure the mass flow of fissile material through a gaseous or liquid process stream. It consists of a source-modulator assembly, a detector assembly, and a cabinet that houses all control, data acquisition, and supporting electronics equipment. The development of this flow monitor was first funded by DOE/NE in September 95, and an initial demonstration by ORNL was described in previous INMM meetings. This methodology was chosen by DOE/NE for implementation in November 1996, and the hardware/software development is complete. Successful BDMS installation and operation of the complete BDMS has been demonstrated in the Paducah Gaseous Diffusion Plant (PGDP), which is operated by Lockheed Martin Utility Services, Inc. for the US Enrichment Corporation and regulated by the Nuclear Regulatory Commission. Equipment for two BDMS units has been shipped to the Russian Federation

  14. Feasibility of Producing and Using Biomass-Based Diesel and Jet Fuel in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Milbrandt, A. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Kinchin, C. [National Renewable Energy Lab. (NREL), Golden, CO (United States); McCormick, R. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2013-12-01

    The study summarizes the best available public data on the production, capacity, cost, market demand, and feedstock availability for the production of biomass-based diesel and jet fuel. It includes an overview of the current conversion processes and current state-of-development for the production of biomass-based jet and diesel fuel, as well as the key companies pursuing this effort. Thediscussion analyzes all this information in the context of meeting the RFS mandate, highlights uncertainties for the future industry development, and key business opportunities.

  15. Storage capacity for fissile material as a function of facility shape (room length-to-width ratio)

    International Nuclear Information System (INIS)

    Altschuler, S.J.

    1975-01-01

    The results of a previous study for applying surface density methods to square room of varying size are shown to be conservative for rectangular rooms as well. The surface density required to produce criticality has been calculated as a function of the facility length-to-width ratio for a variety of room widths and unit sizes, shapes, and fissile material compositions. For a length to width ratio greater than or equal to 6, the critical surface density is essentially constant. This allows further economies since more fissile material can be stored at a given subcritical value of k/ sub eff/(0.90) in a rectangular vault of given usable area than in a square one. (U.S.)

  16. Process and plant for obtaining producer gas from fossil fuels. Verfahren und Anlage zur Gewinnung von Generatorgas aus fossilen Brennstoffen

    Energy Technology Data Exchange (ETDEWEB)

    Anon,

    1983-12-01

    In a plant for generating producer gas from fossil fuels with relatively high humidity, there is predrying of the wet material in two drying chambers situated above the actual reactor shaft. The drying air required for this purpose is drawn off via blowers and heat exchangers preheated from the area of the combustion zone. The preparation of the crude gases produced first in the process is done by a socalled bypass gas system, i.e. the reintroduction of the crude gases enriched with tar oil and steam and diverting prepared hot gases via an annular pipe from the area of the reduction zone.

  17. Fuel utilization in a progressive conversion reactor (PCR)

    International Nuclear Information System (INIS)

    Leyse, C.F.; Judd, J.L.

    1981-05-01

    Preliminary studies indicate that for once-through fuel cycles, the PCR offers potential improvements over current LWRs in the following major areas: improved uranium utilization (reduced uranium demand), degraded plutonium product in spent fuel, reduced plutonium content of spent fuel, reduced amount of spent fuel, reduced fissile content of spent fuel, and reduced separative work

  18. The simultaneous neutron and photon interrogation method for fissile and non-fissile element separation in radioactive waste drums

    International Nuclear Information System (INIS)

    Jallu, F.; Lyoussi, A.; Passard, C.; Payan, E.; Recroix, H.; Nurdin, G.; Buisson, A.; Allano, J.

    2000-01-01

    Measuring α-emitters such as ( 234,235,236,238 U, 238,239,240,242,244 Pu, 237 Np, 241,243 Am, ...), in solid radioactive waste allows us to quantify the α-activity in a drum and then to classify it. The simultaneous photon and neutron interrogation experiment (SIMPHONIE) method dealt with in this paper, combines both active neutron interrogation and induced photofission interrogation techniques simultaneously. Its purpose is to quantify fissile ( 235 U, 239,241 Pu, ...) and non-fissile ( 236,238 U, 238,240 Pu, ...) elements separately in only one measurement. This paper presents the principle of the method, the experimental setup, and the first experimental results obtained using the DGA/ETCA Linac and MiniLinatron pulsed linear electron accelerators located at Arcueil, France. First studies were carried out with U and Pu bare samples

  19. Copper produced from powder by HIP to encapsulate nuclear fuel elements

    International Nuclear Information System (INIS)

    Ekbom, L.B.; Bogegaard, S.

    1989-02-01

    In the Swedish nuclear waste mangement program, nuclear fuel elements are proposed to be encapsulated in copper canisters. To fill the space between the fuel elements two methods have been proposed. Originally lead was proposed to be cast into the canister. According to a second method the space between the fuel rods is filled with copper powder and hot isostatic pressed (HIP) to seal the canister lid and to densify the powder to homogenous copper. This latter method has the advantage that each fuel rod is individually encapsulated in a very corrosion resistant material. This investigation was performed to find out to what extent pure copper powder can be hot isosatic pressed to full density and to achieve properties comparable to that of the oxygen free high conductivity (OFHC) copper of the canister. OFHC copper was molten under helium gas protection and atomized to a fine spherical powder in a pilot plant. The powder was transfered to a glove box with an argon atmosphere. The powder was filled into a steel container, which was evacuated and sealed. HIP was done at 550 degree C and 200 MPa for one hour. The resulting copper was found to have a good ductility and mechanical properties comparable to that of ordinary copper. The constant strainrate stress corrosion test used to test the canister copper showed that the HIP-ed copper has the same good properties as OFHC copper. (authors)

  20. Hydrothermal Conversion in Near-Critical Water – A Sustainable Way of Producing Renewable Fuels

    DEFF Research Database (Denmark)

    Hoffmann, Jessica; Pedersen, Thomas Helmer; Rosendahl, Lasse

    2014-01-01

    Liquid fuels from biomass will form an essential part of meeting the grand challenges within energy. The need for renewable and sustainable energy sources is triggered by a number of factors; like increase in global energy demand, depletion of conventional resources, climate issues and the desire...... hydrothermal conversion of lignocellulosic biomass and upgrading pathways of bio-crude components with focus on hydrodeoxygenation reactions....

  1. Power generation versus fuel production in light water hybrid reactors

    International Nuclear Information System (INIS)

    Greenspan, E.

    1977-06-01

    The economic potentials of fissile-fuel-producing light-water hybrid reactors (FFP-LWHR) and of fuel-self-sufficient (FSS) LWHR's are compared. A simple economic model is constructed that gives the capital investment allowed for the hybrid reactor so that the cost of electricity generated in the hybrid based energy system equals the cost of electricity generated in LWR's. The power systems considered are LWR, FSS-LWHR, and FFP-LWHR plus LWR, both with and without plutonium recycling. The economic potential of FFP-LWHR's is found superior to that of FSS-LWHR's. Moreover, LWHR's may compete, economically, with LWR's. Criteria for determining the more economical approach to hybrid fuel or power production are derived for blankets having a linear dependence between F and M. The examples considered favor the power generation rather than fuel production

  2. The role of accelerators in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Takahashi, Hiroshi

    1990-01-01

    The use of the neutrons produced by medium energy proton accelerators (1-3 GeV) has the considerable potential in reconstructing the nuclear fuel cycle. About 1.5 - 2.5 t of fissile material can be produced annually by injecting a 450 MW proton beam directly into fertile materials. A source of neutrons produced by a proton beam to supply subcritical reactors could alleviate many of the safety problems associated with critical assemblies. It is worthwhile to study an alternative approach to store the waste that would separate long-lived nuclei from high level waste by transmuting them into short-lived or nonradioactive waste. The small beam power of 15-30 MW can incinerate the actinide produced by ten 1 GWe light water reactors. Moreover, an incinerator with 900 MW thermal power can produce 270-240 MWe excess electricity and 100 kg of fissile material by surrounding the core with fertile materials. Accelerator breeders, actinide incinerators, particle fuel suitable to these purposes, the incineration of Cs-137 and Sr-90 fission products and future accelerator technology are described. Plasma beat waves and wake fields, and laser technology are the next steps of development. (K.I.)

  3. Evaluation of criticality criteria for fissile class II packages in transportation

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1976-01-01

    The nuclear criticality safety of packages in transportation is explored systematically by a surface density representation of reflected array criticality of air-spaced units. Typical perturbations to arrays are shown to be related analytically to the corresponding reactivity changes they produce. The reactivity change associated with the removal of three reflecting surfaces from a totally water reflected array is shown to depend upon the fissile material loading of the packages. For U(93.2) metal, the expected reactivity loss can range from 2 to 21%. Replacement of a three-sided reflector of water on a critical array by one of concrete results in a reactivity increase ranging from 0 to 6%. Mass limits established by criticality data for reflected arrays of air-spaced units can provide a minimum, uniform margin of safety, expressible in terms of reactivity, to more reliably specify subcriticality in transport. Mass limits less than those defined by air-spaced units in water-reflected arrays are unnecessary for Fissile Class II packages. (author)

  4. Contributions at the Tripoli Monte Carlo code qualifying on critical experiences and at neutronic interaction study of fissile units

    International Nuclear Information System (INIS)

    Nouri, A.

    1994-01-01

    Criticality studies in nuclear fuel cycle are based on Monte Carlo method. These codes use multigroup cross sections which can verify by experimental configurations or by use of reference codes such Tripoli 2. In this Tripoli 2 code nuclear data are errors attached and asked for experimental studies with critical experiences. This is one of the aim of this thesis. To calculate the keff of interacted fissile units we have used the multigroup Monte Carlo code Moret with convergence problems. A new estimator of reactions rates permit to better approximate the neutrons exchange between units and a new importance function has been tested. 2 annexes

  5. Alternative fuel produced from thermal pyrolysis of waste tires and its use in a DI diesel engine

    International Nuclear Information System (INIS)

    Wang, Wei-Cheng; Bai, Chi-Jeng; Lin, Chi-Tung; Prakash, Samay

    2016-01-01

    Highlights: • The liquid, solid and gas yields from pyrolysis of waste tires were investigated. • For energy and economic consideration, pre-treatments of TPO were avoided. • Various proportions of TPO-diesel mixture were tested in a DI diesel engine. • TPOs derived from various pyrolysis temperatures were also tested in engine. • Fuel consumption, cylinder pressure, engine power, and SO2 emission were discussed. - Abstract: Alternative fuels from waste material have been receiving attentions due to the increasing demand of fossil fuels. Pyrolysis has been a considerable solution for processing waste tires because it gives clean emissions and produces valuable liquid or solid products. Pyrolysis oil from waste tires has become a potential replacement for petroleum diesel due to the similar physical and chemical properties to diesel fuel. In this study, waste tires were pyrolyzed in a lab-scale fixed bed reactor with various reaction temperatures. The liquid, solid and gas product yields from different pyrolysis temperatures were compared, as well as the analyses of property and element for the oil product. Due to the energy and economic consideration, the pre-treatments of TPO before adding into regular diesel were avoided. The TPO derived from various pyrolysis temperatures were mixed with regular diesel at different proportions and subsequently tested in a DI diesel engine. The engine performance, such as fuel consumption, cylinder pressure, engine power, and SO_2 emission, were examined and discussed. The results indicated that increasing the TPO fraction in diesel lead to worse engine performance, but it can be recovered using TPOs produced from higher pyrolysis temperatures.

  6. Screening of IAEA environmental samples for fissile material content

    International Nuclear Information System (INIS)

    Hembree, Doyle M. Jr.; Carter, Joel A.; Devault, Gerald L.; Whitaker, J. Michael; Glasgow, David

    2001-01-01

    Full text: Analysis of environmental samples for the International Atomic Energy Agency (IAEA) Strengthened Safeguards Systems program requires that stringent measures be taken to control contamination. To facilitate contamination control, it is extremely useful to have some estimate of the fissile content of a given sample prior to beginning sample preparation and analysis. This is particularly true for laboratories that employ clean rooms during sample preparation. A review of the analytical results for samples submitted between January 1, 1999 and September 1, 2000 revealed that the total uranium content values ranged from 0.2 to greater than 500,000 ng/sample. Poor estimates of the uranium or plutonium content in the samples have caused some of the laboratories in the IAEA Network of Analytical Laboratories (NWAL) to experience clean laboratory contamination, sample cross contamination, and non-ideal uranium spike additions. This has led to significant increases in analysis costs (e.g., recertification of clean rooms after removing contamination, and rerunning samples) and degradation in data quality. A number of methods have been proposed for screening environmental samples for fissile material content, including gamma spectrometry, x-ray fluorescence, kinetic phosphorimetry (KPA), and inductively coupled plasma-mass spectrometry (ICP-MS). Gamma spectrometry and x-ray fluorescence are suitable for screening samples with microgram or greater quantities of uranium. ICP-MS and KPA are used successfully in some DOE NWAL laboratories to screen environmental samples. A neutron activation analysis (NAA) method that offers numerous advantages over other screening techniques for environmental samples has recently been proposed. Fissile materials such as 239 Pu and 235 U can be made to undergo fission in the intense neutron field to which they are exposed during neutron activation analysis (NAA). Some of the fission products emit neutrons referred to as 'delayed

  7. Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

    Science.gov (United States)

    Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

    2017-07-01

    Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

  8. Nuclear fuels and development of nuclear fuel elements

    International Nuclear Information System (INIS)

    Sundaram, C.V.; Mannan, S.L.

    1989-01-01

    Safe, reliable and economic operation of nuclear fission reactors, the source of nuclear power at present, requires judicious choice, careful preparation and specialised fabrication procedures for fuels and fuel element structural materials. These aspects of nuclear fuels (uranium, plutonium and their oxides and carbides), fuel element technology and structural materials (aluminium, zircaloy, stainless steel etc.) are discussed with particular reference to research and power reactors in India, e.g. the DHRUVA research reactor at BARC, Trombay, the pressurised heavy water reactors (PHWR) at Rajasthan and Kalpakkam, and the Fast Breeder Test Reactor (FBTR) at Kalpakkam. Other reactors like the gas-cooled reactors operating in UK are also mentioned. Because of the limited uranium resources, India has opted for a three-stage nuclear power programme aimed at the ultimate utilization of her abundant thorium resources. The first phase consists of natural uranium dioxide-fuelled, heavy water-moderated and cooled PHWR. The second phase was initiated with the attainment of criticality in the FBTR at Kalpakkam. Fast Breeder Reactors (FBR) utilize the plutonium and uranium by-products of phase 1. Moreover, FBR can convert thorium into fissile 233 U. They produce more fuel than is consumed - hence, the name breeders. The fuel parameters of some of the operating or proposed fast reactors in the world are compared. FBTR is unique in the choice of mixed carbides of plutonium and uranium as fuel. Factors affecting the fuel element performance and life in various reactors e.g. hydriding of zircaloys, fuel pellet-cladding interaction etc. in PHWR and void swelling; irradiation creep and helium embrittlement of fuel element structural materials in FBR are discussed along with measures to overcome some of these problems. (author). 15 refs., 9 tabs., 23 figs

  9. The potential of using organic side-streams produced in Ghana for generation of bio-fuel

    International Nuclear Information System (INIS)

    Laryea, G. N; Abdul-Samii, R.; Tottimeh, G.

    2014-01-01

    Bio-fuel can be generated from organic side-streams of maize, rice, millet, sorghum and groundnut by using fast pyrolysis technology. Data on side-streams of these crops were obtained from the Ministry of Food and Agriculture (MoFA) in 2010 for the study. The study shows that the estimated total crop side-streams generated was 3,475,413 t of which 2,345,903.5 of bio-fuel can be produced, given a potential energy equivalent of 42,226 PJ/y. The result shows a growth rate of 12.9 per cent in energy equivalent potential for synthetic fuel production as compared to the estimated production in 2009. Northern Region had the highest energy potential of 9,676 PJ/y (22.91%) of the total energy equivalent of bio-fuel, whereas, Greater Accra Region had the lowest with 183 PJ/y (0.43%). It is recommended that the available energy potential at the three northern regions of Ghana be utilised effectively when renewable energy policy is improved for a wider applications of side-streams from crops.(au)

  10. Fossil fuel reform in developing states: The case of Trinidad and Tobago, a petroleum producing small Island developing State

    International Nuclear Information System (INIS)

    Scobie, Michelle

    2017-01-01

    Trinidad and Tobago is an oil exporting small island developing state (SIDS) with a 0.12% contribution to global emissions and with important socio-economic challenges. It has producer, electricity and transport fuel subsidies. It is at an interesting juncture in subsidy reform: the government faces the embeddedness of distributive justice norms that are contested by fiscal prudence and environmental stewardship norms. The value of the paper is twofold. First it develops a subsidy intractability framework to explain reform global narratives that highlights: the power of agents, the nature of contested economic, justice and environmental norms and the availability of mechanisms for reform. Second, this framework is used to explain reform narratives and trajectories in Trinidad and Tobago using data from public documents and from a unique elite survey of former and present heads of state, politicians, policy makers and stakeholders. Even in conditions of falling oil prices and national revenue and pressures to reduce emissions, where redistributive justice arguments are heavily embedded in public discourses, those aspects of the subsidy that have developmental or distributive justice goals are more intractable. The results of the study have implications for carbon emission reduction strategies in developing states with fossil fuel reserves. - Highlights: • A subsidy intractability framework is used to analyse fuel subsidy reform. • A sense of entitlement to resources contributes to subsidy intractability. • Global environmental stewardship norms matter less for fuel subsidy reform in SIDS. • Policy space is most determined by international economic conditions in SIDS.

  11. A novel CO2 sequestration system for environmentally producing hydrogen from fossil-fuels

    International Nuclear Information System (INIS)

    Eucker IV, W.

    2007-01-01

    Aqueous monoethanolamine (MEA) scrubbers are currently used to capture carbon dioxide (CO 2 ) from industrial flue gases in various fossil-fuel based energy production systems. MEA is a highly volatile, corrosive, physiologically toxic, and foul-smelling chemical that requires replacement after 1000 operational hours. Room temperature ionic liquids (RTILs), a novel class of materials with negligible vapor pressures and potentiality as benign solvents, may be the ideal replacement for MEA. Ab initio computational modeling was used to investigate the molecular interactions of ILs with CO 2 . The energetic and thermodynamic parameters of the RTILs as CO 2 solvents are on par with MEA. As viable competitors to the present CO 2 separation technology, RTILs may economize the fossil-fuel decarbonization process with the ultimate aim of realizing a green hydrogen economy

  12. Is the biochar produced from sewage sludge a good quality solid fuel?

    Directory of Open Access Journals (Sweden)

    Pulka Jakub

    2016-12-01

    Full Text Available The influence of sewage sludge torrefaction temperature on fuel properties was investigated. Non-lignocellulosic waste thermal treatment experiment was conducted within 1 h residence time, under the following temperatures: 200, 220, 240, 260, 280 and 300°C. Sawdust was used as lignocellulosic reference material. The following parameters of biochar have been measured: moisture, higher heating value, ash content, volatile compounds and sulfur content. Sawdust biochar has been confirmed to be a good quality solid fuel. High ash and sulfur content may be an obstacle for biochar energy reuse. The best temperature profile for sawdust torrefaction and fuel production for 1 h residence time was 220°C. At this temperature the product contained 84% of initial energy while decreased the mass by 25%. The best temperature profile for sewage sludge was 240°C. The energy residue was 91% and the mass residue was 85%. Higher temperatures in both cases caused excessive mass and energy losses.

  13. Study of relationships between microstructures and service properties, of U(Mo) fissile alloys particles

    International Nuclear Information System (INIS)

    Champion, G.

    2013-01-01

    This thesis enters in the Material and Testing Reactors (MTRs) framework where the necessity to use a Low- Enriched Uranium (LEU) fuel has led to the development of a dense fissile material based on U(Mo) alloys. The designed fuel is a composite material, made of dispersed U(Mo) particles embedded in an Al based matrix. Post- Irradiation Examinations of these LEU fuel plates showed that the irradiation behaviour of the fuel is not fit for purpose yet. This is mainly due to the growth of an interaction layer between the fuel and the matrix and to the bad gas retention efficiency of the fuel particles. This thesis had for purpose the development of several solutions in order to modify and/or decrease or even inhibit the fuel/matrix interaction and to increase the gas retention capacities of the fuel. In order to achieve so, two solutions have been tested during this thesis, (i) optimization of the U(Mo) alloy intrinsic microstructural properties and (ii) modification of the fuel meat/matrix interface, through the deposition of a layer acting as a 'diffusion barrier'. Concerning the first axis of study, a characterization campaign of the reference powders has been performed, as a first step, in order to identify the key parameters for the development of products showing an 'optimized' microstructure. Two novel products have then been developed: one based on a combined process associating 'atomization + grinding' and another, which consists in a magnesiothermy process. These products were subjected to characterization: X-Ray and neutron diffraction, electron backscattered diffraction and transmission electron microscopy have been performed in particular. We managed to show that these powders can be an advantage concerning the issue with the gas retention capacities of the fuel. Concerning the growth of the interaction layer, a third product has been developed: an U(Mo) atomized powder, coated with an alumina layer. We managed to show that a thickness between 100 and

  14. Report of Post Irradiation Examination for Dry Process Fuel

    International Nuclear Information System (INIS)

    Par, Jang Jin; Jung, I. H.; Kang, K. H.; Moon, J. S.; Lee, C. R.; Ryu, H. J.; Song, K. C.; Yang, M. S.; Yoo, B. O.; Jung, Y. H.; Choo, Y. S.

    2006-08-01

    The spent PWR fuel typically contains 0.9 wt.% of fissile uranium and 0.6 wt.% of fissile plutonium, which exceeds the natural uranium fissile content of 0.711 wt.%. The neutron economy of a CANDU reactor is sufficient to utilize the DUPIC fuel, even though the neutron-absorbing fission products contained in the spent PWR fuel were remained in the DUPIC fuel. The DUPIC fuel cycle offers advantages to the countries operating both the PWR and CANDU reactors, such as saving the natural uranium, reducing the spent fuel in both PWR and CANDU, and acquiring the extra energy by reuse of the PWR spent fuel. This report contains the results of post-irradiation examination of the DUPIC fuel irradiated four times at HANARO from May 2000 to August 2006 present except the first irradiation test of simulated DUPIC fuel at HANARO on August 1999

  15. Quality assurance for breeder reactor fuel fabrication

    International Nuclear Information System (INIS)

    Marx, E.R.

    1978-01-01

    Fuel performance in the Fast Flux Test Facility (FFTF) depends on fabrication of fuel to rigorous quality standards. The quality program including Management, Procurement, Fabrication, Inspection, Records, and Audits is discussed as well as unique mixed oxide fuel inspections such as homogeneity inspection, analytical chemistry, and nondestructive fissile assay

  16. Sewage sludge based producer gas of rich H{sub 2} content as a fuel for an IC engine

    Energy Technology Data Exchange (ETDEWEB)

    Szwaja, Stanislaw; Cupial, Karol [Czestochowa Univ. of Technology (Poland)

    2010-07-01

    The manuscript presents investigation on hydrogen rich gas combustion in an internal combustion (IC) engine. The gas is obtained from gasification process of sewage sludge which is by-product of waste water treatment in a municipal sewage treatment plant. Recently introduced EU regulations of environmental protection do not allow to use such sludge as a soil fertilizer or substance for landfilling the ground due to its biological toxicity. On another hand, this sludge contains organic content of approximately 45-55% and from this point of view the sludge looks as an attractive material for fuel production through its gasification. This technology, primarily applied for wood gasification, has been also successfully implemented for gasification of sludge. It was found that the producer gas obtained in this way is rich of hydrogen content even up to 25%. This is because of high water content in the sludge that provides favorable conditions for steam reforming resulting in increase of hydrogen in the products of gasification. The high hydrogen content in the producer gas can lead to improper combustion particularly when the combustion takes place in the internal combustion engine. That improper combustion might appear as combustion knock and it is the main problem for the engine in which hydrogen is used as a fuel [1]. Onset of the knock during combustion contributes to rapid increase in heat transfer to the piston crown causing the piston to be quickly overheated that leads to surface erosion and damages. Additionally, engine body vibration coming from the knock significantly shortens engine durability. Conclusions from this investigation provide good premises for combusting the sludge producer gas in the IC engine without any improper combustion anomalies, thus considers this gas as worthy fuel for a stationary engine driven a power generator. The presentation shows results of producer gas combustion in both the spark-ignited and the compression ignition engine with

  17. The effects of the evolution of fuel prices and the environmental regulations on the producers of electric power based on fossil fuel

    International Nuclear Information System (INIS)

    Balasoiu, Constantin; Alecu, Sorin

    2006-01-01

    The production of electric power in the context of the concept of human society's lasting development is influenced in the recent years by a series of external factors, both circumstantial and derived from internal and international regulations. This work proposes a theoretical analysis of additional costs induced by the evolution of fuel prices as well as of the short, medium and long term environmental restrictions for the producers of lignite based electric power in Romania. To this purpose, the authors have considered as theoretical elements of analysis, a 330 MW functioning power station, working entirely on lignite GEL (70% expenses on fuel) with a production cost of 40 Euros/MWh at a 70% degree of usage capacity and 36 Euros/MWh at 100%. The paper addresses the following items: 1. The periods of analysis and the influential factors; 2. The evaluation of additional costs for the observance of EU Directive 2001/80/EC; 3. The evaluation of additional costs induced by the stipulations of the Kyoto Protocol; 4. The evaluation of additional costs induced by the evolution of the price of the fuel. In conclusion accumulating all the influences described in the chapters of this material, the impact in the rise of production costs for the described lignite based power plant is summarized by taking into account: the impact of CO 2 emissions; the impact Directive 2001/80/EC; the impact of the fuel price; the total rise. One can notice, that the biggest influence on the additional production costs comes from the impact of CO 2 emissions, in the outlook of the integration in the EU ETS, which depends on: 1) The way in which the National Allocation Plan for the allowances of CO 2 emissions is made in the power sector. The higher D utl.ref is, the stronger will be their place on the market. 2) The evolution of the price of CO 2 emissions on the EU ETS

  18. Canyon transfer neutron absorber to fissile material ratio analysis. Revision 1

    International Nuclear Information System (INIS)

    Clemmons, J.S.

    1994-01-01

    Waste tank fissile material and non-fissile material estimates are used to evaluate criticality safety for the existing sludge inventory and batches of sludge sent to Extended Sludge Processing (ESP). This report documents the weight ratios of several non-fissile waste constituents to fissile waste constituents from canyon reprocessing waste streams. Weight ratios of Fe, Mn, Al, Mi, and U-238 to fissile material are calculated from monthly loss estimates from the F and H Canyon Low Heat Waste (LHW) and High Heat Waste (HHW) streams. The monthly weight ratios for Fe, Mn and U-238 are then compared to calculated minimum safe weight ratios. Documented minimum safe weight ratios for Al and Ni to fissile material are currently not available. Total mass data for the subject sludge constituents is provided along with scatter plots of the monthly weight ratios for each waste stream

  19. Development of anionic membranes produced by radiation-grafting for alkaline fuel cell applications

    International Nuclear Information System (INIS)

    Pereira, Clotilde Coppini

    2017-01-01

    Anion Exchange Membranes (AEMs) are a promising alternative to the development of more efficient electrolytes for alkaline fuel cells. In general, the AEMs are ionomeric membranes able to conduct hydroxide ions (OH - ) due to the quaternary ammonium groups, which confer high pH equivalent to the AEM. In order to develop alkaline membranes with high chemical and thermal stability, besides satisfactory ionic conductivity for alkaline fuel cells, membranes based on low density polyethylene (LDPE), ultrahigh weight molecular weight polyethylene (UHWHPE), poly(ethylene-co-tetrafluoroethylene) (PETFE) and poly(hexafluoropropylene-co-tetrafluoroethylene) (PFEP) previously irradiated by using 60 Co gamma and electron beam sources, have been synthesized by styrene-grafting, and functionalized with trimethylamine to introduced quaternary ammonium groups. The resulting membranes were characterized by electron paramagnetic resonance (EPR), Raman spectroscopy, thermogravimetry (TG) and electrochemical impedance spectroscopy (EIS). The determination of the grafting degree and water uptake were conducted by gravimetry and ion exchange capacity, by titration. The membranes synthesized with PELD and PEUHMW polymers pre-irradiated at 70 kGy and stored at low temperature (-70 deg C), up to 10 months, showed ionic conductivity results, in hydroxide form (OH - ), of 29 mS.cm -1 and 14 mS.cm -1 at 65 deg C, respectively. The PFEP polymers irradiated by the simultaneous process showed insufficient grating levels for the membrane synthesis, requiring more studies to improve the irradiation and grafting process. The styrene-grafted PETFE membranes, pre-irradiated at 70 kGy and stored at low temperature (-70 deg C), up to 10 months, showed ionic conductivity results, in hydroxide form (OH - ), of 90 mS.cm -1 to 165 mS.cm -1 , in the temperature range 30 to 60 deg C. Such results have demonstrated that LDPE, UHMWPE and PETFE based AEMs are promising electrolytes for alkaline fuel cell

  20. Research priorities in bioconversion of municipal solid waste to produce chemicals, liquid and gaseous fuels

    Energy Technology Data Exchange (ETDEWEB)

    Coombs, J. [BABA Ltd., Reading (United Kingdom)

    1988-09-01

    Areas for future research on the bioconversion of municipal solid wastes are highlighted in order to optimise the potential use of this resource to make chemical, liquid and gaseous fuels. Despite widespread research, a biological understanding of bioconversion technologies, including landfill gas, composting and anaerobic digestion, has yet to be established. Specifically, work on the development and growth of microorganisms in uncontrolled systems and the detailed biochemistry of purified strains needs to be undertaken. The microbial breakdown of xenobiotics to clean up polluted sites, and as an alternative to incineration of toxic organic wastes, is viewed as a desirable outcome of such an understanding. (UK)

  1. Fossil fuel produced radioactivities and their effect on the food chain (II)

    International Nuclear Information System (INIS)

    Okamoto, K.

    1982-01-01

    The effects of radioactivities released from fossil fuel burning are examined. Main radioactivities are 210 Pb and 210 Po. Revised values of the dose due to the intake of leafy vegetables and seafoods are presented. The dose from natural gas from the Northern Sea is shown to be much lower than the dose from coal. This conclusion can probably apply to other natural gas except for that from the North American continent. The dose due to coal burning is found to be much higher than that due to marine disposal of nuclear waste

  2. Fossil fuel produced radioactivities and their effect on the food chain (II)

    Energy Technology Data Exchange (ETDEWEB)

    Okamoto, K [New South Wales Univ., Kensington (Australia). Dept. of Applied Mathematics

    1982-03-01

    The effects of radioactivities released from fossil fuel burning are examined. Main radioactivities are /sup 210/Pb and /sup 210/Po. Revised values of the dose due to the intake of leafy vegetables and seafoods are presented. The dose from natural gas from the Northern Sea is shown to be much lower than the dose from coal. This conclusion can probably apply to other natural gas except for that from the North American continent. The dose due to coal burning is found to be much higher than that due to marine disposal of nuclear waste.

  3. Neutronic studies of fissile and fusile breeding blankets

    International Nuclear Information System (INIS)

    Taczanowski, S.

    1984-08-01

    In light of the need of convincing motivation substantiating expensive and inherently applied research (nuclear energy), first a simple comparative study of fissile breeding economics of fusion fission hybrids, spallators and also fast breeder reactors has been carried out. As a result, the necessity of maximization of fissile production (in the first two ones, in fast breeders rather the reprocessing costs should be reduced) has been shown, thus indicating the design strategy (high support ratio) for these systems. In spite of the uncertainty of present projections onto further future and discrepancies in available data even quite conservative assumptions indicate that hybrids and perhaps even earlier - spallators can become economic at realistic uranium price increase and successfully compete against fast breeders. Then on the basis of the concept of the neutron flux shaping aimed at the correlation of the selected cross-sections with the neutron flux, the indications for the maximization of respective reaction rates has been formulated. In turn, these considerations serve as the starting point for the guidelines of breeding blanket nuclear design, which are as follows: 1) The source neutrons must face the multiplying layer (of proper thickness) of possibly low concentration of nuclides attenuating the neutron multiplication (i.e. structure materials, nongaseous coolants). 2) For the most effective trapping of neutrons within the breeding zone (leakage and void streaming reduction) it must contain an efficient moderator (not valid for fissile breeding blankets). 3) All regions of significant slow flux should contain 6 Li in order to reduce parasite neutron captures in there. (orig./HP)

  4. Contribution to fissile materials transportation in transit storage

    International Nuclear Information System (INIS)

    Silva, Teresinha de Moraes da

    2005-01-01

    The national and international standards for the transportation of fissile materials establish two indexes: Transport Index (Tl) and Criticality Safety Index (ISC). Besides, in non-exclusive transit, the largest of these indexes cannot overtake the value 50. Considering several groups to be transported, the sum of the transportation indexes cannot overtake 200 and the distance between them should be 6 meters This work aimed, as a primary target, to verify when an index is superior to another, in relation to the fissile materials studied, i.e., uranium oxides UO 2 , U 3 O 8 and uranium silicide U 3 Si 2 , taking into account the different enrichment grades. The result found is that the criticality safety index is always greater. As a second goal, it was tried to verify if there is any alteration in the case of these compounds aging process, i.e., alteration in transport index (Tl) due to gamma radiation of daughters radioisotopes in secular equilibrium. No alteration, was verified as the daughters contribution although considerable related to the transport index is very small concerning the criticality safety index. As a third target, it was tried to justify a distance equal to 6 meters, between each group of fissile material. The result showed that, in air media, the distance of 1 meter is sufficient, except for the UO 2 compound at 100% of enrichment, which reaches 2 meter while in the water means the distance of 40cm is enough for the compounds studied. This fact is of great importance when the cost of the necessary area in the transportation and storage is taken into consideration. (author)

  5. Fissile material disposition program final immobilization form assessment and recommendation

    International Nuclear Information System (INIS)

    Cochran, S.G.; Dunlop, W.H.; Edmunds, T.A.; MacLean, L.M.; Gould, T.H.

    1997-01-01

    Lawrence Livermore National Laboratory (LLNL), in its role as the lead laboratory for the development of plutonium immobilization technologies for the Department of Energy's Office of Fissile Materials Disposition (MD), has been requested by MD to recommend an immobilization technology for the disposition of surplus weapons- usable plutonium. The recommendation and supporting documentation was requested to be provided by September 1, 1997. This report addresses the choice between glass and ceramic technologies for immobilizing plutonium using the can-in-canister approach. Its purpose is to provide a comparative evaluation of the two candidate technologies and to recommend a form based on technical considerations

  6. Minimization of actinide waste by multi-recycling of thoriated fuels in the EPR reactor

    Directory of Open Access Journals (Sweden)

    Nuttin A.

    2012-02-01

    Full Text Available The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR has been investigated. If increasing quantities of 238U, the fertile isotope in standard UO2 fuel, are replaced by 232Th, then a greater yield of new fissile material (233U is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that minor actinide and plutonium waste inventories are reduced and hence waste radio-toxicities and decay heats are up to a factor of 20 lower after 103 years. Two innovative fuel types named S90 and S20, ThO2 mixed with 90% and 20% enriched UO2 respectively, are compared as an alternative to standard uranium oxide (UOX and uranium/plutonium mixed oxide (MOX fuels at the longest EPR fuel discharge burn-ups of 65 GWd/t. Fissile and waste inventories are examined, waste radio-toxicities and decay heats are extracted and safety feedback coefficients are calculated.

  7. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  8. Analysis of spent fuel assay with a lead slowing down spectrometer

    International Nuclear Information System (INIS)

    Gavron, A.; Smith, L. Eric; Ressler, Jennifer J.

    2009-01-01

    Assay of fissile materials in spent fuel that are produced or depleted during the operation of a reactor, is of paramount importance to nuclear materials accounting, verification of the reactor operation history, as well as for criticality considerations for storage. In order to prevent future proliferation following the spread of nuclear energy, we must develop accurate methods to assay large quantities of nuclear fuels. We analyze the potential of using a Lead Slowing Down Spectrometer for assaying spent fuel. We conclude that it possible to design a system that will provide around 1% statistical precision in the determination of the 239 Pu, 241 Pu and 235 U concentrations in a PWR spent-fuel assembly, for intermediate-to-high burnup levels, using commercial neutron sources, and a system of 238 U threshold fission detectors. Pending further analysis of systematic errors, it is possible that missing pins can be detected, as can asymmetry in the fuel bundle. (author)

  9. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    International Nuclear Information System (INIS)

    Bromley, B.P.; Edwards, G.W.R.; Sambavalingam, P.

    2016-01-01

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  10. Prediction of small spark ignited engine performance using producer gas as fuel

    Directory of Open Access Journals (Sweden)

    N. Homdoung

    2015-03-01

    Full Text Available Producer gas from biomass gasification is expected to contribute to greater energy mix in the future. Therefore, effect of producer gas on engine performance is of great interest. Evaluation of engine performances can be hard and costly. Ideally, they may be predicted mathematically. This work was to apply mathematical models in evaluating performance of a small producer gas engine. The engine was a spark ignition, single cylinder unit with a CR of 14:1. Simulation was carried out on full load and varying engine speeds. From simulated results, it was found that the simple mathematical model can predict the performance of the gas engine and gave good agreement with experimental results. The differences were within ±7%.

  11. Microstructure Characterization of WCCo-Mo Based Coatings Produced Using High Velocity Oxygen Fuel

    Directory of Open Access Journals (Sweden)

    Serkan Islak

    2015-12-01

    Full Text Available The present study has been carried out in order to investigate the microstructural properties of WCCo-Mo composite coatings deposited onto a SAE 4140 steel substrate by high velocity oxygen fuel (HVOF thermal spray. For this purpose, the Mo quantity added to the WCCo was changed as 10, 20, 30 and 40 wt. % percents. The coatings are compared in terms of their phase composition, microstructure and hardness. Phase compound and microstructure of coating layers were examined using X-ray diffractometer (XRD and scanning electron microscope (SEM. XRD results showed that WCCo-Mo composite coatings were mainly composed of WC, W2C, Co3W3C, Mo2C, MoO2, Mo and Co phases. The average hardness of the coatings increased with increasing Mo content.

  12. Growth kinetic and fuel quality parameters as selective criterion for screening biodiesel producing cyanobacterial strains.

    Science.gov (United States)

    Gayathri, Manickam; Shunmugam, Sumathy; Mugasundari, Arumugam Vanmathi; Rahman, Pattanathu K S M; Muralitharan, Gangatharan

    2018-01-01

    The efficiency of cyanobacterial strains as biodiesel feedstock varies with the dwelling habitat. Fourteen indigenous heterocystous cyanobacterial strains from rice field ecosystem were screened based on growth kinetic and fuel parameters. The highest biomass productivity was obtained in Nostoc punctiforme MBDU 621 (19.22mg/L/day) followed by Calothrix sp. MBDU 701 (13.43mg/L/day). While lipid productivity and lipid content was highest in Nostoc spongiaeforme MBDU 704 (4.45mg/L/day and 22.5%dwt) followed by Calothrix sp. MBDU 701 (1.54mg/L/day and 10.75%dwt). Among the tested strains, Nostoc spongiaeforme MBDU 704 and Nostoc punctiforme MBDU 621 were selected as promising strains for good quality biodiesel production by Preference Ranking Organization Method for Enrichment Evaluation (PROMETHEE) and Graphical Analysis for Interactive Assistance (GAIA) analysis. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. Process for the fabrication of a nuclear fuel

    International Nuclear Information System (INIS)

    Hirose, Yasuo.

    1970-01-01

    Herein disclosed is a process for fabricating a nuclear fuel incorporating either uranium or plutonium. A pellet-like substrate consisting of a packed powder ceramic fuel such as uranium or plutonium is prepared with the horizontal surface of the body provided with a masking. Next, after impregnating the substrate voids with a solution consisting of a fissile material or mixture of fissile material and poison, the solvent is removed by a chemical deposition process which causes the impregnated material to migrate through capillary action toward the vicinity of the fuel body surface. Sintering and pyrolysis of the deposited material and masking are subsequently carried out to yield a fuel body having adjacent to its surface an intensely concentrated layer of either fissile material or a mixture of fissile material and poison. (Owens, K.J.)

  14. Used Fuel Cask Identification through Neutron Profile

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-20

    Currently, most spent fuel is stored near reactors. An interim consolidated fuel storage facility would receive fuel from multiple sites and store it in casks on site for decades. For successful operation of such a facility there is need for a way to restore continuity of knowledge if lost as well as a method that will indicate state of fuel inside the cask. Used nuclear fuel is identifiable by its radiation emission, both gamma and neutron. Neutron emission from fission products, multiplication from remaining fissile material, and the unique distribution of both in each cask produce a unique neutron signature. If two signatures taken at different times do not match, either changes within the fuel content or misidentification of a cask occurred. It was found that identification of cask loadings works well through the profile of emitted neutrons in simulated real casks. Even casks with similar overall neutron emission or average counts around the circumference can be distinguished from each other by analyzing the profile. In conclusion, (1) identification of unaltered casks through neutron signature profile is viable; (2) collecting the profile provides insight to the condition and intactness of the fuel stored inside the cask; and (3) the signature profile is stable over time.

  15. Using the second law of thermodynamics for enrichment and isolation of microorganisms to produce fuel alcohols or hydrocarbons.

    Science.gov (United States)

    Kohn, Richard A; Kim, Seon-Woo

    2015-10-07

    Fermentation of crops, waste biomass, or gases has been proposed as a means to produce desired chemicals and renewable fuels. The second law of thermodynamics has been shown to determine the net direction of metabolite flow in fermentation processes. In this article, we describe a process to isolate and direct the evolution of microorganisms that convert cellulosic biomass or gaseous CO2 and H2 to biofuels such as ethanol, 1-butanol, butane, or hexane (among others). Mathematical models of fermentation elucidated sets of conditions that thermodynamically favor synthesis of desired products. When these conditions were applied to mixed cultures from the rumen of a cow, bacteria that produced alcohols or alkanes were isolated. The examples demonstrate the first use of thermodynamic analysis to isolate bacteria and control fermentation processes for biofuel production among other uses. Copyright © 2015 Elsevier Ltd. All rights reserved.

  16. How much ethanol fuel can be produced from sugarcane in Hawaii

    OpenAIRE

    Kwong, John

    2014-01-01

    This study evaluates how much sugar ethanol Hawaii can produce. Fossilfuel reserves will diminish with time, and alternative energy may not be effectivein totally replacing combustible engines for all application. Factors important tosugar ethanol production and distribution are examined and evaluated.  

  17. Ensuring the 50 year life of a fissile material container

    International Nuclear Information System (INIS)

    Glass, R.E.; Towne, T.L.

    1997-12-01

    Sandia was presented with an opportunity in 1993 to design containers for the long term storage and transport of fissile material. This program was undertaken at the direction of the US Department of Energy and in cooperation with Lawrence Livermore National Laboratory and Los Alamos National Laboratory which were tasked with developing the internal fixturing for the contents. The hardware is being supplied by Allied Signal Federal Manufacturing and Technologies, and the packaging will occur at Mason and Hangar Corporation's Pantex Plant. The unique challenge was to design a container that could be sealed with the fissile material contents; and, anytime during the next 50 years, the container could be transported with only the need for the pre-shipment leak test. This required not only a rigorous design capable of meeting the long term storage and transportation requirements, but also resulted in development of a surveillance program to ensure that the container continues to perform as designed over the 50-year life. This paper addresses the design of the container, the testing that was undertaken to demonstrate compliance with US radioactive materials transport regulations, and the surveillance program that has been initiated to ensure the 50-year performance

  18. Automated monitoring of fissile and fertile materials in incinerator residue

    International Nuclear Information System (INIS)

    Schoenig, F.C. Jr.; Glendinning, S.G.; Tunnell, G.W.; Zucker, M.S.

    1986-01-01

    This patent describes an apparatus for determining the fissile and fertile material content of incinerator residue contained in a manipulatable container. The apparatus comprises a main body member formed of neutron moderating material and formed with a well for receiving the container; a first plug formed of neutron reflecting material for closing the top of the well; and a second plug containing a first neutron source for alternatively closing the top of the well and for directing neutrons into the well. It also includes a second neutron source selectively positionable in the bottom of the well for directing neutrons into the well; manipulating means for placing the container in the well and removing the container therefrom and for selectively placing one of the first and second plugs in the top of the well. Neutron detectors are positioned within the neutron moderating material of the main body member around the sides of the well. At least one gamma ray detector is positioned adjacent the bottom of the well. A means receives and processes the signals from the neutron and gamma ray detectors when the container is in the well for determining the fissile and fertile material content of the incinerator residue in the container

  19. Calculation of the minimum critical mass of fissile nuclides

    International Nuclear Information System (INIS)

    Wright, R.Q.; Hopper, Calvin Mitchell

    2008-01-01

    The OB-1 method for the calculation of the minimum critical mass of fissile actinides in metal/water systems was described in a previous paper. A fit to the calculated minimum critical mass data using the extended criticality parameter is the basis of the revised method. The solution density (grams/liter) for the minimum critical mass is also obtained by a fit to calculated values. Input to the calculation consists of the Maxwellian averaged fission and absorption cross sections and the thermal values of nubar. The revised method gives more accurate values than the original method does for both the minimum critical mass and the solution densities. The OB-1 method has been extended to calculate the uncertainties in the minimum critical mass for 12 different fissile nuclides. The uncertainties for the fission and capture cross sections and the estimated nubar uncertainties are used to determine the uncertainties in the minimum critical mass, either in percent or grams. Results have been obtained for U-233, U-235, Pu-236, Pu-239, Pu-241, Am-242m, Cm-243, Cm-245, Cf-249, Cf-251, Cf-253, and Es-254. Eight of these 12 nuclides are included in the ANS-8.15 standard.

  20. Homogeneous Thorium Fuel Cycles in Candu Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, B.; Dyck, G.R.; Edwards, G.W.R.; Magill, M. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada)

    2009-06-15

    The CANDU{sup R} reactor has an unsurpassed degree of fuel-cycle flexibility, as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle [1]. These features facilitate the introduction and full exploitation of thorium fuel cycles in Candu reactors in an evolutionary fashion. Because thorium itself does not contain a fissile isotope, neutrons must be provided by adding a fissile material, either within or outside of the thorium-based fuel. Those same Candu features that provide fuel-cycle flexibility also make possible many thorium fuel-cycle options. Various thorium fuel cycles can be categorized by the type and geometry of the added fissile material. The simplest of these fuel cycles are based on homogeneous thorium fuel designs, where the fissile material is mixed uniformly with the fertile thorium. These fuel cycles can be competitive in resource utilization with the best uranium-based fuel cycles, while building up a 'mine' of U-233 in the spent fuel, for possible recycle in thermal reactors. When U-233 is recycled from the spent fuel, thorium-based fuel cycles in Candu reactors can provide substantial improvements in the efficiency of energy production from existing fissile resources. The fissile component driving the initial fuel could be enriched uranium, plutonium, or uranium-233. Many different thorium fuel cycle options have been studied at AECL [2,3]. This paper presents the results of recent homogeneous thorium fuel cycle calculations using plutonium and enriched uranium as driver fuels, with and without U-233 recycle. High and low burnup cases have been investigated for both the once-through and U-233 recycle cases. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). 1. Boczar, P.G. 'Candu Fuel-Cycle Vision', Presented at IAEA Technical Committee Meeting on 'Fuel Cycle Options for LWRs and HWRs', 1998 April 28 - May 01, also Atomic Energy

  1. 78 FR 44075 - Notice of Data Availability Concerning Renewable Fuels Produced From Barley Under the RFS Program

    Science.gov (United States)

    2013-07-23

    ... and diesel fuel or renewable fuels such as biodiesel and renewable diesel. Regulated categories... production plants. Fuel and feedstock transport includes emissions from transporting bushels of harvested..., Mean (Low/High) 11,290 (2,784/21,679) Fuel Production 39,069 19,200 Fuel and Feedstock Transport 4,861...

  2. Solid oxide fuel cell electrolytes produced via very low pressure suspension plasma spray and electrophoretic deposition

    Science.gov (United States)

    Fleetwood, James D.

    Solid oxide fuel cells (SOFCs) are a promising element of comprehensive energy policies due to their direct mechanism for converting the oxidization of fuel, such as hydrogen, into electrical energy. Both very low pressure plasma spray and electrophoretic deposition allow working with high melting temperature SOFC suspension based feedstock on complex surfaces, such as in non-planar SOFC designs. Dense, thin electrolytes of ideal composition for SOFCs can be fabricated with each of these processes, while compositional control is achieved with dissolved dopant compounds that are incorporated into the coating during deposition. In the work reported, sub-micron 8 mole % Y2O3-ZrO2 (YSZ) and gadolinia-doped ceria (GDC), powders, including those in suspension with scandium-nitrate dopants, were deposited on NiO-YSZ anodes, via very low pressure suspension plasma spray (VLPSPS) at Sandia National Laboratories' Thermal Spray Research Laboratory and electrophoretic deposition (EPD) at Purdue University. Plasma spray was carried out in a chamber held at 320 - 1300 Pa, with the plasma composed of argon, hydrogen, and helium. EPD was characterized utilizing constant current deposition at 10 mm electrode separation, with deposits sintered from 1300 -- 1500 °C for 2 hours. The role of suspension constituents in EPD was analyzed based on a parametric study of powder loading, powder specific surface area, polyvinyl butyral (PVB) content, polyethyleneimine (PEI) content, and acetic acid content. Increasing PVB content and reduction of particle specific surface area were found to eliminate the formation of cracks when drying. PEI and acetic acid content were used to control suspension stability and the adhesion of deposits. Additionally, EPD was used to fabricate YSZ/GDC bilayer electrolyte systems. The resultant YSZ electrolytes were 2-27 microns thick and up to 97% dense. Electrolyte performance as part of a SOFC system with screen printed LSCF cathodes was evaluated with peak

  3. Large-scale fuel cycle centres

    International Nuclear Information System (INIS)

    Smiley, S.H.; Black, K.M.

    1977-01-01

    The US Nuclear Regulatory Commission (NRC) has considered the nuclear energy centre concept for fuel cycle plants in the Nuclear Energy Centre Site Survey 1975 (NECSS-75) Rep. No. NUREG-0001, an important study mandated by the US Congress in the Energy Reorganization Act of 1974 which created the NRC. For this study, the NRC defined fuel cycle centres as consisting of fuel reprocessing and mixed-oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle centre sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000-300,000MW(e). The types of fuel cycle facilities located at the fuel cycle centre permit the assessment of the role of fuel cycle centres in enhancing the safeguard of strategic special nuclear materials - plutonium and mixed oxides. Siting fuel cycle centres presents a smaller problem than siting reactors. A single reprocessing plant of the scale projected for use in the USA (1500-2000t/a) can reprocess fuel from reactors producing 50,000-65,000MW(e). Only two or three fuel cycle centres of the upper limit size considered in the NECSS-75 would be required in the USA by the year 2000. The NECSS-75 fuel cycle centre evaluation showed that large-scale fuel cycle centres present no real technical siting difficulties from a radiological effluent and safety standpoint. Some construction economies may be achievable with fuel cycle centres, which offer opportunities to improve waste-management systems. Combined centres consisting of reactors and fuel reprocessing and mixed-oxide fuel fabrication plants were also studied in the NECSS. Such centres can eliminate shipment not only of Pu but also mixed-oxide fuel. Increased fuel cycle costs result from implementation of combined centres unless the fuel reprocessing plants are commercial-sized. Development of Pu-burning reactors could reduce any economic penalties of combined centres. The need for effective fissile

  4. Response of range grasses to water produced from in situ fossil fuel processing

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, Q D; Moore, T S; Sexton, J C

    1984-11-01

    In situ-produced waters collected while retorting oil shale and tar sands to produce oil, and coal to produce gas, were tested for their effects on plant growth. Basin wildrye (Elymus cinereus), western wheatgrass (Agropyron smithii) 'Rosana', alkali sacaton (Sporobolus airoides), bluebunch wheatgrass (Agropyron spicatum) and Nuttall alkaligrass (Puccinellia airoides) were utilized. Root weight, shoot weight, total dry weight, leaf area and root/shoot weight ratios were determined. All experiments were conducted under greenhouse conditions using hydroponic techniques and horticultural grade perlite for plant support. Measurements were collected after a 10-week growth period. Results show that differences in plant growth can be monitored using dry biomass, leaf area and root to shoot ratio measurements when plants are subjected to retort waters. Plant species reaction to a water may be different. Generally, alkali sacaton, basin wildrye and western wheatgrass are least susceptible to toxicity by the majority of retort waters tested. Bluebunch wheatgrass is most susceptible. Waters from different retort procedures vary in toxicity to different plant species.

  5. Biotechnology for producing fuels and chemicals from biomass. Volume II. Fermentation chemicals from biomass

    Energy Technology Data Exchange (ETDEWEB)

    Villet, R. (ed.)

    1981-02-01

    The technological and economic feasibility of producing some selected chemicals by fermentation is discussed: acetone, butanol, acetic acid, citric acid, 2,3-butanediol, and propionic acid. The demand for acetone and butanol has grown considerably. They have not been produced fermentatively for three decades, but instead by the oxo and aldol processes. Improved cost of fermentative production will hinge on improving yields and using cellulosic feedstocks. The market for acetic acid is likely to grow 5% to 7%/yr. A potential process for production is the fermentation of hydrolyzed cellulosic material to ethanol followed by chemical conversion to acetic acid. For about 50 years fermentation has been the chief process for citric acid production. The feedstock cost is 15% to 20% of the overall cost of production. The anticipated 5%/yr growth in demand for citric acid could be enhanced by using it to displace phosphates in detergent manufacture. A number of useful chemicals can be derived from 2,3-butanediol, which has not been produced commercially on a large scale. R and D are needed to establish a viable commercial process. The commercial fermentative production of propionic acid has not yet been developed. Recovery and purification of the product require considerable improvement. Other chemicals such as lactic acid, isopropanol, maleic anhydride, fumarate, and glycerol merit evaluation for commercial fermentative production in the near future.

  6. Microbial development in distillers wet grains produced during fuel ethanol production from corn (Zea mays)

    Energy Technology Data Exchange (ETDEWEB)

    Lehman, R.M.; Rosentrater, K.A. [United States Dept. of Agriculture, Brookings, SD (United States). North Central Agricultural Research Laboratory

    2007-09-15

    The microbiology of post-production distillers wet grains (DWG) was investigated over a period of 9 days at an industrial ethanol plant. Samples of the DWG were physically and chemically characterized. Compositional analyses were conducted for protein, fiber, and fat. Fixed suspensions of DWG were dispersed and disrupted by sonication. Bacterial cells were enumerated under epifluorescent illumination. Solid media and standard dilution were used to enumerate total colony-forming units (CFU) of lactic-acid producing bacteria (LAB), and aerobic heterotrophic organisms. The DWG had a pH of approximately 4.4, a moisture content of 53.5 per cent, and 4 x 10{sup 5} total yeast cells. Thirteen morphologically distinct isolates were identified during the study, 10 of which were yeasts and molds from 6 different genera. Two of the yeasts were of the lactic-acid Pediococcus pentosaceus strain, and 1 of the yeasts was an aerobic heterotrophic bacteria. Results showed that the matrix of the DWG produced severe technical difficulties for several of the culture-independent community-level analyses. It was concluded that numbers of potentially beneficial bacteria appeared to increase over the time period relative to potential spoilage agents. Molds capable of producing mycotoxins colonized the DWG and grew to high densities over the 9 day period. 31 refs., 3 tabs., 2 figs.

  7. Chemical inhibition of the contaminant Lactobacillus fermentum from distilleries producing fuel bioethanol

    Directory of Open Access Journals (Sweden)

    Pedro de Oliva Neto

    2014-06-01

    Full Text Available The purpose of this study was to determine the Minimum Inhibitory Concentration (MIC of pure or mixed chemicals for Saccharomyces cerevisiae and Lactobacillus fermentum in the samples isolated from distilleries with serious bacterial contamination problems. The biocides, which showed the best results were: 3,4,4' trichlorocarbanilide (TCC, tested at pH 4.0 (MIC = 3.12 mg/l, TCC with benzethonium chloride (CBe at pH 6.0 (MIC = 3.12 mg/l and TCC mixed with benzalkonium chloride (CBa at pH 6.0 (MIC = 1.53 mg /l. If CBa was used in sugar cane milling in 1:1 ratio with TCC, a 8 times reduction of CBa was possible. This formulation also should be tested in fermentation steps since it was more difficult for the bacterium to develop resistance to biocide. There was no inhibition of S. cerevisiae and there were only antibiotics as an option to bacterial control of fuel ethanol fermentation by S. cerevisiae.

  8. Combustion Characteristics of Chlorine-Free Solid Fuel Produced from Municipal Solid Waste by Hydrothermal Processing

    Directory of Open Access Journals (Sweden)

    Kunio Yoshikawa

    2012-11-01

    Full Text Available An experimental study on converting municipal solid waste (MSW into chlorine-free solid fuel using a combination of hydrothermal processing and water-washing has been performed. After the product was extracted from the reactor, water-washing experiments were then conducted to obtain chlorine-free products with less than 3000 ppm total chlorine content. A series of combustion experiments were then performed for the products before and after the washing process to determine the chlorine content in the exhaust gas and those left in the ash after the combustion process at a certain temperature. A series of thermogravimetric analyses were also conducted to compare the combustion characteristics of the products before and after the washing process. Due to the loss of ash and some volatile matter after washing process, there were increases in the fixed carbon content and the heating value of the product. Considering the possible chlorine emission, the washing process after the hydrothermal treatment should be necessary only if the furnace temperature is more than 800 °C.

  9. Young and Especially Senescent Endothelial Microvesicles Produce NADPH: The Fuel for Their Antioxidant Machinery

    Directory of Open Access Journals (Sweden)

    Guillermo Bodega

    2018-01-01

    Full Text Available In a previous study, we demonstrated that endothelial microvesicles (eMVs have a well-developed enzymatic team involved in reactive oxygen species detoxification. In the present paper, we demonstrate that eMVs can synthesize the reducing power (NAD(PH that nourishes this enzymatic team, especially those eMVs derived from senescent human umbilical vein endothelial cells. Moreover, we have demonstrated that the molecules that nourish the enzymatic machinery involved in NAD(PH synthesis are blood plasma metabolites: lactate, pyruvate, glucose, glycerol, and branched-chain amino acids. Drastic biochemical changes are observed in senescent eMVs to optimize the synthesis of reducing power. Mitochondrial activity is diminished and the glycolytic pathway is modified to increase the activity of the pentose phosphate pathway. Different dehydrogenases involved in NADPH synthesis are also increased. Functional experiments have demonstrated that eMVs can synthesize NADPH. In addition, the existence of NADPH in eMVs was confirmed by mass spectrometry. Multiphoton confocal microscopy images corroborate the synthesis of reducing power in eMVs. In conclusion, our present and previous results demonstrate that eMVs can act as autonomous reactive oxygen species scavengers: they use blood metabolites to synthesize the NADPH that fuels their antioxidant machinery. Moreover, senescent eMVs have a stronger reactive oxygen species scavenging capacity than young eMVs.

  10. Agricultural residues as fuel for producer gas generation. Report from a test series with coconut shells, coconut husks, wheat straw and sugar cane

    Energy Technology Data Exchange (ETDEWEB)

    Hoeglund, C

    1981-08-01

    This paper reports on results from a series of tests with four different types of agricultural residues as fuel for producer gas generation. The fuels are coconut shells, coconut husks, pelletized wheat straw and pressed sugar cane. The tests were made with a 73 Hp agricultural tractor diesel engine equipped with a standard gasifier developed for wood chips in Sweden, and run on a testbed at the Swedish National Machinery Testing Institute. The engine was operated on approximately 10 per cent diesel oil and 90 per cent producer gas. The gas composition, its calorific value and temperature, the pressure drop and the engine power were monitored. Detailed elementary analysis of the fuel and gas were carried out. Observations were also made regarding the important aspects of bridging and slagging in the gasifier. The tests confirmed that coconut shells make an excellent fuel for producer gas generation. After 8 hours of running no problems with slags and bridging were experienced. Coconut husks showed no bridging but some slag formation. The gasifier operated satisfactorily for this fuel. Pelletized wheat straw and pressed sugar cane appeared unsuitable as fuel in the unmodified test gasifier (Type F 300) due to slag formation. It is important to note, however, that the present results are not optimal for any of the fuel used, the gasifier being designed for wood-chips and not for the test-fuels used. Tests using appropriately modified gasifiers are planned for the future.

  11. MOX fuel irradiation behaviour: Results from X-ray microbeam analysis

    International Nuclear Information System (INIS)

    Walker, C.T.; Goll, W.; Matsumura, T.

    1997-01-01

    The behaviour of plutonium, xenon and caesium were investigated in two sections of irradiated MOX fuel produced by the OCOM process. In one fuel (OCOM30), the MOX agglomerates contained 18 wt% fissile plutonium, and had a low volume fraction of 0.17; in the other (OCOM15) the agglomerates contained 9 wt% fissile plutonium, and had a high volume fraction of 0.34. Both fuels had been irradiated under normal power reactor conditions to a burn-up of approximately 44 GWd/t. The main aim of the work was to establish whether the above differences in composition affected the percentage fission gas released by the fuels. Since U/Pu interdiffusion did not occurred during the irradiation, both fuels remained inhomogeneous on the microscopic scale. However, the concentration of plutonium in the MOX agglomerates decreases by about 50% as a result of fission, whereas the plutonium content of the UO 2 matrix increased by about a factor of four to approximately 2 wt% due to neutron capture by 238 U. The agglomerates in the OCOM15 fuel generally exhibited a finer structure due to the lower burn-up. More than 80% of the fission gas had been released from the oxide lattice of the MOX agglomerates in both fuels. However, a very high fraction of this gas precipitated and remained in the pore structure of the agglomerates. Consequently, puncturing revealed that for both fuels the percentage of gas released to the rod free volume increased from less than 0.5% at 10 GWd/t to a maximum of 3.5% at 45 GWd/t. The conclusion is that the percentage of gas released by MOX fuel is largely unaffected of the level of inhomogeneity of the fuel. In both fuels caesium showed near complete retention in both the MOX agglomerates and the UO 2 matrix. (author). 8 refs, 11 figs, 3 tabs

  12. Mathematical model for choosing the nuclear safe matrix compositions for fissile material immobilization

    International Nuclear Information System (INIS)

    Gorshtein, A.I.; Matyunin, Yu.I.; Poluehktov, P.P.

    2000-01-01

    A mathematical model is proposed for preliminary choice of the nuclear safe matrix compositions for fissile material immobilization. The IBM PC computer software for nuclear safe matrix composition calculations is developed. The limiting concentration of fissile materials in the some used and perspective nuclear safe matrix compositions for radioactive waste immobilization is calculated [ru

  13. Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

    Science.gov (United States)

    Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

    2013-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

  14. Experience of work with radioactive materials and nuclear fuel at the reactor WWR-K

    International Nuclear Information System (INIS)

    Maltseva, R.M.; Petukhov, V.K.

    1998-01-01

    In the report there are considered questions concerning the handling with fresh and spent fuel, experimental devices, containing high enriched uranium, being fissile materials of the bulk form, radioisotopes, obtained in the reactor, and radioactive waste, formed during the operation of the reactor, and organization of storage, account and control of radioactive and fissile materials is described. (author)

  15. Single-column ion chromatography with determination of hydrazoic acid produced in spent nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Ma Guilan; Tan Shuping

    2006-01-01

    The reaction of hydrazine and its derivative with ammonium metavanadate may produce hydrazoic acid (HN 3 ). A single-column ion chromatography is used for the determination of HN 3 after neutralizing the rest acid in the sample with sodium hydroxide. Chromatography separation of HN 3 is carried out on a 25 cm x 0.46 cm (inside diameter) stainless steel column packed with Vydac IC302 ion Chromatography packing. The eluent is 1 mmol/L o-phthalic acid, and the ion is detected by conductivity detector. The detection limit in the presence chromatography is 5 μg/mL, the linear range is from 5 to 201 μg/mL, the linear correlation coefficient is 0.9994, respectively. The analysis accuracy is 2% for standard sample, and the detection limit is 51 μg/mL for HN 3 in the real sample. (authors)

  16. Supply Chain Sustainability Analysis of Fast Pyrolysis and Hydrotreating Bio-Oil to Produce Hydrocarbon Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Adom, Felix K. [Argonne National Lab. (ANL), Argonne, IL (United States); Cai, Hao [Argonne National Lab. (ANL), Argonne, IL (United States); Dunn, Jennifer B. [Argonne National Lab. (ANL), Argonne, IL (United States); Hartley, Damon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Searcy, Erin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tan, Eric [National Renewable Energy Lab. (NREL), Golden, CO (United States); Jones, Sue [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Snowden-Swan, Lesley [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-03-01

    The Department of Energy’s (DOE) Bioenergy Technology Office (BETO) aims at developing and deploying technologies to transform renewable biomass resources into commercially viable, high-performance biofuels, bioproducts and biopower through public and private partnerships (DOE, 2015). BETO and its national laboratory teams conduct in-depth techno-economic assessments (TEA) of technologies to produce biofuels. These assessments evaluate feedstock production, logistics of transporting the feedstock, and conversion of the feedstock to biofuel. There are two general types of TEAs. A design case is a TEA that outlines a target case for a particular biofuel pathway. It enables identification of data gaps and research and development needs, and provides goals and targets against which technology progress is assessed. On the other hand, a state of technology (SOT) analysis assesses progress within and across relevant technology areas based on actual experimental results relative to technical targets and cost goals from design cases, and includes technical, economic, and environmental criteria as available.

  17. A method of producing a multilayer barrier structure for a solid oxide fuel cell

    DEFF Research Database (Denmark)

    2010-01-01

    The present invention provides a method of producing a multilayer barrier structure for a solid oxide cell stack, comprising the steps of: - providing a metal interconnect, wherein the metal interconnect is a ferritic stainless steel layer; - applying a first metal oxide layer on said metal...... oxide; and - reacting the metal oxide in said first metal oxide layer with the metal of said metal interconnect during the SOC-stack initialisation, and a solid oxide stack comprising an anode contact layer and support structure, an anode layer, an electrolyte layer, a cathode layer, a cathode contact...... layer, a metallic interconnect, and a multilayer barrier structure which is obtainable by the above method and through an initialisation step, which is carried out under controlled conditions for atmosphere composition and current load, which depends on the layer composition facilitating the formation...

  18. A treaty on the cutoff of fissile material for nuclear weapons - What to cover? How to verify?

    International Nuclear Information System (INIS)

    Schaper, A.

    1998-01-01

    Since 1946, a cutoff has been proposed. In 1993, the topic was placed on the agenda of the CD. The establishment of an Ad Hoc Committee in the CD with a mandate to negotiate a fissile material cutoff treaty struggled with difficulties for more than a year. The central dispute was whether the mandate should refer to existing un-safeguarded stockpiles. The underlying conflict of the CTBT negotiations can be summarized as nuclear disarmament versus nuclear nonproliferation The same conflict is now blocking progress with FMCT negotiations in the CD. At the center of technical proliferation concerns is direct use material that can be used for nuclear warheads without any further enrichment or reprocessing. Those materials are plutonium and highly enriched uranium (HEU). A broader category of materials is defined as all those containing any fissile isotopes, called special fissionable materials. In order ta verify that no direct use materials are abused for military purposes, also special fissionable materials must be controlled. An even broader category is simply called nuclear materials. Pu and HEU can be distinguished into the following categories of utilisation: 1. military direct use material in operational nuclear weapons and their logistics pipeline, 2. military direct use material held in reserve for military purposes, in assembled weapons or in other forms, 3. military direct use material withdrawn from dismantled weapons, 4. military direct use material considered excess and designated for transfer into civilian use, 5. military direct use material considered excess and declared for transfer into civilian use, 6. direct use material currently in reactors or their logistics pipelines and storages, and 7. irradiated Pu and HEU in spent fuel from reactors, or in vitrified form for final disposal. Large quantities of materials are neither inside weapons nor declared excess. So far, there are no legal obligations for NWS for limitations, declarations, or

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  20. Warheads and Fissile Materials:Declarations and Counting

    International Nuclear Information System (INIS)

    Sutcliffe, W.G.

    1991-01-01

    This paper reviews some of the issues about verifying the dismantlement of nuclear warheads and controlling nuclear materials in the context of arms control objectives. It is asserted that information about the stockpiles of nuclear warheads and materials is necessary to analyze the impacts and verification requirements of arms control measures including warhead dismantlement and fissile material controls. It is proposed that the US and the Soviets engage in a series of declarations about their stockpiles of nuclear weapons and materials. It is also asserted that currently it is more important to verify that warheads are retired to safe, secure facilities than to verify their dismantlement. It is proposed that production of new or rebuilt warheads be limited to less than the number retired each year. Verifying the number of new and rebuilt warheads deployed and the number retired avoids many of the difficulties in verifying dismantlement and material controls

  1. Anodic biofilms in microbial fuel cells harbor low numbers of higher-power-producing bacteria than abundant genera

    Energy Technology Data Exchange (ETDEWEB)

    Kiely, Patrick D.; Call, Douglas F.; Yates, Matthew D.; Regan, John M.; Logan, Bruce E. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Civil and Environmental Engineering

    2010-09-15

    Microbial fuel cell (MFC) anode communities often reveal just a few genera, but it is not known to what extent less abundant bacteria could be important for improving performance. We examined the microbial community in an MFC fed with formic acid for more than 1 year and determined using 16S rRNA gene cloning and fluorescent in situ hybridization that members of the Paracoccus genus comprised most ({proportional_to}30%) of the anode community. A Paracoccus isolate obtained from this biofilm (Paracoccus denitrificans strain PS-1) produced only 5.6 mW/m{sup 2}, whereas the original mixed culture produced up to 10 mW/m{sup 2}. Despite the absence of any Shewanella species in the clone library, we isolated a strain of Shewanella putrefaciens (strain PS-2) from the same biofilm capable of producing a higher-power density (17.4 mW/m{sup 2}) than the mixed culture, although voltage generation was variable. Our results suggest that the numerical abundance of microorganisms in biofilms cannot be assumed a priori to correlate to capacities of these predominant species for high-power production. Detailed screening of bacterial biofilms may therefore be needed to identify important strains capable of high-power generation for specific substrates. (orig.)

  2. Anodic biofilms in microbial fuel cells harbor low numbers of higher-power-producing bacteria than abundant genera

    KAUST Repository

    Kiely, Patrick D.; Call, Douglas F.; Yates, Matthew D.; Regan, John M.; Logan, Bruce E.

    2010-01-01

    Microbial fuel cell (MFC) anode communities often reveal just a few genera, but it is not known to what extent less abundant bacteria could be important for improving performance. We examined the microbial community in an MFC fed with formic acid for more than 1 year and determined using 16S rRNA gene cloning and fluorescent in situ hybridization that members of the Paracoccus genus comprised most (~30%) of the anode community. A Paracoccus isolate obtained from this biofilm (Paracoccus denitrificans strain PS-1) produced only 5.6 mW/m 2, whereas the original mixed culture produced up to 10 mW/m 2. Despite the absence of any Shewanella species in the clone library, we isolated a strain of Shewanella putrefaciens (strain PS-2) from the same biofilm capable of producing a higher-power density (17.4 mW/m2) than the mixed culture, although voltage generation was variable. Our results suggest that the numerical abundance of microorganisms in biofilms cannot be assumed a priori to correlate to capacities of these predominant species for high-power production. Detailed screening of bacterial biofilms may therefore be needed to identify important strains capable of high-power generation for specific substrates. © 2010 Springer-Verlag.

  3. Anodic biofilms in microbial fuel cells harbor low numbers of higher-power-producing bacteria than abundant genera

    KAUST Repository

    Kiely, Patrick D.

    2010-07-15

    Microbial fuel cell (MFC) anode communities often reveal just a few genera, but it is not known to what extent less abundant bacteria could be important for improving performance. We examined the microbial community in an MFC fed with formic acid for more than 1 year and determined using 16S rRNA gene cloning and fluorescent in situ hybridization that members of the Paracoccus genus comprised most (~30%) of the anode community. A Paracoccus isolate obtained from this biofilm (Paracoccus denitrificans strain PS-1) produced only 5.6 mW/m 2, whereas the original mixed culture produced up to 10 mW/m 2. Despite the absence of any Shewanella species in the clone library, we isolated a strain of Shewanella putrefaciens (strain PS-2) from the same biofilm capable of producing a higher-power density (17.4 mW/m2) than the mixed culture, although voltage generation was variable. Our results suggest that the numerical abundance of microorganisms in biofilms cannot be assumed a priori to correlate to capacities of these predominant species for high-power production. Detailed screening of bacterial biofilms may therefore be needed to identify important strains capable of high-power generation for specific substrates. © 2010 Springer-Verlag.

  4. Charge distribution on plutonium-containing aerosols produced in mixed-oxide reactor fuel fabrication and the laboratory

    International Nuclear Information System (INIS)

    Yeh, H.C.; Newton, G.J.; Teague, S.V.

    1976-01-01

    The inhalation toxicity of potentially toxic aerosols may be affected by the electrostatic charge on the particles. Charge may influence the deposition site during inhalation and therefore its subsequent clearance and dose patterns. The electrostatic charge distributions on plutonium-containing aerosols were measured with a miniature, parallel plate, aerosol electrical mobility spectrometer. Two aerosols were studied: a laboratory-produced 238 PuO 2 aerosol (15.8 Ci/g) and a plutonium mixed-oxide aerosol (PU-MOX, natural UO 2 plus PuO 2 , 0.02 Ci/g) formed during industrial centerless grinding of mixed-oxide reactor fuel pellets. Plutonium-238 dioxide particles produced in the laboratory exhibited a small net positive charge within a few minutes after passing through a 85 Kr discharger due to alpha particle emission removal of valence electrons. PU-MOX aerosols produced during centerless grinding showed a charge distribution essentially in Boltzmann equilibrium. The gross alpha aerosol concentrations (960-1200 nCi/l) within the glove box were sufficient to provide high ion concentrations capable of discharging the charge induced by mechanical and/or nuclear decay processes

  5. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Homan, F.J.; Balthesen, E.; Turner, R.F.

    1977-01-01

    Significant advances have occurred in the development of HTGR fuel and fuel cycle. These accomplishments permit a wide choice of fuel designs, reactor concepts, and fuel cycles. Fuels capable of providing helium outlet temperatures of 750 0 C are available, and fuels capable of 1000 0 C outlet temperatures may be expected from extension of present technology. Fuels have been developed for two basic HTGR designs, one using a spherical (pebble bed) element and the other a prismatic element. Within each concept a number of variations of geometry, fuel composition, and structural materials are permitted. Potential fuel cycles include both low-enriched and high-enriched Th- 235 U, recycle Th- 233 U, and Th-Pu or U-Pu cycles. This flexibility offered by the HTGR is of great practical benefit considering the rapidly changing economics of power production. The inflation of ore prices has increased optimum conversion ratios, and increased the necessity of fuel recycle at an early date. Fuel element makeup is very similar for prismatic and spherical designs. Both use spherical fissile and fertile particles coated with combinations of pyrolytic carbon and silicon carbide. Both use carbonaceous binder materials, and graphite as the structural material. Weak-acid resin (WAR) UO 2 -UC 2 fissile fuels and sol-gel-derived ThO 2 fertile fuels have been selected for the Th- 233 U cycle in the prismatic design. Sol-gel-derived UO 2 UC 2 is the reference fissile fuel for the low-enriched pebble bed design. Both the United States and Federal Republic of Germany are developing technology for fuel cycle operations including fabrication, reprocessing, refabrication, and waste handling. Feasibility of basic processes has been established and designs developed for full-scale equipment. Fuel and fuel cycle technology provide the basis for a broad range of applications of the HTGR. Extension of the fuels to higher operating temperatures and development and commercial demonstration of fuel

  6. Literature review of thermal and radiation performance parameters for high-temperature, uranium dioxide fueled cermet materials

    International Nuclear Information System (INIS)

    Haertling, C.; Hanrahan, R.J.

    2007-01-01

    High-temperature fissile-fueled cermet literature was reviewed. Data are presented primarily for the W-UO 2 as this was the system most frequently studied; other reviewed systems include cermets with Mo, Re, or alloys as a matrix. Failure mechanisms for the cermets are typically degradation of mechanical integrity and loss of fuel. Mechanical failure can occur through stresses produced from dissimilar expansion coefficients, voids created from diffusion of dissimilar materials or formation of metal hydride and subsequent volume expansion. Fuel loss failure can occur by high temperature surface vaporization or by vaporization after loss of mechanical integrity. Techniques found to aid in retaining fuel include the use of coatings around UO 2 fuel particles, use of oxide stabilizers in the UO 2 , minimizing grain sizes in the metal matrix, minimizing impurities, controlling the cermet sintering atmosphere, and cladding around the cermet

  7. Pressing device for producing compacts from source material in powder form in particular pulverized nuclear reactor fuel

    International Nuclear Information System (INIS)

    Heller, G.; Adelmann, M.; Konigs, W.; Wendorf, W.

    1984-01-01

    Pressing device for producing compacts from source material in powder form, in particular pulverized nuclear reactor fuel having a die-plate contained in platen and a bore associated with a ram, for receiving source material powder, a filling shoe, and a reservoir for powder connected by a hose to the filling shoe. The device is characterized by a passing wheel in the filling shoe as filling aid means; a tube containing a feedscrew disposed between the reservoir and hose as metering means; the reservoir having a bottom part with a can type place-on part with an opening eccentric to the axis; a coupling part and a cover part are placed on the open part of the can, these parts are also provided with a passageway to the feedscrew eccentric to the longitudinal axis

  8. Anode microbial communities produced by changing from microbial fuel cell to microbial electrolysis cell operation using two different wastewaters

    KAUST Repository

    Kiely, Patrick D.; Cusick, Roland; Call, Douglas F.; Selembo, Priscilla A.; Regan, John M.; Logan, Bruce E.

    2011-01-01

    Conditions in microbial fuel cells (MFCs) differ from those in microbial electrolysis cells (MECs) due to the intrusion of oxygen through the cathode and the release of H2 gas into solution. Based on 16S rRNA gene clone libraries, anode communities in reactors fed acetic acid decreased in species richness and diversity, and increased in numbers of Geobacter sulfurreducens, when reactors were shifted from MFCs to MECs. With a complex source of organic matter (potato wastewater), the proportion of Geobacteraceae remained constant when MFCs were converted into MECs, but the percentage of clones belonging to G. sulfurreducens decreased and the percentage of G. metallireducens clones increased. A dairy manure wastewater-fed MFC produced little power, and had more diverse microbial communities, but did not generate current in an MEC. These results show changes in Geobacter species in response to the MEC environment and that higher species diversity is not correlated with current. © 2010 Elsevier Ltd.

  9. Enhancing the properties of Fischer-Tropsch fuel produced from syngas over Co/SiO2 catalyst: Lubricity and Calorific Value

    Science.gov (United States)

    Doustdar, O.; Wyszynski, M. L.; Mahmoudi, H.; Tsolakis, A.

    2016-09-01

    Bio-fuel produced from renewable sources is considered the most viable alternatives for the replacement of mineral diesel fuel in compression ignition engines. There are several options for biomass derived fuels production involving chemical, biological and thermochemical processes. One of the best options is Fischer Tropsch Synthesis, which has an extensive history of gasoline and diesel production from coal and natural gas. FTS fuel could be one of the best solutions to the fuel emission due to its high quality. FTS experiments were carried out in 16 different operation conditions. Mini structured vertical downdraft fixed bed reactor was used for the FTS. Instead of Biomass gasification, a simulated N2 -rich syngas cylinder of, 33% H2 and 50% N2 was used. FT fuels products were analyzed in GCMS to find the hydrocarbon distributions of FT fuel. Calorific value and lubricity of liquid FT product were measured and compared with commercial diesel fuel. Lubricity has become an important quality, particularly for biodiesel, due to higher pressures in new diesel fuel injection (DFI) technology which demands better lubrication from the fuel and calorific value which is amount of energy released in combustion paly very important role in CI engines. Results show that prepared FT fuel has desirable properties and it complies with standard values. FT samples lubricities as measured by ASTM D6079 standard vary from 286μm (HFRR scar diameter) to 417μm which are less than limit of 520μm. Net Calorific value for FT fuels vary from 9.89 MJ/kg to 43.29 MJ/kg, with six of the samples less than EN 14213 limit of 35MJ/kg. Effect of reaction condition on FT fuel properties was investigated which illustrates that in higher pressure Fischer-Tropsch reaction condition liquid product has better properties.

  10. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    O'Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO 2 and UO 2 ), typically containing 95% or more UO 2 . DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO 2 powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO 2 powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required

  11. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

  12. Optimizing Immobilized Enzyme Performance in Cell-Free Environments to Produce Liquid Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Belfort, Georges [Rensselaer Polytechnic Inst., Troy, NY (United States). Dept. of Chemical and Biological Engineering; Grimaldi, Joseph J. [Rensselaer Polytechnic Inst., Troy, NY (United States). Dept. of Chemical and Biological Engineering

    2015-01-27

    Limitations on biofuel production using cell culture (Escherichia coli, Clostridium, Saccharomyces cerevisiae, brown microalgae, blue-green algae and others) include low product (alcohol) concentrations (≤0.2 vol%) due to feedback inhibition, instability of cells, and lack of economical product recovery processes. To overcome these challenges, an alternate simplified biofuel production scheme was tested based on a cell-free immobilized enzyme system. Using this cell free system, we were able to obtain about 2.6 times higher concentrations of iso-butanol using our non-optimized system as compared with live cell systems. This process involved two steps: (i) converts acid to aldehyde using keto-acid decarboxylase (KdcA), and (ii) produces alcohol from aldehyde using alcohol dehydrogenase (ADH) with a cofactor (NADH) conversion from inexpensive formate using a third enzyme, formate dehydrogenase (FDH). To increase stability and conversion efficiency with easy separations, the first two enzymes were immobilized onto methacrylate resin. Fusion proteins of labile KdcA (fKdcA) were expressed to stabilize the covalently immobilized KdcA. Covalently immobilized ADH exhibited long-term stability and efficient conversion of aldehyde to alcohol over multiple batch cycles without fusions. High conversion rates and low protein leaching were achieved by covalent immobilization of enzymes on methacrylate resin. The complete reaction scheme was demonstrated by immobilizing both ADH and fKdcA and using FDH free in solution. The new system without in situ removal of isobutanol achieved a 55% conversion of ketoisovaleric acid to isobutanol at a concentration of 0.5 % (v/v). Further increases in titer will require continuous removal of the isobutanol using our novel brush membrane system that exhibits a 1.5 fold increase in the separation factor of isobutanol from water versus that obtained for commercial silicone rubber membranes. These bio-inspired brush membranes are based on the

  13. Glass material oxidation and dissolution system: Converting miscellaneous fissile materials to glass

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Ferrada, J.J.

    1996-01-01

    The cold war and the development of nuclear energy have resulted in significant inventories of miscellaneous fissile materials (MFMs). MFMs include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel (SNF), (3) certain hot cell wastes, and (4) many one-of-a-kind materials. Major concerns associated with the long-term management of these materials include: safeguards and nonproliferation issues; health, environment, and safety concerns. waste management requirements; and high storage costs. These issues can be addressed by converting the MFMs to glass for secure, long-term storage or repository disposal; however, conventional glass-making processes require oxide-like feed materials. Converting MFMs to oxide-like materials with subsequent vitrification is a complex and expensive process. A new vitrification process has been invented, the Glass Material Oxidation and Dissolution System (GMODS), which directly converts metals, ceramics, and amorphous solids to glass; oxidizes organics with the residue converted to glass; and converts chlorides to borosilicate glass and a secondary sodium chloride (NaCl) stream. Laboratory work has demonstrated the conversion of cerium (a plutonium surrogate), uranium, Zircaloy, stainless steel, multiple oxides, and other materials to glass. However, significant work is required to develop GMODS further for applications at an industrial scale. If implemented, GMODS will provide a new approach to manage these materials

  14. Processing fissile material mixtures containing zirconium and/or carbon

    Science.gov (United States)

    Johnson, Michael Ernest; Maloney, Martin David

    2013-07-02

    A method of processing spent TRIZO-coated nuclear fuel may include adding fluoride to complex zirconium present in a dissolved TRIZO-coated fuel. Complexing the zirconium with fluoride may reduce or eliminate the potential for zirconium to interfere with the extraction of uranium and/or transuranics from fission materials in the spent nuclear fuel.

  15. Recycling of nuclear matters. Myths and realities. Calculation of recycling rate of the plutonium and uranium produced by the French channel of spent fuel reprocessing

    International Nuclear Information System (INIS)

    Coeytaux, X.; Schneider, M.

    2000-05-01

    The recycling rate of plutonium and uranium are: from the whole of the plutonium separated from the spent fuel ( inferior to 1% of the nuclear matter content) attributed to France is under 50% (under 42 tons on 84 tons); from the whole of plutonium produced in the French reactors is less than 20% (42 tons on 224 tons); from the whole of the uranium separated from spent fuels attributed to France is about 10 % (1600 tons on 16000 tons); from the whole of the uranium contained in the spent fuel is slightly over 5%. (N.C.)

  16. R ampersand D plan for immobilization technologies: fissile materials disposition program. Revision 1.0

    International Nuclear Information System (INIS)

    Shaw, H.F.; Armantrout, G.A.

    1996-09-01

    In the aftermath of the Cold War, the US and Russia have agreed to large reductions in nuclear weapons. To aid in the selection of long- term fissile material management options, the Department of Energy's Fissile Materials Disposition Program (FMDP) is conducting studies of options for the storage and disposition of surplus plutonium (Pu). One set of alternatives for disposition involve immobilization. The immobilization alternatives provide for fixing surplus fissile materials in a host matrix in order to create a solid disposal form that is nuclear criticality-safe, proliferation-resistant and environmentally acceptable for long-term storage or disposal

  17. UF6 fissile mass flow simulation at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; March-Leuba, J.; Valentine, T.E.; Mattingly, J.K.; Uckan, T.; McEvers, J.A.

    1997-01-01

    Basis for measuring fissile mass flow in slurries, liquid, and gaseous streams is activation of a fissile stream by neutrons and then detection of delayed radiation from resulting fission products. This paper describes recent simulation measurements with the first prototype of the system for fissile mass flow measurements with HEU UF 6 gas for use in blenddown facilities. Theory was only 15% higher than actual measured; thus calibration factor would be 0.85. This simulation of HEU gas flow confirms well the understanding of the physical phenomena associated with this measurement system

  18. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    This chapter explains the distinction between fissile and fertile materials, examines briefly the processes involved in fuel manufacture and management, describes the alternative nuclear fuel cycles and considers their advantages and disadvantages. Fuel management is usually divided into three stages; the front end stage of production and fabrication, the back end stage which deals with the fuel after it is removed from the reactor (including reprocessing and waste treatment) and the stage in between when the fuel is actually in the reactor. These stages are illustrated and explained in detail. The plutonium fuel cycle and thorium-uranium-233 fuel cycle are explained. The differences between fuels for thermal reactors and fast reactors are explained. (U.K.)

  19. Operational Characteristics of an Accelerator Driven Fissile Solution System

    International Nuclear Information System (INIS)

    Kimpland, Robert Herbert

    2016-01-01

    Operational characteristics represent the set of responses that a nuclear system exhibits during normal operation. Operators rely on this behavior to assess the status of the system and to predict the consequences of off-normal events. These characteristics largely refer to the relationship between power and system operating conditions. The static and dynamic behavior of a chain-reacting system, operating at sufficient power, is primarily governed by reactivity effects. The science of reactor physics has identified and evaluated a number of such effects, including Doppler broadening and shifts in the thermal neutron spectrum. Often these reactivity effects are quantified in the form of feedback coefficients that serve as coupling coefficients relating the neutron population and the physical mechanisms that drive reactivity effects, such as fissile material temperature and density changes. The operational characteristics of such nuclear systems usually manifest themselves when perturbations between system power (neutron population) and system operating conditions arise. Successful operation of such systems requires the establishment of steady equilibrium conditions. However, prior to obtaining the desired equilibrium (steady-state) conditions, an approach from zero-power (startup) must occur. This operational regime may possess certain limiting system conditions that must be maintained to achieve effective startup. Once steady-state is achieved, a key characteristic of this operational regime is the level of stability that the system possesses. Finally, a third operational regime, shutdown, may also possess limiting conditions of operation that must be maintained. This report documents the operational characteristics of a ''generic'' Accelerator Driven Fissile Solution (ADFS) system during the various operational regimes of startup, steady-state operation, and shutdown. Typical time-dependent behavior for each operational regime will be illustrated, and key system

  20. Method of producing a diesel fuel blend having a pre-determined flash-point and pre-determined increase in cetane number

    Science.gov (United States)

    Waller, Francis Joseph; Quinn, Robert

    2004-07-06

    The present invention relates to a method of producing a diesel fuel blend having a pre-determined flash-point and a pre-determined increase in cetane number over the stock diesel fuel. Upon establishing the desired flash-point and increase in cetane number, an amount of a first oxygenate with a flash-point less than the flash-point of the stock diesel fuel and a cetane number equal to or greater than the cetane number of the stock diesel fuel is added to the stock diesel fuel in an amount sufficient to achieve the pre-determined increase in cetane number. Thereafter, an amount of a second oxygenate with a flash-point equal to or greater than the flash-point of the stock diesel fuel and a cetane number greater than the cetane number of the stock diesel fuel is added to the stock diesel fuel in an amount sufficient to achieve the pre-determined increase in cetane number.

  1. Material correlations and models for the irradiation behavior of fissile and fertile material in SNR-300, Mark-II and KNK II, third core

    International Nuclear Information System (INIS)

    Fenneker; Steinmetz; Toebbe

    1986-07-01

    The report contains the material correlations and models used in the fuel pin design code IAMBUS for the irradiation behavior of PuO 2 -UO 2 fissile materials and UO 2 fertile materials of the SNR-300 Mark-II reload and the KNK II third core. They are applicable for pellet densities of more than 90 % of the theoretical density. The presented models of the fuel behavior and the applied material correlations have been derived either from single experiments or from the comparison of theoretically predicted integral fuel behavior with the results of fuel pin irradiation experiments. The material correlations have been examined and extended in the frame of the collaborations INTERATOM/KWU and INTERATOM/KfK. French and British results were included, when available from the European fast reactor knowledge exchange [de

  2. Fuel analysis of a PBMR

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Francois L, J. L.

    2015-09-01

    In this paper a neutronic analysis of fuel for a Pebble Bed Modular Reactor is presented, based on their composition and geometric distribution, having as main objective the use and utilization of thorium for the production of fuel for the operation of this reactor. For the study of these characteristics is necessary to use a code capable of carry out a reliable calculation of the main parameters of the fuel. Using the Monte Carlo method is suitable for simulating the neutron transport in the reactor core, which is the basis of Serpent code, with which the calculations for the analysis will be made. The results show the desirability of the use of thorium, since presents good conversion levels of fertile material to fissile, to produce U 233 by neutron capture, taking as a very important factor the distribution of materials in the core, which in this work had better results based on the neutron multiplication effective factor, formed by three right circular cylinders circumscribed, making that the core has three areas constituted by a mixture of plutonium oxide in the central and external areas, and thorium oxide in the intermediate area. (Author)

  3. Methods and apparatuses for the development of microstructured nuclear fuels

    Science.gov (United States)

    Jarvinen, Gordon D [Los Alamos, NM; Carroll, David W [Los Alamos, NM; Devlin, David J [Santa Fe, NM

    2009-04-21

    Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission products mixture.

  4. Analysis of Advanced Fuel Kernel Technology

    International Nuclear Information System (INIS)

    Oh, Seung Chul; Jeong, Kyung Chai; Kim, Yeon Ku; Kim, Young Min; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung

    2010-03-01

    The reference fuel for prismatic reactor concepts is based on use of an LEU UCO TRISO fissile particle. This fuel form was selected in the early 1980s for large high-temperature gas-cooled reactor (HTGR) concepts using LEU, and the selection was reconfirmed for modular designs in the mid-1980s. Limited existing irradiation data on LEU UCO TRISO fuel indicate the need for a substantial improvement in performance with regard to in-pile gaseous fission product release. Existing accident testing data on LEU UCO TRISO fuel are extremely limited, but it is generally expected that performance would be similar to that of LEU UO 2 TRISO fuel if performance under irradiation were successfully improved. Initial HTGR fuel technology was based on carbide fuel forms. In the early 1980s, as HTGR technology was transitioning from high-enriched uranium (HEU) fuel to LEU fuel. An initial effort focused on LEU prismatic design for large HTGRs resulted in the selection of UCO kernels for the fissile particles and thorium oxide (ThO 2 ) for the fertile particles. The primary reason for selection of the UCO kernel over UO 2 was reduced CO pressure, allowing higher burnup for equivalent coating thicknesses and reduced potential for kernel migration, an important failure mechanism in earlier fuels. A subsequent assessment in the mid-1980s considering modular HTGR concepts again reached agreement on UCO for the fissile particle for a prismatic design. In the early 1990s, plant cost-reduction studies led to a decision to change the fertile material from thorium to natural uranium, primarily because of a lower long-term decay heat level for the natural uranium fissile particles. Ongoing economic optimization in combination with anticipated capabilities of the UCO particles resulted in peak fissile particle burnup projection of 26% FIMA in steam cycle and gas turbine concepts

  5. Physics design of fissile mass-flow monitoring system

    International Nuclear Information System (INIS)

    Mattingly, J.K.; March-Leuba, J.; Valentine, T.E.; Mihalczo, J.T.; Uckan, T.

    1997-01-01

    The system measures the flow rate and uranium-235 content in liquid or gas streams; it does not penetrate the process piping. A moderated fission neutron source is used to periodicially introduce a burst of thermal neutrons into the fluid stream to induce fission; delayed gamma emissions from the resulting fission fragments are detected by high-efficiency scintillators downstream of the neutron source. The fluid flow rate is measure from the time between initiation of the thermal neutron burst and detection of the fission product gamma emissions, and the U-235 content is inferred from the intensity of the gamma burst detected. Design of the fissile mass flow monitor requires satisfaction of several competing constraints. Efficient operation of the monitor requires that source-induced fission rate and detection efficiency be maximized while the source-induced background rate is simultaneoulsy minimized. Near optical nuclear design of the system was achieved using numerous Monte Carlo calculations and measurements. This paper addresses calculational aspects of the physics design for the system applied to UF 6 gas

  6. Verification of classified fissile material using unclassified attributes

    International Nuclear Information System (INIS)

    Nicholas, N.J.; Fearey, B.L.; Puckett, J.M.; Tape, J.W.

    1998-01-01

    This paper reports on the most recent efforts of US technical experts to explore verification by IAEA of unclassified attributes of classified excess fissile material. Two propositions are discussed: (1) that multiple unclassified attributes could be declared by the host nation and then verified (and reverified) by the IAEA in order to provide confidence in that declaration of a classified (or unclassified) inventory while protecting classified or sensitive information; and (2) that attributes could be measured, remeasured, or monitored to provide continuity of knowledge in a nonintrusive and unclassified manner. They believe attributes should relate to characteristics of excess weapons materials and should be verifiable and authenticatable with methods usable by IAEA inspectors. Further, attributes (along with the methods to measure them) must not reveal any classified information. The approach that the authors have taken is as follows: (1) assume certain attributes of classified excess material, (2) identify passive signatures, (3) determine range of applicable measurement physics, (4) develop a set of criteria to assess and select measurement technologies, (5) select existing instrumentation for proof-of-principle measurements and demonstration, and (6) develop and design information barriers to protect classified information. While the attribute verification concepts and measurements discussed in this paper appear promising, neither the attribute verification approach nor the measurement technologies have been fully developed, tested, and evaluated

  7. Simulator for an Accelerator-Driven Subcritical Fissile Solution System

    International Nuclear Information System (INIS)

    Klein, Steven Karl; Day, Christy M.; Determan, John C.

    2015-01-01

    LANL has developed a process to generate a progressive family of system models for a fissile solution system. This family includes a dynamic system simulation comprised of coupled nonlinear differential equations describing the time evolution of the system. Neutron kinetics, radiolytic gas generation and transport, and core thermal hydraulics are included in the DSS. Extensions to explicit operation of cooling loops and radiolytic gas handling are embedded in these systems as is a stability model. The DSS may then be converted to an implementation in Visual Studio to provide a design team the ability to rapidly estimate system performance impacts from a variety of design decisions. This provides a method to assist in optimization of the system design. Once design has been generated in some detail the C++ version of the system model may then be implemented in a LabVIEW user interface to evaluate operator controls and instrumentation and operator recognition and response to off-normal events. Taken as a set of system models the DSS, Visual Studio, and LabVIEW progression provides a comprehensive set of design support tools.

  8. Simulator for an Accelerator-Driven Subcritical Fissile Solution System

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Day, Christy M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Determan, John C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-14

    LANL has developed a process to generate a progressive family of system models for a fissile solution system. This family includes a dynamic system simulation comprised of coupled nonlinear differential equations describing the time evolution of the system. Neutron kinetics, radiolytic gas generation and transport, and core thermal hydraulics are included in the DSS. Extensions to explicit operation of cooling loops and radiolytic gas handling are embedded in these systems as is a stability model. The DSS may then be converted to an implementation in Visual Studio to provide a design team the ability to rapidly estimate system performance impacts from a variety of design decisions. This provides a method to assist in optimization of the system design. Once design has been generated in some detail the C++ version of the system model may then be implemented in a LabVIEW user interface to evaluate operator controls and instrumentation and operator recognition and response to off-normal events. Taken as a set of system models the DSS, Visual Studio, and LabVIEW progression provides a comprehensive set of design support tools.

  9. Portal monitoring for detecting fissile materials and chemical explosives

    International Nuclear Information System (INIS)

    Albright, D.

    1992-01-01

    The portal monitoring of pedestrians, packages, equipment, and vehicles entering or leaving areas of high physical security has been common for many years. Many nuclear facilities rely on portal monitoring to prevent the theft or diversion of plutonium and highly enriched uranium. At commercial airports, portals are used to prevent firearms and explosives from being smuggled onto airplanes. An August 1989 Federal Aviation Administration (FAA) regulation requires US airlines to screen luggage on international flights for chemical explosives. This paper reports that portal monitoring is now being introduced into arms-control agreements. Because some of the portal-monitoring equipment that would be useful in verifying arms-control agreements is already widely used as part of the physical security systems at nuclear facilities and commercial airports, the authors review these uses of portal monitoring, as well as its role in verifying the INF treaty. Then the authors survey the major types of portal-monitoring equipment that would be most useful in detecting nuclear warheads or fissile material

  10. Immobilization as a route to surplus fissile materials disposition

    International Nuclear Information System (INIS)

    Gray, L.W.; Kan, T.

    1995-01-01

    In the aftermath of the Cold War, the US and Russia have agreed to large reductions in nuclear weapons. To aid in the selection of long-term management options, DOE has undertaken a multifaceted study to select options for storage and disposition of plutonium (Pu) in keeping with the national policy that Pu must be subjected to the highest standards of safety, security, and accountability. One alternative being considered is immobilization. To arrive at a suitable immobilization form, the authors first reviewed published information on high-level waste (HLW) immobilization technologies in order to identify 72 possible Pu immobilization forms to be prescreened. Surviving forms were screened using multiattribute analysis to determine the most promising technologies. Promising immobilization families were further evaluated to identify chemical, engineering, environmental, safety, and health problems that remain to be solved prior to making technical decisions as to the viability of using the form for long-term disposition of plutonium. All data, analyses, and reports are being provided to the DOE Fissile Materials Disposition Project Office to support the Record of Decision that is anticipated in the fourth quarter of FY96

  11. The mass transfer mechanism of fissile material due to fission

    International Nuclear Information System (INIS)

    Shafrir, N.H.

    1975-01-01

    A thin 252 Cf source of a mean thickness of an approXimately mono-atomic layer was used as an experimental model for the study of the basic mechanism of the knock-on process taking place in fissile material. Because of the thinness of the source it can be assumed that mainly primary knock-ons are formed. The ejection rate of knock-ons created by direct collisions between fission fragments and source atoms was measured as follows: the ejected atoms were collected in high vacuum on a catcher foil and 252 Cf determined by alpha spectroscopy using a silicon surface barrier detector. The number of 252 Cf ejected from the source in unit time could thus be determined while considering the anisotropy of ejection, geometry and counting efficiency. Taking into account the chemical composition of the source, eta(theor.) = 252 Cf atoms/fission was obtained. This result can be considered in reasonable agreement with experiment confirming that under the experimental conditions described, practically no knock-on cascade is formed. (B.G.)

  12. Fast-neutron capture in fissile and fertile nuclides

    International Nuclear Information System (INIS)

    Peelle, R.W.

    1982-01-01

    Extensive graphical and numerical presentations, available to the working group, assisted us in exploring the rich data base established through the labors of many skilled persons. Consistent with the meeting setting, the working group discussion concentrated on data for fast-breeder reactor (FBR) applications. All but 1 to 3% of the magnitude of cross section sensitivities of FBR parameters come from the energy region below approx. = 1.5 MeV, so the statistical model is the relevant theoretical concept. The Meeting emphasizes energies above approx. = 10 keV where resonance fluctuations are not a dominant factor. However, we should remember that approximately half the FBR sensitivity to 238 U capture data, as relfected in integral parameters, lies below 25 keV where resonance fluctuations are strong and resonance self-protection is a most important consideration in reactor physics. There are similar low-energy aspects to 239 Pu capture in that approx. = 30% of the FBR-parameter data sensitivity lies below approx. = 4 keV. Even with the discussion largely cofined to the approx. = 10 to 1500 keV region, the working group could only scratch the surface of the available body of information. The reader is referred to the papers presented at the Meeting and to the references contained therein in order to obtain a more detailed understanding of current issues related to fissile and fertile fast-neutron capture

  13. Gasification of heavy fuels to produce electrical energy and hydrogen; Gasificacion de combustibles pesados para producir energia electrica e hidrogeno

    Energy Technology Data Exchange (ETDEWEB)

    Vera Garcia, Oscar Alberto [Universidad Nacional Autonoma de Mexico(UNAM), Mexico, D.F. (Mexico)

    2006-11-15

    A description is presented of the different types of integrated gasifiers that at the moment are used in the synthesis gas production to be used, with different fuels in the generation of electricity in Combined Cycle. Three cases of application of integrated gasifiers are analyzed. The first it is the engine power upgrade of a Combined Cycle power plant to natural gas to burn fuel of bad quality in an integrated gasifier (CCGI). The second one examines the incorporation of a shift reactor in which the synthesis gas is transformed into CO{sub 2} and H{sub 2} which are used to move the turbine to gas, adapted for pure hydrogen. Finally is studied the amount of other by-products that can be obtained from these co-generation cycles such as CO{sub 2} to be used in secondary recovery of oil wells, N{sub 2} to be used in the fertilizer industry or in the proper oil production and H{sub 2} to be used in the oil industry or the generation with fuel cells. All the cases are studied in quantitative form, making the balance of mass and energy of each one of them. In order to give more practical sense to the calculations, the engineering data of the Valladolid Power station of Comision Federal de Electricidad (CFE) have been taken as base. This article provides a basic idea, but very practical, to estimate the fuel consumption of the different modes of arrangement of a CCGI power station, as well as the volumes of the different gases that can be produced and the modifications to the size of the equipment that is required. [Spanish] Se presenta una descripcion de los diferentes tipos de gasificadores integrados que actualmente se utilizan en la produccion de gas de sintesis para ser utilizados, con diferentes combustibles, en la generacion de electricidad con Ciclo Combinado. Se analizan tres casos de aplicacion de gasificadores integrados. El primero es la repotenciacion de una planta de Ciclo Combinado a gas natural para quemar combustible de mala calidad en un gasificador

  14. Accelerator molten-salt breeding and thorium fuel cycle

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Nakahara, Yasuaki; Kato, Yoshio; Ohno, Hideo; Mitachi, Kohshi.

    1990-01-01

    The recent efforts at the development of fission energy utilization have not been successful in establishing fully rational technology. A new philosophy should be established on the basis of the following three principles: (1) thorium utilization, (2) molten-salt fuel concept, and (3) separation of fissile-breeding and power-generating functions. Such philosophy is called 'Thorium Molten-Salt Nuclear Energy Synergetics [THORIMS-NES]'. The present report first addresses the establishment of 233 U breeding fuel cycle, focusing on major features of the Breeding and Chemical Processing Centers and a small molten-salt power station (called FUJI-II). The development of fissile producing breeders is discussed in relation to accelerator molten-salt breeder (AMSB), impact fusion molten-salt breeder, and inertial-confined fusion hybrid molten-salt breeder. Features of the accelerator molten-salt breeder are described, focusing on technical problems with accelerator breeders (or spallators), design principle of the accelerator molten-salt breeder, selection of molten salt compositions, and nuclear- and reactor-chemical aspects of AMSB. Discussion is also made of further research and development efforts required in the future for AMSB. (N.K.)

  15. Evaluation of spent fuel properties from a conceptual PEACER core

    International Nuclear Information System (INIS)

    Lim, Jae Yong; Kim, Myung Hyun; Kim, Chang Hyo; Hwang, Il Soon

    2003-01-01

    In this paper, a new conceptual core design, PEACER was evaluated in aspect of core performance and spent fuel properties. The core shape is like a pancake to increase axial neutron leakage. Square lattice array was applied which was suitable to decrease the flow speed of Pb-Bi coolant. Although over 30% TRU produced by pyroprocessing was loaded in U-Zr metal fuel, the cycle length of 1 year was achieved and the relative assembly power peaking was less than 1.3. In order to confirm nuclear performance of PEACER core design, several performance indices were adopted and developed. Simple indices such as FIR and FG were used to evaluate fissile breeding. BCM, TG, SNS, and OR calculated by plutonium composition vectors were chosen to distinguish the competency of proliferation resistance. For the estimation of transmutation capability, D-value and extended effective fission half-life time(T EX ) were used. According to these indices, the PEACER core had the better performance compared with other conventional reactor cores although fissile breeding was not acquired

  16. Detection of tiny amounts of fissile materials in large-sized containers with radioactive waste

    Science.gov (United States)

    Batyaev, V. F.; Skliarov, S. V.

    2018-01-01

    The paper is devoted to non-destructive control of tiny amounts of fissile materials in large-sized containers filled with radioactive waste (RAW). The aim of this work is to model an active neutron interrogation facility for detection of fissile ma-terials inside NZK type containers with RAW and determine the minimal detectable mass of U-235 as a function of various param-eters: matrix type, nonuniformity of container filling, neutron gen-erator parameters (flux, pulse frequency, pulse duration), meas-urement time. As a result the dependence of minimal detectable mass on fissile materials location inside container is shown. Nonu-niformity of the thermal neutron flux inside a container is the main reason of the space-heterogeneity of minimal detectable mass in-side a large-sized container. Our experiments with tiny amounts of uranium-235 (<1 g) confirm the detection of fissile materials in NZK containers by using active neutron interrogation technique.

  17. Re-energizing energy supply: Electrolytically-produced hydrogen as a flexible energy storage medium and fuel for road transport

    Science.gov (United States)

    Emonts, Bernd; Schiebahn, Sebastian; Görner, Klaus; Lindenberger, Dietmar; Markewitz, Peter; Merten, Frank; Stolten, Detlef

    2017-02-01

    "Energiewende", which roughly translates as the transformation of the German energy sector in accordance with the imperatives of climate change, may soon become a byword for the corresponding processes most other developed countries are at various stages of undergoing. Germany's notable progress in this area offers valuable insights that other states can draw on in implementing their own transitions. The German state of North Rhine-Westphalia (NRW) is making its own contribution to achieving the Energiewende's ambitious objectives: in addition to funding an array of 'clean and green' projects, the Virtual Institute Power to Gas and Heat was established as a consortium of seven scientific and technical organizations whose aim is to inscribe a future, renewable-based German energy system with adequate flexibility. Thus, it is tasked with conceiving of and evaluating suitable energy path options. This paper outlines one of the most promising of these pathways, which is predicated on the use of electrolytically-produced hydrogen as an energy storage medium, as well as the replacement of hydrocarbon-based fuel for most road vehicles. We describe and evaluate this path and place it in a systemic context, outlining a case study from which other countries and federated jurisdictions therein may draw inspiration.

  18. Fission of 209 Bi by 60-270 MeV tagged photons: cross section measurement and analysis of photo fissility

    International Nuclear Information System (INIS)

    Terranova, M.L.; Tavares, O.A.P.

    1996-07-01

    Tagged photons produced by the ROKK-2 facility have been used to measure the photofission cross section of 209 Bi in the energy range 60-270 MeV. Photofission events were detected by using a nuclear fragment detector designed for fission experiments, based on multiwire spark counters. Fissility values have been deduced and compared with available data obtained in other laboratories by using monochromatic photons. These data, together with early measurements obtained near photofission threshold, have been analysed in the framework of a two-step model which considers the primary photo interaction occurring via the quasi-deuteron and/or photo mesonic processes, followed by a mechanism of evaporation-fission competition for the excited residual nucleus. The model was found to reproduce the main experimental features of 209 Bi photo fissility up to 300 MeV. (author). 52 refs., 7 figs., 2 tabs

  19. Update to the Fissile Materials Disposition program SST/SGT transportation estimation

    International Nuclear Information System (INIS)

    John Didlake

    1999-01-01

    This report is an update to ''Fissile Materials Disposition Program SST/SGT Transportation Estimation,'' SAND98-8244, June 1998. The Department of Energy Office of Fissile Materials Disposition requested this update as a basis for providing the public with an updated estimation of the number of transportation loads, load miles, and costs associated with the preferred alternative in the Surplus Plutonium Disposition Final Environmental Impact Statement (EIS)

  20. Uranium requirements for advanced fuel cycles in expanding nuclear power systems

    International Nuclear Information System (INIS)

    Banerjee, S.; Tamm, H.

    1978-01-01

    When considering advanced fuel cycle strategies in rapidly expanding nuclear power systems, equilibrium analyses do not apply. A computer simulation that accounts for system delay times and fissile inventories has been used to study the effects of different fuel cycles and different power growth rates on uranium consumption. The results show that for a given expansion rate of installed capacity, the main factors that affect resource requirements are the fissile inventory needed to introduce the advanced fuel cycle and the conversion (or breeding) ratio. In rapidly expanding systems, the effect of fissile inventory dominates, whereas in slowly expanding systems, conversion or breeding ratio dominates. Heavy-water-moderated and -cooled reactors, with their high conversion ratios, appear to be adaptable vehicles for accommodating fuel cycles covering a wide range of initial fissile inventories. They are therefore particularly suitable for conserving uranium over a wide range of nuclear power system expansion rates

  1. Irradiation performance of (Th,U)O2 fuel designed for advanced cycle applications

    International Nuclear Information System (INIS)

    Hastings, I.J.; Celli, A.; Onofrei, M.; Swanson, M.L.

    1982-06-01

    Our reference fabrication route for Advanced Cycle thoria-based fuel is conventional in that it produces cold-pressed and sintered pellets. However, we are also evaluating alternative fuels which offer the potential for simpler fabrication in a remote facility, and in some cases improved high burnup performance. These alternatives are impregnated, spherepac, and extruded thoria-based fuels. Spherepac fuel has been irradiated at a linear power of 50-60 kW/m to about 180 MW.h/kg H.E. There have been unexplained defects in fuel with both free-standing and collapsible cladding. Impegnated fuel has operated to 650 MW.h/kg H.E. at 50-60 kW/m. An experiment examining fuel from the sol-gel extrusion process has reached 450 Mw.h/kg H.E. at a maximum linear power of 60 kW/m. The latter two experiments have operated without defects and with fission gas release less than that for UO 2 under identical conditions. The extruded fuel has a pellet geometry similar to that for conventional fuel and is AECL's first practical demonstration of thoria-based fuel with the fissile component distributed homogeneously on an atomic scale

  2. Modern new nuclear fuel characteristics and radiation protection aspects.

    Science.gov (United States)

    Terry, Ian R

    2005-01-01

    The glut of fissile material from reprocessing plants and from the conclusion of the cold war has provided the opportunity to design new fuel types to beneficially dispose of such stocks by generating useful power. Thus, in addition to the normal reactor core complement of enriched uranium fuel assemblies, two other types are available on the world market. These are the ERU (enriched recycled uranium) and the MOX (mixed oxide) fuel assemblies. Framatome ANP produces ERU fuel assemblies by taking feed material from reprocessing facilities and blending this with highly enriched uranium from other sources. MOX fuel assemblies contain plutonium isotopes, thus exploiting the higher neutron yield of the plutonium fission process. This paper describes and evaluates the gamma, spontaneous and alpha reaction neutron source terms of these non-irradiated fuel assembly types by defining their nuclear characteristics. The dose rates which arise from these terms are provided along with an overview of radiation protection aspects for consideration in transporting and delivering such fuel assemblies to power generating utilities.

  3. Impact on burnup performance of coated particle fuel design in pebble bed reactor with ROX fuel

    International Nuclear Information System (INIS)

    Ho, Hai Quan; Obara, Toru

    2015-01-01

    The pebble bed reactor (PBR), a kind of high-temperature gas-cooled reactor (HTGR), is expected to be among the next generation of nuclear reactors as it has excellent passive safety features, as well as online refueling and high thermal efficiency. Rock-like oxide (ROX) fuel has been studied at the Japan Atomic Energy Agency (JAEA) as a new once-through type fuel concept. Rock-like oxide used as fuel in a PBR can be expected to achieve high burnup and improve chemical stabilities. In the once-through fuel concept, the main challenge is to achieve as high a burnup as possible without failure of the spent fuel. The purpose of this study was to investigate the impact on burnup performance of different coated fuel particle (CFP) designs in a PBR with ROX fuel. In the study, the AGR-1 Coated Particle design and Deep-Burn Coated Particle design were used to make the burnup performance comparison. Criticality and core burnup calculations were performed by MCPBR code using the JENDL-4.0 library. Results at equilibrium showed that the two reactors utilizing AGR-1 Coated Particle and Deep-Burn Coated Particle designs could be critical with almost the same multiplication factor k eff . However, the power peaking factor and maximum power per fuel ball in the AGR-1 coated particle design was lower than that of Deep-Burn coated particle design. The AGR-1 design also showed an advantage in fissions per initial fissile atoms (FIFA); the AGR-1 coated particle design produced a higher FIFA than the Deep-Burn coated particle design. These results suggest that the difference in coated particle fuel design can have an effect on the burnup performance in ROX fuel. (author)

  4. Cross section measurements of fissile nuclei for slow neutrons; Mesures de sections efficaces de noyaux fissiles pour les neutrons lents

    Energy Technology Data Exchange (ETDEWEB)

    Auclair, J M; Hubert, P; Joly, R; Vendryes, G; Jacrot, B; Netter, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Galula, M [Centre National de la Recherche Scientifique (CNRS), 91 - Gif-sur-Yvette (France)

    1955-07-01

    It presents the experimental measurements of cross section of fissile nuclei for slow neutrons to improve the understanding of some heavy nuclei of great importance in the study of nuclear reactors. The different experiments are divided in three categories. In the first part, it studied the variation with energy of the cross sections of natural uranium, {sup 233}U, {sup 235}U and {sup 239}Pu. Two measurement techniques are used: the time-of-flight spectrometer and the crystal spectrometer. In a second part, the fission cross sections of {sup 233}U and {sup 239}Pu for thermal neutrons are compared using a neutron flux from EL-2 going through a double fission chamber. The matter quantity contained in each source is measured by counting the {alpha} activity with a solid angle counter. Finally, the average cross section of {sup 236}U for a spectra of neutrons from the reactor is measured by studying the {beta} activity of {sup 237}U formed by the reaction {sup 236}U (n, {gamma}) {sup 237}U in a sample of {sup 236}U irradiated in the Saclay reactor (EL-2). (M.P.)

  5. Prospective studies of HTR fuel cycles involving plutonium

    International Nuclear Information System (INIS)

    Bonin, B.; Greneche, D.; Carre, F.; Damian, F.; Doriath, J.Y.

    2002-01-01

    High Temperature Gas Cooled reactors (HTRs) are able to accommodate a wide variety of mixtures of fissile and fertile materials without any significant modification of the core design. This flexibility is due to an uncoupling between the parameters of cooling geometry, and the parameters which characterize neutronic optimisation (moderation ratio or heavy nuclide concentration and distribution). Among other advantageous features, an HTR core has a better neutron economy than a LWR because there is much less parasitic capture in the moderator (capture cross section of graphite is 100 times less than the one of water) and in internal structures. Moreover, thanks to the high resistance of the coated particles, HTR fuels are able to reach very high burn-ups, far beyond the possibilities offered by other fuels (except the special case of molten salt reactors). These features make HTRs especially interesting for closing the nuclear fuel cycle and stabilizing the plutonium inventory. A large number of fuel cycle studies are already available today, on 3 main categories of fuel cycles involving HTRs : i) High enriched uranium cycle, based on thorium utilization as a fertile material and HEU as a fissile material; ii) Low enriched uranium cycle, where only LEU is used (from 5% to 12%); iii) Plutonium cycle based on the utilization of plutonium only as a fissile material, with (or without) fertile materials. Plutonium consumption at high burnups in HTRs has already been tested with encouraging results under the DRAGON project and at Peach Bottom. To maximize plutonium consumption, recent core studies have also been performed on plutonium HTR cores, with special emphasis on weapon-grade plutonium consumption. In the following, we complete the picture by a core study for a HTR burning reactor-grade plutonium. Limits in burnup due to core neutronics are investigated for this type of fuel. With these limits in mind, we study in some detail the Pu cycle in the special case of a

  6. Bio-based production of fuels and industrial chemicals by repurposing antibiotic-producing type I modular polyketide synthases: opportunities and challenges

    DEFF Research Database (Denmark)

    Yuzawa, Satoshi; Keasling, Jay D.; Katz, Leonard

    2017-01-01

    Complex polyketides comprise a large number of natural products that have broad application in medicine and agriculture. They are produced in bacteria and fungi from large enzyme complexes named type I modular polyketide synthases (PKSs) that are composed of multifunctional polypeptides containin...... have applications as fuels or industrial chemicals....

  7. FIELD-PRODUCED JP-8 STANDARD FOR CALIBRATION OF LOWER EXPLOSIVE LIMIT METERS USED BY JET FUEL TANK MAINTENANCE PERSONNEL

    Science.gov (United States)

    Thousands of military personnel and tens of thousands of civilian workers perform jet fuel tank entry procedures. Before entering the confined space of a jet fuel tank, OSHA regulations (29CFR1910.146) require the internal atmosphere be tested with a calibrated, direct-reading...

  8. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    Henriquez, C; Navarro, G; Pereda, C

    2000-01-01

    Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

  9. Quantification of Fissile Materials by Photon Activation Method in a Highly Shielded Enclosure

    International Nuclear Information System (INIS)

    Dighe, P.M.; Pithawa, C.K.; Goswami, A.; Dixit, K.P.; Mittal, K.C.; Sunil, C.; Sarkar, P.K.; Mukhopadhyay, P.K.; Patil, R.K.; Srivastava, G.P.; Ganesan, S.; Venugopal, V.

    2010-01-01

    For active and non-destructive quantitative identification of heavily shielded fissile materials, photo fission is one of the most often used techniques. High energy photon beams can be conveniently generated with the help of electron LINACs. 10MeV energy electron LINACs are extensively used for various industrial applications such as food irradiation, X-ray radiography, etc. The radiological safety consideration favours the use of electron beam of upto 10 MeV energy. The photonuclear data available on 10 MeV end point energy is very scarce. The present paper gives the results of our initial experiments carried out using natural uranium samples at 10 MeV LINAC facility. Water cooled tantalum target converter was used to produce intense Bremsstrahlung to induce photofission in the samples. Neutron detection system consists of six numbers of high sensitivity Helium-3 proportional counters and gamma detection system consists of two numbers of 76 mm diameter BGO scintillators. Delayed neutron and delayed gamma radiations were measured and analyzed. The mass to count rate relationship has been established for both delayed neutron and gamma radiations. Delayed gamma decay constants of natural uranium have been derived for the 10 MeV end point energy. (author)

  10. Current status and recommended future studies of underground supercriticality of fissile material

    International Nuclear Information System (INIS)

    Bowman, C.D.

    1996-06-01

    More than a year has passed since we released our original report pointing out the possibility of natural or induced rearrangement of fissile material underground into a critical mass, the possibility of positive feedback in underground configurations, the confinement of the rock to produce significant yield, and the possibility of venting or explosion. The nuclear weapons and repository storage groups at both Los Alamos and Livermore have been critical of our work while others have defended our calculations on wet and dry criticality. The conditions we identified for positive and negative feedback are no longer contested. The role of confinement of the rock in enhancing the yield from the explosion is still unsettled, and that is addressed later in this paper. The likelihood of confinement, venting, or explosive dispersion also remains unsettled and that is addressed here as well. Some critics of our work have tried to show that the probability of reconfiguration by natural processes is very small. They argue further that emplacement can be done in such a way as to make the probability even smaller. Of course these additional efforts will raise the cost of waste emplacement and the question arises as to how much is enough. The answer to this question seems to not be an easy one

  11. 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

  12. ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    International Nuclear Information System (INIS)

    O'Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1997-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program's preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO 2 and UO 2 ), typically containing 95% or more UO 2 . DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement

  13. ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1997-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement.

  14. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  15. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Jin Ha; Kim, Myung Hyun

    2016-01-01

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  16. Criticality safety issues arising from the treatment of liquid effluent streams from the reprocessing of thermal oxide fuel

    International Nuclear Information System (INIS)

    Thorne, P.R.; Farrington, L.M.

    1991-01-01

    The BNFL THORP plant will reprocess irradiated oxide fuel from thermal reactors to recover plutonium dioxide and uranium trioxide in a pure form. A consequence of the reprocessing is that several liquid effluent streams are produced which can contain residual fissile material. Generally, the treatment of these effluent streams is carried out in large vessels which are not geometrically favourable with regard to nuclear safety. This is possible because the concentration of fissile material in solution is far less than the safely subcritical infinite sea concentrations. The situation is complicated by the presence of precipitated solids in some vessels and crud layers in others. Experimental measurements have been used to characterise these solids in order to extend the usual safe limits, and to provide an acceptable operating regime. Based on the experimental characterisation of the solids, the neutronics computer codes WIMS and MONK have been used to determine the optimum possible conditions existing, and to determine the safe fissile mass limits for these systems. The limits which are derived have been used to provide alarm and trip levels for instrumentation which has been employed in a novel way. It has been shown that the plant can be operated successfully and remains acceptably safe taking into account the presence of solids in the liquid effluent streams. (author)

  17. Theoretical, physical and experimental study of fissile aqueous media; Etudes theorique, physique et experimentale des milieux fissiles aqueux

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-04-01

    This thesis consists of a set of theoretical and experimental studies. a) Theoretical calculation methods used for cross-sections and the critical parameters; b) Comparison of the theoretical and experimental results: it is shown that the agreement between these results cannot be improved above a certain limit because of the accuracy with which are known the composition and the dimensions of the media and the microscopic cross-sections; c) Determination of the ratios {eta}{sup 9}-bar / {eta}{sup 5}-bar, {eta}{sup 3}-bar / {eta}{sup 5}-bar for fissile aqueous media ({eta}-bar: number of neutrons emitted per neutron absorbed, averaged over the reactor neutron spectrum). Evaluation of the accuracy to which these ratios are known; d) Effect of {sup 240}Pu: the measurements are carried out on Pu with a {sup 240}Pu content of 1.5 per cent, 3.11 per cent and 9.95 per cent; Calculation of the resonance integral I240 using the experimental results gives values in reasonable agreement with the results obtained by other more conventional methods. e) Measurement of the spectrum indices for aqueous media containing Pu, U5 and U3. With these latter it is possible to obtain mean fission cross-section ratios {sigma}f239-bar / {sigma}f235-bar for these different spectra. A calculation-experiment comparison is carried out using various theoretical methods. (author) [French] Cette these groupe un ensemble d'etudes theoriques et experimentales. a) Methodes theoriques de calcul utilisees pour les sections efficaces et les parametres critiques; b) Comparaisons des resultats theoriques et experimentaux: on montre que l'accord entre ces resultats ne peut etre ameliore au-dela de certaines limites vu la precision avec laquelle sont connues la composition et les dimensions des milieux et les sections efficaces macroscopiques; c) Determination des rapports {eta}{sup 9}-bar / {eta}{sup 5}-bar, {eta}{sup 3}-bar / {eta}{sup 5}-bar pour les milieux fissiles aqueux ({eta}: nombre de

  18. Theoretical, physical and experimental study of fissile aqueous media; Etudes theorique, physique et experimentale des milieux fissiles aqueux

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-04-01

    This thesis consists of a set of theoretical and experimental studies. a) Theoretical calculation methods used for cross-sections and the critical parameters; b) Comparison of the theoretical and experimental results: it is shown that the agreement between these results cannot be improved above a certain limit because of the accuracy with which are known the composition and the dimensions of the media and the microscopic cross-sections; c) Determination of the ratios {eta}{sup 9}-bar / {eta}{sup 5}-bar, {eta}{sup 3}-bar / {eta}{sup 5}-bar for fissile aqueous media ({eta}-bar: number of neutrons emitted per neutron absorbed, averaged over the reactor neutron spectrum). Evaluation of the accuracy to which these ratios are known; d) Effect of {sup 240}Pu: the measurements are carried out on Pu with a {sup 240}Pu content of 1.5 per cent, 3.11 per cent and 9.95 per cent; Calculation of the resonance integral I240 using the experimental results gives values in reasonable agreement with the results obtained by other more conventional methods. e) Measurement of the spectrum indices for aqueous media containing Pu, U5 and U3. With these latter it is possible to obtain mean fission cross-section ratios {sigma}f239-bar / {sigma}f235-bar for these different spectra. A calculation-experiment comparison is carried out using various theoretical methods. (author) [French] Cette these groupe un ensemble d'etudes theoriques et experimentales. a) Methodes theoriques de calcul utilisees pour les sections efficaces et les parametres critiques; b) Comparaisons des resultats theoriques et experimentaux: on montre que l'accord entre ces resultats ne peut etre ameliore au-dela de certaines limites vu la precision avec laquelle sont connues la composition et les dimensions des milieux et les sections efficaces macroscopiques; c) Determination des rapports {eta}{sup 9}-bar / {eta}{sup 5}-bar, {eta}{sup 3}-bar / {eta}{sup 5}-bar pour les milieux fissiles aqueux ({eta}: nombre de neutrons emis

  19. The modular ALMR (PRISM) fuel cycle

    International Nuclear Information System (INIS)

    Thompson, M.L.

    1990-01-01

    The modular reactor concept, PRISM (power reactor, innovative, small module), originated by General Electric in conjunction with the integral fast reactor (IFR) metal fuel being developed by Argonne National Laboratory (ANL), is the reference US Department of Energy advanced liquid-metal reactor (ALMR). The reference ALMR is unique in several ways; for example, it can produce (or breed) substantially more fissile material than it consumes. It is also unique in that it has the capability to utilize as fuel the long-life radioactive actinides (primarily plutonium, and the minor actinides, neptunium, americium, and curium) present as waste in light water reactor (LWR) spent fuels. This capability provides a means for converting long-life actinide radioactive wastes to elements whose lifetimes and thus storage needs are much shorter, namely, hundreds of years. This could clearly focus and potentially alleviate a controversial aspect (waste disposal) of the nuclear option. While it does not change the need for, or timing of, an initial high-level waste (HLW) repository, the conversion of actinides could change in a dramatic way the time period required for safe storage of nuclear waste and potentially the number and criteria for future repositories. This work considers the potential for utilizing LWR actinides in the ALMR fuel cycle

  20. Bio-based production of fuels and industrial chemicals by repurposing antibiotic-producing type I modular polyketide synthases: opportunities and challenges.

    Science.gov (United States)

    Yuzawa, Satoshi; Keasling, Jay D; Katz, Leonard

    2017-04-01

    Complex polyketides comprise a large number of natural products that have broad application in medicine and agriculture. They are produced in bacteria and fungi from large enzyme complexes named type I modular polyketide synthases (PKSs) that are composed of multifunctional polypeptides containing discrete enzymatic domains organized into modules. The modular nature of PKSs has enabled a multitude of efforts to engineer the PKS genes to produce novel polyketides of predicted structure. We have repurposed PKSs to produce a number of short-chain mono- and di-carboxylic acids and ketones that could have applications as fuels or industrial chemicals.

  1. Bio-based production of fuels and industrial chemicals by repurposing antibiotic-producing type I modular polyketide synthases: opportunities and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Yuzawa, Satoshi [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Biological Systems and Engineering Division; Keasling, Jay D. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Biological Systems and Engineering Division; Univ. of California, Berkeley, CA (United States). QB3 Inst.; Joint BioEnergy Inst. (JBEI), Emeryville, CA (United States); Univ. of California, Berkeley, CA (United States). Dept. of Bioengineering; Univ. of California, Berkeley, CA (United States). Dept. of Chemical and Biomolecular Engineering; Technical Univ. of Denmark, Horsholm (Denmark). Novo Nordisk Foundation Center for Biosustainability; Katz, Leonard [Univ. of California, Berkeley, CA (United States). QB3 Inst.

    2016-11-16

    Complex polyketides comprise a large number of natural products that have broad application in medicine and agriculture. They are produced in bacteria and fungi from large enzyme complexes named type I modular polyketide synthases (PKSs) that are composed of multifunctional polypeptides containing discrete enzymatic domains organized into modules. The modular nature of PKSs has enabled a multitude of efforts to engineer the PKS genes to produce novel polyketides of predicted structure. Finally, we have repurposed PKSs to produce a number of short-chain mono- and di-carboxylic acids and ketones that could have applications as fuels or industrial chemicals.

  2. Open literature review of threats including sabotage and theft of fissile material transport in Japan

    International Nuclear Information System (INIS)

    Cochran, John Russell; Furaus, James Phillip; Marincel, Michelle K.

    2005-01-01

    This report is a review of open literature concerning threats including sabotage and theft related to fissile material transport in Japan. It is intended to aid Japanese officials in the development of a design basis threat. This threat includes the external threats of the terrorist, criminal, and extremist, and the insider threats of the disgruntled employee, the employee forced into cooperation via coercion, the psychotic employee, and the criminal employee. Examination of the external terrorist threat considers Japanese demographics, known terrorist groups in Japan, and the international relations of Japan. Demographically, Japan has a relatively homogenous population, both ethnically and religiously. Japan is a relatively peaceful nation, but its history illustrates that it is not immune to terrorism. It has a history of domestic terrorism and the open literature points to the Red Army, Aum Shinrikyo, Chukaku-Ha, and Seikijuku. Japan supports the United States in its war on terrorism and in Iraq, which may make Japan a target for both international and domestic terrorists. Crime appears to remain low in Japan; however sources note that the foreign crime rate is increasing as the number of foreign nationals in the country increases. Antinuclear groups' recent foci have been nuclear reprocessing technology, transportation of MOX fuel, and possible related nuclear proliferation issues. The insider threat is first defined by the threat of the disgruntled employee. This threat can be determined by studying the history of Japan's employment system, where Keiretsu have provided company stability and lifetime employment. Recent economic difficulties and an increase of corporate crime, due to sole reliability on the honor code, have begun to erode employee loyalty

  3. Repository emplacement costs for Al-clad high enriched uranium spent fuel

    International Nuclear Information System (INIS)

    McDonell, W.R.; Parks, P.B.

    1994-01-01

    A range of strategies for treatment and packaging of Al-clad high-enriched uranium (HEU) spent fuels to prevent or delay the onset of criticality in a geologic repository was evaluated in terms of the number of canisters produced and associated repository costs incurred. The results indicated that strategies in which neutron poisons were added to consolidated forms of the U-Al alloy fuel generally produced the lowest number of canisters and associated repository costs. Chemical processing whereby the HEU was removed from the waste form was also a low cost option. The repository costs generally increased for isotopic dilution strategies, because of the substantial depleted uranium added. Chemical dissolution strategies without HEU removal were also penalized because of the inert constituents in the final waste glass form. Avoiding repository criticality by limiting the fissile mass content of each canister incurred the highest repository costs

  4. Criticality issues with highly enriched fuels in a repository environment

    International Nuclear Information System (INIS)

    Taylor, L.L.; Sanchez, L.C.; Rath, J.S.

    1998-03-01

    This paper presents preliminary analysis of a volcanic tuff repository containing a combination of low enrichment commercial spent nuclear fuels (SNF) and DOE-owned SNF packages. These SNFs were analyzed with respect to their criticality risks. Disposal of SNF packages containing significant fissile mass within a geologic repository must comply with current regulations relative to criticality safety during transportation and handling within operational facilities. However, once the repository is closed, the double contingency credits for criticality safety are subject to unremediable degradation, (e.g., water intrusion, continued presence of neutron absorbers in proximity to fissile material, and fissile material reconfiguration). The work presented in this paper focused on two attributes of criticality in a volcanic tuff repository for near-field and far-field scenarios: (1) scenario conditions necessary to have a criticality, and (2) consequences of a nuclear excursion that are components of risk. All criticality consequences are dependent upon eventual water intrusion into the repository and subsequent breach of the disposal package. Key criticality parameters necessary for a critical assembly are: (1) adequate thermal fissile mass, (2) adequate concentration of fissile material, (3) separation of neutron poison from fissile materials, and (4) sufficient neutron moderation (expressed in units of moderator to fissile atom ratios). Key results from this study indicated that the total energies released during a single excursion are minimal (comparable to those released in previous solution accidents), and the maximum frequency of occurrence is bounded by the saturation and temperature recycle times, thus resulting in small criticality risks

  5. Nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Natori, Hisahide; Kurihara, Kunitoshi.

    1982-01-01

    Purpose: To increase the fuel safety by decreasing the gap conductance between fuels and cladding tubes, as well as improve the reactor core controllability by rendering the void coefficient negative. Constitution: Fuel assemblies in a pressure tube comprise a tie-rod, fuel rods in a central region, and fuel rods with burnable poison in the outer circumference region. Here, B 4 C is used as the burnable poison by 1.17 % by weight ratio. The degrees of enrichment for the fissile plutonium as PuO 2 -UO 2 fuel used in the assemblies are 2.7 %, 2.7 % and 1.5 % respectively in the innermost layer, the intermediate layer and the outermost layer. This increases the burn-up degree to improve the plant utilizability, whereby the void coefficient is rendered negative to improve the reactor core controllability. (Horiuchi, T.)

  6. Compound light ion fuel cycles: An approach to optimization

    International Nuclear Information System (INIS)

    Kernbichler, W.; Heindler, M.

    1985-01-01

    Together with the relatively high complexity and the low power density anticipated for fusion reactors have produced different attitude towards the long term perspective of fusion as a commercial energy source. The favourite pathway is to trust in optimization aiming at low tritium inventory, the availability of low-activation structure materials, the increase of redundancy, etc. In contrast, a respectable minority suggests turning away from d-t fusion or to envisage fusion as powerful neutron rather than energy source (fusion as fissile fuel or synfuel factory). We here intend to investigate the potentiality of fusion based on alternatives to d-t fuel. Such so called ''advanced fuels'' require higher burn temperatures and advanced reactor concepts (high-beta confinement schemes to compensate for their inherently lower reactivities. The experience that has been gained in fusion oriented plasma research admittedly justifies optimism for advanced fuels to a still lesser extent than for d-t. It can however be argued that it may pay off to choose a developmental direction with higher risk for failure but aiming at a more desirable end product. In order to explore this eventual desirability of advanced fuel fusion, we assume, as has been done in the case of d-t, that the first category of problems can be successfully handled. Our goal is thus to examine the potentiality of advanced fuels with respect to the second category of problems which largely determines the attractivity of utilization in fusion reactors

  7. Current state of spent fuel management in the Russian Federation

    International Nuclear Information System (INIS)

    Makarchuk, T.F.; Spichev, V.V.; Tikhonov, N.S.; Simanovsky, V.M.; Tokarenko, A.I.; Bespalov, V.N.

    1998-01-01

    Twenty nine power units of nine nuclear power plants of total installed capacity 22 GW(e) are now in operation in the Russian Federation. They produce approximately 12% of electric power in the country. The annual spent fuel arising is about 790 tU. The spent fuel from VVER-440 and BN-600 is reprocessed at the RT-1 plant near Chelyabinsk. The VVER-1000 spent fuel is planned to be reprocessed at the reprocessing plant RT-2 which is under construction near Krasnoyarsk. The RBMK-1000 spent fuel is not reprocessed because of its low fissile content. It is meant to be stored in intermediate storage facilities at the NPP sites and in a centralized storage facility during a period not less than 50 years and then to be disposed of in geological formations. State of the art of spent fuel reprocessing, storage and transportation is considered in the paper. Problems of nuclear fuel cycle back-end in Russia are taken into account. (author)

  8. Effects of fuel particle size distributions on neutron transport in stochastic media

    International Nuclear Information System (INIS)

    Liang, Chao; Pavlou, Andrew T.; Ji, Wei

    2014-01-01

    Highlights: • Effects of fuel particle size distributions on neutron transport are evaluated. • Neutron channeling is identified as the fundamental reason for the effects. • The effects are noticeable at low packing and low optical thickness systems. • Unit cells of realistic reactor designs are studied for different size particles. • Fuel particle size distribution effects are not negligible in realistic designs. - Abstract: This paper presents a study of the fuel particle size distribution effects on neutron transport in three-dimensional stochastic media. Particle fuel is used in gas-cooled nuclear reactor designs and innovative light water reactor designs loaded with accident tolerant fuel. Due to the design requirements and fuel fabrication limits, the size of fuel particles may not be perfectly constant but instead follows a certain distribution. This brings a fundamental question to the radiation transport computation community: how does the fuel particle size distribution affect the neutron transport in particle fuel systems? To answer this question, size distribution effects and their physical interpretations are investigated by performing a series of neutron transport simulations at different fuel particle size distributions. An eigenvalue problem is simulated in a cylindrical container consisting of fissile fuel particles with five different size distributions: constant, uniform, power, exponential and Gaussian. A total of 15 parametric cases are constructed by altering the fissile particle volume packing fraction and its optical thickness, but keeping the mean chord length of the spherical fuel particle the same at different size distributions. The tallied effective multiplication factor (k eff ) and the spatial distribution of fission power density along axial and radial directions are compared between different size distributions. At low packing fraction and low optical thickness, the size distribution shows a noticeable effect on neutron

  9. Fissile material detection and control facility with pulsed neutron sources and digital data processing

    International Nuclear Information System (INIS)

    Romodanov, V.L.; Chernikova, D.N.; Afanasiev, V.V.

    2010-01-01

    Full text: In connection with possible nuclear terrorism, there is long-felt need of devices for effective control of radioactive and fissile materials in the key points of crossing the state borders (airports, seaports, etc.), as well as various customs check-points. In International Science and Technology Center Projects No. 596 and No. 2978, a new physical method and digital technology have been developed for the detection of fissile and radioactive materials in models of customs facilities with a graphite moderator, pulsed neutron source and digital processing of responses from scintillation PSD detectors. Detectability of fissile materials, even those shielded with various radiation-absorbing screens, has been shown. The use of digital processing of scintillation signals in this facility is a necessary element, as neutrons and photons are discriminated in the time dependence of fissile materials responses at such loads on the electronic channels that standard types of spectrometers are inapplicable. Digital processing of neutron and photon responses practically resolves the problem of dead time and allows implementing devices, in which various energy groups of neutrons exist for some time after a pulse of source neutrons. Thus, it is possible to detect fissile materials deliberately concealed with shields having a large cross-section of absorption of photons and thermal neutrons. Two models of detection and the control of fissile materials were advanced: 1. the model based on graphite neutrons moderator and PSD scintillators with digital technology of neutrons and photons responses separation; 2. the model based on plastic scintillators and detecting of time coincidences of fission particles by digital technology. Facilities that count time coincidences of neutrons and photons occurring in the fission of fissile materials can use an Am Li source of neutrons, e.g. that is the case with the AWCC system. The disadvantages of the facility are related to the issues

  10. Fuel conditioning facility material accountancy

    International Nuclear Information System (INIS)

    Yacout, A.M.; Bucher, R.G.; Orechwa, Y.

    1995-01-01

    The operation of the Fuel conditioning Facility (FCF) is based on the electrometallurgical processing of spent metallic reactor fuel. It differs significantly, therefore, from traditional PUREX process facilities in both processing technology and safeguards implications. For example, the fissile material is processed in FCF only in batches and is transferred within the facility only as solid, well-characterized items; there are no liquid steams containing fissile material within the facility, nor entering or leaving the facility. The analysis of a single batch lends itself also to an analytical relationship between the safeguards criteria, such as alarm limit, detection probability, and maximum significant amount of fissile material, and the accounting system's performance, as it is reflected in the variance associated with the estimate of the inventory difference. This relation, together with the sensitivity of the inventory difference to the uncertainties in the measurements, allows a thorough evaluation of the power of the accounting system. The system for the accountancy of the fissile material in the FCF has two main components: a system to gather and store information during the operation of the facility, and a system to interpret this information with regard to meeting safeguards criteria. These are described and the precision of the inventory closure over one batch evaluated

  11. Standard problem exercise to validate criticality codes for large arrays of packages of fissile materials

    International Nuclear Information System (INIS)

    Whitesides, G.E.; Stephens, M.E.

    1986-01-01

    A study has been conducted by an Office of Economic Cooperation and Development-Committee on the Safety of Nuclear Installations (OECD-CSNI) Working Group that examined computational methods used to compute k/sub eff/ for large greater than or equal to5 3 arrays of fissile material (in which each unit is a substantial fraction of a critical mass). Five fissile materials that might typically be transported were used in the study. The ''packages'' used for this exercise were simplified to allow studies unperturbed by the variety of structural materials which would exist in an actual package. The only material present other than the fissile material was a variation in the moderator (water) surrounding the fissile material. Consistent results were obtained from calculations using several computational methods. That is, when the bias demonstrated by each method for actual critical experiments was used to ''correct'' the results obtained for systems for which there were no experimental data, there was good agreement between the methods. Two major areas of concern were raised by this exercise. First, the lack of experimental data for arrays with size greater than 5 3 limits validation for large systems. Second, there is a distinct possibility that the comingling of two shipments of unlike units could result in a reduction of the safety margins. Additional experiments and calculations will be required to satisfactorily resolve the remaining questions regarding the safe transport of large arrays of fissile materials

  12. High-power, photofission-inducing bremsstrahlung source for intense pulsed active detection of fissile material

    Directory of Open Access Journals (Sweden)

    J. C. Zier

    2014-06-01

    Full Text Available Intense pulsed active detection (IPAD is a promising technique for detecting fissile material to prevent the proliferation of special nuclear materials. With IPAD, fissions are induced in a brief, intense radiation burst and the resulting gamma ray or neutron signals are acquired during a short period of elevated signal-to-noise ratio. The 8 MV, 200 kA Mercury pulsed-power generator at the Naval Research Laboratory coupled to a high-power vacuum diode produces an intense 30 ns bremsstrahlung beam to study this approach. The work presented here reports on Mercury experiments designed to maximize the photofission yield in a depleted-uranium (DU object in the bremsstrahlung far field by varying the anode-cathode (AK diode gap spacing and by adding an inner-diameter-reducing insert in the outer conductor wall. An extensive suite of diagnostics was fielded to measure the bremsstrahlung beam and DU fission yield as functions of diode geometry. Delayed fission neutrons from the DU proved to be a valuable diagnostic for measuring bremsstrahlung photons above 5 MeV. The measurements are in broad agreement with particle-in-cell and Monte Carlo simulations of electron dynamics and radiation transport. These show that with increasing AK gap, electron losses to the insert and outer conductor wall increase and that the electron angles impacting the bremsstrahlung converter approach normal incidence. The diode conditions for maximum fission yield occur when the gap is large enough to produce electron angles close to normal, yet small enough to limit electron losses.

  13. High-power, photofission-inducing bremsstrahlung source for intense pulsed active detection of fissile material

    Science.gov (United States)

    Zier, J. C.; Mosher, D.; Allen, R. J.; Commisso, R. J.; Cooperstein, G.; Hinshelwood, D. D.; Jackson, S. L.; Murphy, D. P.; Ottinger, P. F.; Richardson, A. S.; Schumer, J. W.; Swanekamp, S. B.; Weber, B. V.

    2014-06-01

    Intense pulsed active detection (IPAD) is a promising technique for detecting fissile material to prevent the proliferation of special nuclear materials. With IPAD, fissions are induced in a brief, intense radiation burst and the resulting gamma ray or neutron signals are acquired during a short period of elevated signal-to-noise ratio. The 8 MV, 200 kA Mercury pulsed-power generator at the Naval Research Laboratory coupled to a high-power vacuum diode produces an intense 30 ns bremsstrahlung beam to study this approach. The work presented here reports on Mercury experiments designed to maximize the photofission yield in a depleted-uranium (DU) object in the bremsstrahlung far field by varying the anode-cathode (AK) diode gap spacing and by adding an inner-diameter-reducing insert in the outer conductor wall. An extensive suite of diagnostics was fielded to measure the bremsstrahlung beam and DU fission yield as functions of diode geometry. Delayed fission neutrons from the DU proved to be a valuable diagnostic for measuring bremsstrahlung photons above 5 MeV. The measurements are in broad agreement with particle-in-cell and Monte Carlo simulations of electron dynamics and radiation transport. These show that with increasing AK gap, electron losses to the insert and outer conductor wall increase and that the electron angles impacting the bremsstrahlung converter approach normal incidence. The diode conditions for maximum fission yield occur when the gap is large enough to produce electron angles close to normal, yet small enough to limit electron losses.

  14. Refuse derived fuel (RDF) plasma torch gasification as a feasible route to produce low environmental impact syngas for the cement industry.

    Science.gov (United States)

    López-Sabirón, Ana M; Fleiger, Kristina; Schäfer, Stefan; Antoñanzas, Javier; Irazustabarrena, Ane; Aranda-Usón, Alfonso; Ferreira, Germán A

    2015-08-01

    Plasma torch gasification (PTG) is currently researched as a technology for solid waste recovery. However, scientific studies based on evaluating its environmental implications considering the life cycle assessment (LCA) methodology are lacking. Therefore, this work is focused on comparing the environmental effect of the emissions of syngas combustion produced by refuse derived fuel (RDF) and PTG as alternative fuels, with that related to fossil fuel combustion in the cement industry. To obtain real data, a semi-industrial scale pilot plant was used to perform experimental trials on RDF-PTG.The results highlight that PTG for waste to energy recovery in the cement industry is environmentally feasible considering its current state of development. A reduction in every impact category was found when a total or partial substitution of alternative fuel for conventional fuel in the calciner firing (60 % of total thermal energy input) was performed. Furthermore, the results revealed that electrical energy consumption in PTG is also an important parameter from the LCA approach. © The Author(s) 2015.

  15. 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

    fuel, Anticipated evolution of fuel in dry storage, Anticipated evolution of fuel in deep geological disposal); Boiling-water reactor fuel (Similarities, and differences with PWR fuel, Axial and radial zoning, Rod and channel box sizes, Poisoning and reactivity control, Cladding specific characteristics, Trends in fuel evolution); 3 - Liquid-metal-cooled fast reactor fuel: Fast-neutron irradiation damage in structural materials (Fast-neutron-induced damage in metals, What materials should be used?); Fuels and targets for fast-reactor transmutation (Fast reactors: reactors affording the ability to carry out effective actinide transmutation, Recycling: homogeneous, or heterogeneous?); 4 - gas-cooled reactor fuel: Particle fuel (From the initial concept to the advanced TRISO particle concept, Kernel fabrication processes, Particle coating by chemical vapor deposition, Fuel element fabrication: particle compaction, Characterization of fuel particles, and elements, From HTR fuel to VHTR and GFR fuels: the GAIA facility at CEA/Cadarache); Irradiation behavior of particle fuels (Particle fuel: a variety of failure modes for a high-strength object, The amoeba effect, Fission product behavior, and diffusion in particle fuels); Mechanical modeling of particle fuel; Very-high-temperature reactor (VHTR) fuel; Gas-cooled fast reactor (GFR) fuel (The specifications for GFR fuel, GFR fissile material, First containment baffler materials, GFR fuel element concepts); 5 - Research reactor fuels (A considerable feedback from experience, Conversion of French reactors to low-enriched (≤20% U-235)U 3 Si 2 fuel, Conversion of all reactors: R and D requirements for high-performance reactors, An 'advanced' research reactor fuel: UMo, The startup fuel for the Jules Horowitz Reactor (JHR) will still be U 3 Si 2 -Al; 6 - An instrument for future fuel research: the Jules Horowitz Reactor (JHR): Fuel irradiation experiments in JHR, JHR: a flexible instrument; 7 - Glossary-Index

  16. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    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

    2009-07-01

    nature of spent nuclear fuel, Anticipated evolution of fuel in dry storage, Anticipated evolution of fuel in deep geological disposal); Boiling-water reactor fuel (Similarities, and differences with PWR fuel, Axial and radial zoning, Rod and channel box sizes, Poisoning and reactivity control, Cladding specific characteristics, Trends in fuel evolution); 3 - Liquid-metal-cooled fast reactor fuel: Fast-neutron irradiation damage in structural materials (Fast-neutron-induced damage in metals, What materials should be used?); Fuels and targets for fast-reactor transmutation (Fast reactors: reactors affording the ability to carry out effective actinide transmutation, Recycling: homogeneous, or heterogeneous?); 4 - gas-cooled reactor fuel: Particle fuel (From the initial concept to the advanced TRISO particle concept, Kernel fabrication processes, Particle coating by chemical vapor deposition, Fuel element fabrication: particle compaction, Characterization of fuel particles, and elements, From HTR fuel to VHTR and GFR fuels: the GAIA facility at CEA/Cadarache); Irradiation behavior of particle fuels (Particle fuel: a variety of failure modes for a high-strength object, The amoeba effect, Fission product behavior, and diffusion in particle fuels); Mechanical modeling of particle fuel; Very-high-temperature reactor (VHTR) fuel; Gas-cooled fast reactor (GFR) fuel (The specifications for GFR fuel, GFR fissile material, First containment baffler materials, GFR fuel element concepts); 5 - Research reactor fuels (A considerable feedback from experience, Conversion of French reactors to low-enriched ({<=}20% U-235)U{sub 3}Si{sub 2} fuel, Conversion of all reactors: R and D requirements for high-performance reactors, An 'advanced' research reactor fuel: UMo, The startup fuel for the Jules Horowitz Reactor (JHR) will still be U{sub 3}Si{sub 2}-Al; 6 - An instrument for future fuel research: the Jules Horowitz Reactor (JHR): Fuel irradiation experiments in JHR, JHR: a flexible

  17. Thorium fuel cycle analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, K [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

    1980-07-01

    Systems analysis of the thorium cycle, a nuclear fuel cycle accomplished by using thorium, is reported in this paper. Following a brief review on the history of the thorium cycle development, analysis is made on the three functions of the thorium cycle; (1) auxiliary system of U-Pu cycle to save uranium consumption, (2) thermal breeder system to exert full capacity of the thorium resource, (3) symbiotic system to utilize special features of /sup 233/U and neutron sources. The effects of the thorium loading in LWR (Light Water Reactor), HWR (Heavy Water Reactor) and HTGR (High Temperature Gas-cooled Reactor) are considered for the function of auxiliary system of U-Pu cycle. Analysis is made to find how much uranium is saved by /sup 233/U recycling and how the decrease in Pu production influences the introduction of FBR (Fast Breeder Reactor). Study on thermal breeder system is carried out in the case of MSBR (Molten Salt Breeder Reactor). Under a certain amount of fissile material supply, the potential system expansion rate of MSBR, which is determined by fissile material balance, is superior to that of FBR because of the smaller specific fissile inventory of MSBR. For symbiotic system, three cases are treated; i) nuclear heat supply system using HTGR, ii) denatured fuel supply system for nonproliferation purpose, and iii) hybrid system utilizing neutron sources other than fission reactor.

  18. Shippingport LWBR (Th/U Oxide) Fuel Characteristics for Disposal Criticality Analysis

    International Nuclear Information System (INIS)

    Taylor, L. L.; Loo, H. H.

    1999-01-01

    Department of Energy (DOE)-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into eight characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. The Shippingport Light Water Breeder Reactor (LWBR) fuels incorporate more of the conventional materials (zirconium cladding/heavy metal oxides) and fabrication details (rods and spacers) that make them comparable to a typical commercial fuel assembly. The LWBR seed/blanket configuration tested a light-water breeder concept with Th-232/U-233 binary fuel matrix. Reactor design used several assembly configurations at different locations within the same core . The seed assemblies contain the greatest fissile mass per (displaced) unit volume, but the blanket assemblies actually contain more fissile mass in a larger volume; the atom-densities are comparable

  19. Determination of reactor fuel burnup using passive neutron assay

    International Nuclear Information System (INIS)

    Kodeli, I.; Trkov, A.; Najzer, M.; Ertek, C.

    1988-01-01

    Passive neutron assay (PNA) method was developed to verify the fissile inventory of the irradiated reactor fuels. The characteristics of the method were studied at 'Jozef Stefan' Institute. The dependence of neutron source in the fuel on burnup, cooling time, initial enrichment and specific power were investigated and the accuracy of the method, using available computer codes was estimated. (author)

  20. Far-Field Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    International Nuclear Information System (INIS)

    J.P. Nicot

    2000-01-01

    The objective of this calculation is to estimate the quantity of fissile material that could accumulate in fractures in the rock beneath plutonium-ceramic (Pu-ceramic) and Mixed-Oxide (MOX) waste packages (WPs) as they degrade in the potential monitored geologic repository at Yucca Mountain. This calculation is to feed another calculation (Ref. 31) computing the probability of criticality in the systems described in Section 6 and then ultimately to a more general report on the impact of plutonium on the performance of the proposed repository (Ref. 32), both developed concurrently to this work. This calculation is done in accordance with the development plan TDP-DDC-MD-000001 (Ref. 9), item 5. The original document described in item 5 has been split into two documents: this calculation and Ref. 4. The scope of the calculation is limited to only very low flow rates because they lead to the most conservative cases for Pu accumulation and more generally are consistent with the way the effluent from the WP (called source term in this calculation) was calculated (Ref. 4). Ref. 4 (''In-Drift Accumulation of Fissile Material from WPs Containing Plutonium Disposition Waste Forms'') details the evolution through time (breach time is initial time) of the chemical composition of the solution inside the WP as degradation of the fuel and other materials proceed. It is the chemical solution used as a source term in this calculation. Ref. 4 takes that same source term and reacts it with the invert; this calculation reacts it with the rock. In addition to reactions with the rock minerals (that release Si and Ca), the basic mechanisms for actinide precipitation are dilution and mixing with resident water as explained in Section 2.1.4. No other potential mechanism such as flow through a reducing zone is investigated in this calculation. No attempt was made to use the effluent water from the bottom of the invert instead of using directly the effluent water from the WP. This

  1. Far-Field Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    J.P. Nicot

    2000-09-29

    The objective of this calculation is to estimate the quantity of fissile material that could accumulate in fractures in the rock beneath plutonium-ceramic (Pu-ceramic) and Mixed-Oxide (MOX) waste packages (WPs) as they degrade in the potential monitored geologic repository at Yucca Mountain. This calculation is to feed another calculation (Ref. 31) computing the probability of criticality in the systems described in Section 6 and then ultimately to a more general report on the impact of plutonium on the performance of the proposed repository (Ref. 32), both developed concurrently to this work. This calculation is done in accordance with the development plan TDP-DDC-MD-000001 (Ref. 9), item 5. The original document described in item 5 has been split into two documents: this calculation and Ref. 4. The scope of the calculation is limited to only very low flow rates because they lead to the most conservative cases for Pu accumulation and more generally are consistent with the way the effluent from the WP (called source term in this calculation) was calculated (Ref. 4). Ref. 4 (''In-Drift Accumulation of Fissile Material from WPs Containing Plutonium Disposition Waste Forms'') details the evolution through time (breach time is initial time) of the chemical composition of the solution inside the WP as degradation of the fuel and other materials proceed. It is the chemical solution used as a source term in this calculation. Ref. 4 takes that same source term and reacts it with the invert; this calculation reacts it with the rock. In addition to reactions with the rock minerals (that release Si and Ca), the basic mechanisms for actinide precipitation are dilution and mixing with resident water as explained in Section 2.1.4. No other potential mechanism such as flow through a reducing zone is investigated in this calculation. No attempt was made to use the effluent water from the bottom of the invert instead of using directly the effluent water from the

  2. The role of congress in future disposal of fissile materials from dismantled nuclear weapons

    International Nuclear Information System (INIS)

    Donnelly, W.H.; Davis, Z.S.

    1991-01-01

    Assuming the Soviet Union remains intact as a major power and the superpowers do not retrogress to a new Cold War era, it is likely that the United States and the Soviet Union will eventually agree to deep cuts in their nuclear arsenals. Future arms control agreements may be coupled with companion agreements to stop production of fissile materials for nuclear weapons, to dismantle the warheads of the nuclear weapons, and to dispose of their fissile materials to prevent reuse in new warheads. Such agreements would be negotiated by the U.S. executive branch but probably would require ratification, funding, and enabling legislation from the U.S. Congress if they are to succeed. There follows a brief review of the ideas for disposal of fissile materials from dismantled nuclear warheads and the potential role and influence of the Congress in the negotiation, ratification, and implementation of U.S.-Soviet agreements for such disposal

  3. Requirements for timber and cadmium used in shielding for fissile material transport packaging

    International Nuclear Information System (INIS)

    1982-02-01

    This Code of Practice has been prepared as a guide for designers who require packaging for fissile materials. It should be noted that this document covers design requirements only and it is not a manufacturing specification which can be quoted on a manufacturing contract without qualification. Compliance with the regulations regarding the safe transport of fissile materials may be achieved by the provision of an effective shield embodying:- (a) a moderating material -usually one rich in hydrogen, such as wood - in order to thermalise incoming neutrons, and (b) a material - such as cadmium - with a large absorption cross-section for thermal neutrons, located between the moderator and the fissile material, in order to capture the incoming neutrons. This Code describes the requirements in two sections, one for each of these materials. (author)

  4. An approximate method to estimate the minimum critical mass of fissile nuclides

    International Nuclear Information System (INIS)

    Wright, R.Q.; Jordan, W.C.

    1999-01-01

    When evaluating systems in criticality safety, it is important to approximate the answer before any analysis is performed. There is currently interest in establishing the minimum critical parameters for fissile actinides. The purpose is to describe the OB-1 method for estimating the minimum critical mass for thermal systems based on one-group calculations and 235 U spheres fully reflected by water. The observation is made that for water-moderated, well-thermalized systems, the transport and leakage from the system are dominated by water. Under these conditions two fissile mixtures will have nearly the same critical volume provided the infinite media multiplication factor (k ∞ ) for the two systems is the same. This observation allows for very simple estimates of critical concentration and mass as a function of the hydrogen-to-fissile (H/X) moderation ratio by comparison to the known 235 U system

  5. Verification arrangements for the proposed fissile material cut-off treaty

    International Nuclear Information System (INIS)

    Bragin, V.

    2001-01-01

    Since the mid-1950's, an agreement to terminate the production of fissile material for nuclear weapons has been on the agenda. On December 16, 1993, the UNGA adopted Resolution A/RES/48/75/L which recommends ''the negotiation in the most appropriate international forum of a non-discriminatory, multilateral and internationally and effectively verifiable treaty banning the production of fissile material for nuclear weapons and other nuclear explosive devices''. The proposed Fissile Material Cut-off Treaty (FMCT) is still one of the most important items on the multilateral disarmament and non-proliferation agenda. Successful achievement of the FMCT would be an important step towards the goal of eliminating nuclear weapons. (author)

  6. Neutronic studies of the long life core concept: Part 1, Design and performance of 1000 MWe uranium oxide fueled low power density LMR cores

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1987-04-01

    The parametric behavior of some key neutronic performance parameters for low power density LMR cores fueled with uranium oxide is investigated. The results are compared to reference homogeneous and heterogeneous cores with normal fuel management and Pu fueling. It can be concluded that with respect to minimizing the initial fissile mass and thereby economizing on the inventory costs and carrying charges, the superior neutron economy of the LMR fuel cycle is best exploited through normal fuel management with Pu recycling. In the once-through mode the LMR fuel cycle has disadvantages due to a higher fissile inventory and is not competitive with the LWR fuel cycle

  7. Development method for measuring thickness of nuclei and coating of fuel plates

    International Nuclear Information System (INIS)

    Borges Junior, Reinaldo

    2013-01-01

    One of the most important components of a nuclear reactor is the Nuclear Fuel. Currently, the most advanced commercial fuel, whose applicability in Brazilian reactors has been developed by IPEN since 1985, is the silicide U 3 Si 2 . This is formed by fuel plates with nuclei dispersion (where the fissile material (U 3 Si 2 ) is homogeneously dispersed in a matrix of aluminum) coated aluminum. This fuel is produced in Brazil with developed technology, the result of the efforts made by the group of manufacturing nuclear fuel (CCN - Center of Nuclear Fuel) of IPEN. Considering the necessity of increasing the power of the IEA- R1 and Brazilian Multipurpose Reactor Building (RMB), for the production of radioisotopes - mainly for the area of medicine - there will be significant increase in the production of nuclear fuel at IPEN. Given this situation, if necessary, make the development of more modern and automated classification techniques. Aiming at this goal, this work developed a new computational method for measuring thickness of core and cladding of fuel plates, which are able to perform such measurements in less time and with more meaningful statistical data when compared with the current method of measurement. (author)

  8. A novel concept of QUADRISO particles Part III: applications to the plutonium-thorium fuel cycle

    International Nuclear Information System (INIS)

    Talamo, A.

    2009-01-01

    In the present study, a plutonium-thorium fuel cycle is investigated including the 233 U production and utilization. A prismatic thermal High Temperature Gas Reactor (HTGR) and the novel concept of quadruple isotropic (QUADRISO) coated particles, designed at the Argonne National Laboratory, have been used for the study. In absorbing QUADRISO particles, a burnable poison layer surrounds the central fuel kernel to flatten the reactivity curve as a function of time. At the beginning of life, the fuel in the QUADRISO particles is hidden from neutrons, since they get absorbed in the burnable poison before they reach the fuel kernel. Only when the burnable poison depletes, neutrons start streaming into the fuel kernel inducing fission reactions and compensating the fuel depletion of ordinary TRISO particles. In fertile QUADRISO particles, the absorber layer is replaced by natural thorium with the purpose of flattening the excess of reactivity by the thorium resonances and producing 233 U. The above configuration has been compared with a configuration where fissile (neptunium-plutonium oxide from Light Water Reactors irradiated fuel) and fertile (natural thorium oxide) fuels are homogeneously mixed in the kernel of ordinary TRISO particles. For the 233 U utilization, the core has been equipped with europium oxide absorbing QUADRISO particles.

  9. Viability of inert matrix fuel in reducing plutonium amounts in reactors

    International Nuclear Information System (INIS)

    2006-08-01

    Reactors worldwide have produced more than 2000 tonnes of plutonium, contained in spent fuel or as separated forms through reprocessing. Disposition of fissile materials has become a primary concern of nuclear non-proliferation efforts. There is a significant interest in IAEA Member States to develop proliferation resistant nuclear fuel cycles for incineration of plutonium such as inert matrix fuels (IMFs). The present report summarises R and D work on inert matrix fuel for plutonium and (to a lesser extent) minor actinide stock-pile reduction, and discusses the possible strategies to include inert matrix fuel approaches to the nuclear fuel cycle. The publication reviews the status of potential IMF candidates and describes several identified candidate materials for both fast and thermal reactors: MgO, ZrO2, SiC, Zr alloy, SiAl, ZrN; some of these have undergone test irradiations and post-irradiation examination. Also discussed are modelling of IMF fuel performance and safety analysis. System studies have identified strategies for both implementation of IMF fuel as homogeneous or heterogeneous phases, as assemblies or core loadings and in existing reactors in the shorter term, as well as in new reactors in the longer term

  10. Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vapirev, E; Jordanov, T; Khristoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1996-12-31

    The possibility of using highly enriched uranium available from military inventories for production of mixed oxide fuel (MOX) has been proposed. The fuel is based on U-235 dioxide as fissile isotope and Th-232 dioxide as a non-fissile isotope. It is shown that although the fuel conversion coefficient to U-233 is expected to be less than 1, the proposed fuel has several important advantages resulting in cost reduction of the nuclear fuel cycle. The expected properties of MOX fuel (cross-sections, generated chains, delayed neutrons) are estimated. Due to fuel generation the initial enrichment is expected to be 1% less for production of the same energy. In contrast to traditional fuel no long living actinides are generated which reduces the disposal and reprocessing cost. 7 refs.

  11. 75 FR 59622 - Supplemental Determination for Renewable Fuels Produced Under the Final RFS2 Program From Canola Oil

    Science.gov (United States)

    2010-09-28

    ..., heating oil or jet fuel). In addition, this rule includes a new regulatory provision establishing a... work would be completed through a supplemental final rulemaking process. This supplemental final rule... the final RFS2 rule, EPA will revisit our lifecycle analyses in the future as new information becomes...

  12. Analysis of fuel management in the KIPT neutron source facility

    Energy Technology Data Exchange (ETDEWEB)

    Zhong Zhaopeng, E-mail: zzhong@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Gohar, Yousry; Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2011-05-15

    Research highlights: > Fuel management of KIPT ADS was analyzed. > Core arrangement was shuffled in stage wise. > New fuel assemblies was added into core periodically. > Beryllium reflector could also be utilized to increase the fuel life. - Abstract: Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility consisting of an electron accelerator driven sub-critical assembly. The neutron source driving the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The sub-critical assembly surrounding the target is fueled with low enriched WWR-M2 type hexagonal fuel assemblies. The U-235 enrichment of the fuel material is <20%. The facility will be utilized for basic and applied research, producing medical isotopes, and training young specialists. With the 100 KW electron beam power, the total thermal power of the facility is {approx}360 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products continuously reduce the system reactivity during the operation, decrease the neutron flux level, and consequently impact the facility performance. To preserve the neutron flux level during the operation, the fuel assemblies should be added and shuffled for compensating the lost reactivity caused by burnup. Beryllium reflector could also be utilized to increase the fuel life time in the sub-critical core. This paper studies the fuel cycles and shuffling schemes of the fuel assemblies of the sub-critical assembly to preserve the system reactivity and the neutron flux level during the operation.

  13. Analysis of fuel management in the KIPT neutron source facility

    International Nuclear Information System (INIS)

    Zhong Zhaopeng; Gohar, Yousry; Talamo, Alberto

    2011-01-01

    Research highlights: → Fuel management of KIPT ADS was analyzed. → Core arrangement was shuffled in stage wise. → New fuel assemblies was added into core periodically. → Beryllium reflector could also be utilized to increase the fuel life. - Abstract: Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility consisting of an electron accelerator driven sub-critical assembly. The neutron source driving the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The sub-critical assembly surrounding the target is fueled with low enriched WWR-M2 type hexagonal fuel assemblies. The U-235 enrichment of the fuel material is <20%. The facility will be utilized for basic and applied research, producing medical isotopes, and training young specialists. With the 100 KW electron beam power, the total thermal power of the facility is ∼360 kW including the fission power of ∼260 kW. The burnup of the fissile materials and the buildup of fission products continuously reduce the system reactivity during the operation, decrease the neutron flux level, and consequently impact the facility performance. To preserve the neutron flux level during the operation, the fuel assemblies should be added and shuffled for compensating the lost reactivity caused by burnup. Beryllium reflector could also be utilized to increase the fuel life time in the sub-critical core. This paper studies the fuel cycles and shuffling schemes of the fuel assemblies of the sub-critical assembly to preserve the system reactivity and the neutron flux level during the operation.

  14. Criticality Safety in the Handling of Fissile Material. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-05-15

    This Safety Guide provides guidance and recommendations on how to meet the relevant requirements for ensuring subcriticality when dealing with fissile material and for planning the response to criticality accidents. The guidance and recommendations are applicable to both regulatory bodies and operating organizations. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences of this if it were to occur. The Safety Guide makes recommendations on how to ensure subcriticality in systems involving fissile materials during normal operation, anticipated operational occurrences, and, in the case of accident conditions, within design basis accidents, from initial design through commissioning, operation, and decommissioning and disposal.

  15. Fissility of actinide nuclei induced by 60-130 MeV photons

    International Nuclear Information System (INIS)

    Morcelle, Viviane; Tavares, Odilon A.P.

    2004-06-01

    Nuclear fissilities obtained from recent photofission reaction cross section measurements carried out at Saskatchewan Accelerator Laboratory (Saskatoon, Canada) in the energy range 60-130 MeV for 232 Th, 233 U, 235 U, 238 U, and 237 Np nuclei have been analysed in a systematic way. To this aim, a semiempirical approach has been developed based on the quasi-deuteron nuclear photoabsorption model followed by the process of competition between neutron evaporation and fission for the excited nucleus. The study reproduces satisfactorily well the increasing trend of nuclear fissility with parameter Z 2 =A. (author)

  16. International conference on military conversion and science. Utilization/disposal of the excess fissile weapon materials: scientific, technological and socio-economic aspects

    International Nuclear Information System (INIS)

    Kouzminov, V.; Martellini, M.

    1996-01-01

    The Proceedings of the Conference includes the papers presented by the eminent specialists in the field of utilisation and/or disposal of excess fissile materials, each with a separate abstract, as well as the Conference opening and introduction speeches. According to the concerned subjects presentations were divided into following five sessions: perspectives of nuclear research and development; Technical problems and possibilities of civilian utilization of Highly enriched uranium (HEU) and plutonium including alternate strategies (application of MOX fuel) and operational and safety problems; Comparison of different options for weapon-grade Pu utilization connected to present programme for recycling of civilian Pu; Socio-economic aspects including cost of Pu conversion and fabrication of MOX fuel; Effects of different strategies of waste disposal including environmental and safety related issues

  17. Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

    Science.gov (United States)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-01-01

    Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

  18. Development of geological disposal system for spent fuels and high-level radioactive wastes in Korea

    International Nuclear Information System (INIS)

    Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won

    2013-01-01

    Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel) for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

  19. DEVELOPMENT OF GEOLOGICAL DISPOSAL SYSTEMS FOR SPENT FUELS AND HIGH-LEVEL RADIOACTIVE WASTES IN KOREA

    Directory of Open Access Journals (Sweden)

    HEUI-JOO CHOI

    2013-02-01

    Full Text Available Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

  20. MOX fuel reprocessing and recycling

    International Nuclear Information System (INIS)

    Guillet, J.L.

    1990-01-01

    This paper is devoted to the reprocessing of MOX fuel in UP2-800 plant at La Hague, and to the MOX successive reprocessing and recycling. 1. MOX fuel reprocessing. In a first step, the necessary modifications in UP2-800 to reprocess MOX fuel are set out. Early in the UP2-800 project, actions have been taken to reprocess MOX fuel without penalty. They consist in measures regarding: Dissolution; Radiological shieldings; Nuclear instrumentation; Criticality. 2. Mox successive reprocessing and recycling. The plutonium recycling in the LWR is now a reality and, as said before, the MOX fuel reprocessing is possible in UP2-800 plant at La Hague. The following actions in this field consist in verifying the MOX successive reprocessing and recycling possibilities. After irradiation, the fissile plutonium content of irradiated MOX fuel is decreased and, in this case, the re-use of plutonium in the LWR need an important increase of initial Pu enrichment inconsistent with the Safety reactor constraints. Cogema opted for reprocessing irradiated MOX fuel in dilution with the standard UO2 fuel in appropriate proportions (1 MOX for 4 UO2 fuel for instance) in order to save a fissile plutonium content compatible with MOX successive recycling (at least 3 recyclings) in LWR. (author). 2 figs

  1. Current issues in the transport of radioactive waste and spent fuel: work by the World Nuclear Transport Institute

    Energy Technology Data Exchange (ETDEWEB)

    Neau, H-J.; Bonnardel-Azzarelli, B. [World Nuclear Transport Inst., London (United Kingdom)

    2014-07-01

    Various kinds of radioactive waste are generated from nuclear power and fuel cycle facilities. These materials have to be treated, stored and eventually sent to a repository site. Transport of wastes between these various stages is crucial for the sustainable utilization of nuclear energy. The IAEA Regulations for the Safe Transport of Radioactive Material (SSR-6) have, for many decades, provided a safe and efficient framework for radioactive materials transport and continue to do so. However, some shippers have experienced that in the transport of certain specific radioactive wastes, difficulties can be encountered. For example, some materials produced in the decommissioning of nuclear facilities are unique in terms of composition or size and can be difficult to characterize as surface contaminated objects (SCO) or homogeneous. One way WNTI (World Nuclear Transport Institute) helps develop transport methodologies is through the use of Industry Working Groups, bringing together WNTI members with common interests, issues and experiences. The Back-End Transport Industry Working Group focuses on the following issues currently. - Characterization of Waste: techniques and methods to classify wastes - Large Objects: slightly contaminated large objects (ex. spent steam generators) transport - Dual Use Casks: transportable storage casks for spent nuclear fuels, including the very long term storage of spent fuel - Fissile Exceptions: new fissile exceptions provisions of revised TS-R-1 (SSR-6) The paper gives a broad overview of current issues for the packaging and transport of radioactive wastes and the associated work of the WNTI. (author)

  2. Fuel element for high-temperature nuclear power reactors

    International Nuclear Information System (INIS)

    Schloesser, J.

    1974-01-01

    The fuel element of the HTGR consists of a spherical graphite body with a spherical cavity. A deposit of fissile material, e.g. coated particles of uranium carbide, is fixed to the inner wall using binders. In addition to the fissile material, there are concentric deposits of fertile material, e.g. coated thorium carbide particles. The remaining cavity is filled with a graphite mass, preferably graphite powder, and the filling opening with a graphite stopper. At the beginning of the reactor operation, the fissile material layer provides the whole power. With progressing burn-up, the energy production is taken over by the fertile layer, which provides the heat production until the end of burn-up. Due to the relatively small temperature difference between the outer wall of the outer graphite body and the maximum fuel temperature, the power of the fuel element can be increased. (DG) [de

  3. Disposal criticality analysis methodology for fissile waste forms

    International Nuclear Information System (INIS)

    Davis, J.W.; Gottlieb, P.

    1998-03-01

    A general methodology has been developed to evaluate the criticality potential of the wide range of waste forms planned for geologic disposal. The range of waste forms include commercial spent fuel, high level waste, DOE spent fuel (including highly enriched), MOX using weapons grade plutonium, and immobilized plutonium. The disposal of these waste forms will be in a container with sufficiently thick corrosion resistant barriers to prevent water penetration for up to 10,000 years. The criticality control for DOE spent fuel is primarily provided by neutron absorber material incorporated into the basket holding the individual assemblies. For the immobilized plutonium, the neutron absorber material is incorporated into the waste form itself. The disposal criticality analysis methodology includes the analysis of geochemical and physical processes that can breach the waste package and affect the waste forms within. The basic purpose of the methodology is to guide the criticality control features of the waste package design, and to demonstrate that the final design meets the criticality control licensing requirements. The methodology can also be extended to the analysis of criticality consequences (primarily increased radionuclide inventory), which will support the total performance assessment for the respository

  4. Nuclear Fuel Reprocessing

    International Nuclear Information System (INIS)

    Simpson, Michael F.; Law, Jack D.

    2010-01-01

    This is a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. Nuclear reprocessing is the chemical treatment of spent fuel involving separation of its various constituents. Principally, it is used to recover useful actinides from the spent fuel. Radioactive waste that cannot be re-used is separated into streams for consolidation into waste forms. The first known application of nuclear reprocessing was within the Manhattan Project to recover material for nuclear weapons. Currently, reprocessing has a peaceful application in the nuclear fuel cycle. A variety of chemical methods have been proposed and demonstrated for reprocessing of nuclear fuel. The two most widely investigated and implemented methods are generally referred to as aqueous reprocessing and pyroprocessing. Each of these technologies is described in detail in Section 3 with numerous references to published articles. Reprocessing of nuclear fuel as part of a fuel cycle can be used both to recover fissionable actinides and to stabilize radioactive fission products into durable waste forms. It can also be used as part of a breeder reactor fuel cycle that could result in a 14-fold or higher increase in energy utilization per unit of natural uranium. Reprocessing can also impact the need for geologic repositories for spent fuel. The volume of waste that needs to be sent to such a repository can be reduced by first subjecting the spent fuel to reprocessing. The extent to which volume reduction can occur is currently under study by the United States Department of Energy via research at various national laboratories and universities. Reprocessing can also separate fissile and non-fissile radioactive elements for transmutation.

  5. Vertical integration of local fuel producers into rural district heating systems – Climate impact and production costs

    International Nuclear Information System (INIS)

    Kimming, M.; Sundberg, C.; Nordberg, Å.; Hansson, P.-A.

    2015-01-01

    Farmers can use their own agricultural biomass residues for heat production in small-scale systems, enabling synergies between the district heating (DH) sector and agriculture. The barriers to entry into the Swedish heat market were extremely high as long as heat distribution were considered natural monopoly, but were recently lowered due to the introduction of a regulated third party access (TPA) system in the DH sector. This study assesses the potential impact on greenhouse gas emissions and cost-based heat price in the DH sector when farmers vertically integrate into the heat supply chain and introduce more local and agricultural crops and residues into the fuel mix. Four scenarios with various degree of farmer integration, were assessed using life cycle assessment (LCA) methodology, and by analysis of the heat production costs. The results show that full integration of local farm and forest owners in the value chain can reduce greenhouse gas emissions and lower production costs/heat price, if there is an incentive to utilise local and agricultural fuels. The results imply that farmer participation in the DH sector should be encouraged by e.g. EU rural development programmes. - Highlights: • Five DH production systems based on different fuels and ownership were analysed. • Lower GHG emissions were obtained when farmers integrate fully into the DH chain. • Lower heat price was obtained by full vertical integration of farmers. • Salix and straw-based production resulted in the lowest GHG and heat price

  6. Use of an oscillation technique to measure effective cross-sections of fissionable samples in critical assemblies; Mesure des sections efficaces effectives d'echantillons fissiles par une methode d'oscillation dans les-assemblages critiques

    Energy Technology Data Exchange (ETDEWEB)

    Tretiakoff, O; Vidal, R; Carre, J C; Robin, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The authors describe the technique used to measure the effective absorption and neutron-yield cross-sections of a fissionable sample. These two values are determined by analysing the signals due to the variation in reactivity (over-all signal) and the local perturbation in the flux (local signal) produced by the oscillating sample. These signals are standardized by means of a set of samples containing quantities of fissionable material ({sup 235}U) and an absorber, boron, which are well known. The measurements are made for different neutron spectra characterized by lattice parameters which constitute the central zone within which the sample moves. This technique is used to study the effective cross-sections of uranium-plutonium alloys for different heavy-water and graphite lattices in the MINERVE and MARIUS critical assemblies. The same experiments are carried out on fuel samples of different irradiations in order to determine the evolution of effective cross-sections as a function of the spectrum and the irradiations. (authors) [French] On decrit la methode utilisee pour mesurer les sections efficaces effectives d'absorption et de production de neutrons d'un echantillon fissile. Ces deux grandeurs sont determinees en analysant les signaux dus a la variation de reactivite (signal global) et a la perturbation locale de flux (signal local) produits par l'echantillon oscillant. Ces signaux sont etalonnes a l'aide d'un jeu d'echantillons dont les teneurs en materiau fissile ({sup 235}U) et en absorbeur (bore) sont bien connues. Les mesures sont realisees pour differents spectres de neutrons caracterises par les parametres du reseau constituant la zone centrale a l'interieur de laquelle se deplace l'echantillon. A l'aide de cette methode on etudie les sections efficaces effectives d'alliage uranium-plutonium pour differents reseaux a eau lourde et a graphite dans les assemblages crtiques MINERVE et MARIUS. Les memes experiences sont effectuees sur des echantillons de

  7. HTGR fuel reprocessing pilot plant: results of the sequential equipment operation

    International Nuclear Information System (INIS)

    Strand, J.B.; Fields, D.E.; Kergis, C.A.

    1979-05-01

    The second sequential operation of the HTGR fuel reprocessing cold-dry head-end pilot plant equipment has been successfully completed. Twenty standard LHGTR fuel elements were crushed to a size suitable for combustion in a fluid bed burner. The graphite was combusted leaving a product of fissile and fertile fuel particles. These particles were separated in a pneumatic classifier. The fissile particles were fractured and reburned in a fluid bed to remove the inner carbon coatings. The remaining products are ready for dissolution and solvent extraction fuel recovery

  8. Conversion ratio and consumption of fissile material in PWR reactors

    International Nuclear Information System (INIS)

    Tiba, C.

    1977-01-01

    It has been shown that the uranium resources will be insufficient for future projected demand. The many solutions to this problem are considered and, in particular, the effect of enrichment on the conversion ratio and hence total uranium comsumption is studied. The developed computacional method employs the one-group neutron diffusion theory. The model is verified by calculating typical burn-up, conversion ratio, U-235 comsumption and plutonium production values in PWR's, and comparing results with those in the published literature. The associated costs of U and U-Pu fuel cycles are also studied for various enrichment values [pt

  9. Transfer of fissile material through shielding coatings in emergency heating of HTGR coated particles

    International Nuclear Information System (INIS)

    Gudkov, A.N.; Zhuravkov, S.G.; Koptev, M.A.; Kurepin, A.D.

    1990-01-01

    The measurement results of leakage dynamics of fissile material from the coated particles within a temperature range of 1200 + 2000 deg. C are given. The methods of carrying out the experiments are briefly described. The relation of the leakage rate of uranium-235 from CP (coated particles) with the pyrocarbonic coatings has been obtained. (author)

  10. Nuclear energy - Fissile materials - Principles of criticality safety in storing, handling and processing

    International Nuclear Information System (INIS)

    1995-01-01

    This International Standard specifies the basic principles and limitations which govern operations with fissile materials. It discusses general criticality safety criteria for equipment design and for the development of operating controls, while providing guidance for the assessment of procedures, equipment, and operations. It does not cover quality assurance requirements or details of equipment or operational procedures, nor does it cover the effects of radiation on man or materials, or sources of such radiation, either natural or as the result of nuclear chain reactions. Transport of fissile materials outside the boundaries of nuclear establishments is not within the scope of this International Standard and should be governed by appropriate national and international standards and regulations. These criteria apply to operations with fissile materials outside nuclear reactors but within the boundaries of nuclear establishments. They are concerned with the limitations which must be imposed on operations because of the unique properties of these materials which permit them to support nuclear chain reactions. These principles apply to quantities of fissile materials in which nuclear criticality can be established

  11. Detector and front-end electronics of a fissile mass flow monitoring system

    International Nuclear Information System (INIS)

    Paulus, M.J.; Uckan, T.; Lenarduzzi, R.; Mullens, J.A.; Castleberry, K.N.; McMillan, D.E.; Mihalczo, J.T.

    1997-01-01

    A detector and front-end electronics unit with secure data transmission has been designed and implemented for a fissile mass flow monitoring system for fissile mass flow of gases and liquids in a pipe. The unit consists of 4 bismuth germanate (BGO) scintillation detectors, pulse-shaping and counting electronics, local temperature sensors, and on-board local area network nodes which locally acquire data and report to the master computer via a secure network link. The signal gain of the pulse-shaping circuitry and energy windows of the pulse-counting circuitry are periodicially self calibrated and self adjusted in situ using a characteristic line in the fissile material pulse height spectrum as a reference point to compensate for drift such as in the detector gain due to PM tube aging. The temperature- dependent signal amplitude variations due to the intrinsic temperature coefficients of the PM tube gain and BGO scintillation efficiency have been characterized and real-time gain corrections introduced. The detector and electronics design, measured intrinsic performance of the detectors and electronics, and the performance of the detector and electronics within the fissile mass flow monitoring system are described

  12. Recommended nuclear criticality safety experiments in support of the safe transportation of fissile material

    International Nuclear Information System (INIS)

    Tollefson, D.A.; Elliott, E.P.; Dyer, H.R.; Thompson, S.A.

    1993-01-01

    Validation of computer codes and nuclear data (cross-section) libraries using benchmark quality critical (or certain subcritical) experiments is an essential part of a nuclear criticality safety evaluation. The validation results establish the credibility of the calculational tools for use in evaluating a particular application. Validation of the calculational tools is addressed in several American National Standards Institute/American Nuclear Society (ANSI/ANS) standards, with ANSI/ANS-8.1 being the most relevant. Documentation of the validation is a required part of all safety analyses involving significant quantities of fissile materials. In the case of transportation of fissile materials, the safety analysis report for packaging (SARP) must contain a thorough discussion of benchmark experiments, detailing how the experiments relate to the significant packaging and contents materials (fissile, moderating, neutron absorbing) within the package. The experiments recommended in this paper are needed to address certain areas related to transportation of unirradiated fissile materials in drum-type containers (packagings) for which current data are inadequate or are lacking

  13. Detection of tiny amounts of fissile materials in large-sized containers with radioactive waste

    Directory of Open Access Journals (Sweden)

    Batyaev V.F.

    2018-01-01

    Full Text Available The paper is devoted to non-destructive control of tiny amounts of fissile materials in large-sized containers filled with radioactive waste (RAW. The aim of this work is to model an active neutron interrogation facility for detection of fissile ma-terials inside NZK type containers with RAW and determine the minimal detectable mass of U-235 as a function of various param-eters: matrix type, nonuniformity of container filling, neutron gen-erator parameters (flux, pulse frequency, pulse duration, meas-urement time. As a result the dependence of minimal detectable mass on fissile materials location inside container is shown. Nonu-niformity of the thermal neutron flux inside a container is the main reason of the space-heterogeneity of minimal detectable mass in-side a large-sized container. Our experiments with tiny amounts of uranium-235 (<1 g confirm the detection of fissile materials in NZK containers by using active neutron interrogation technique.

  14. 10 CFR 71.59 - Standards for arrays of fissile material packages.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Standards for arrays of fissile material packages. 71.59 Section 71.59 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE.... The value of the CSI may be zero provided that an unlimited number of packages are subcritical, such...

  15. Safety considerations in the fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Baker, A.R.; Burton, W.R.; Taylor, H.A.

    1977-01-01

    The fuel cycle safety problems for fast reactors, as compared with thermal reactors, are enhanced by the higher fissile content and heat rating of the fuel. Additionally recycling leads to the build up of substantial isotopes which contribute to the alpha and neutron hazards. The plutonium arisings in a nuclear power reactor programme extending into the next century are discussed. A requirement is to be able to return the product plutonium to a reactor about 9 months after the end of irradiation and it is anticipated that progress will be made slowly towards this fuel cycle, having regard to the necessity for maintaining safe and reliable operations. Consideration of the steps in the fuel cycle has indicated that it will be best to store the irradiated fuel on the reactor sites while I131 decays and decay heat falls before transporting and a suitable transport flask is being developed. Reprocessing development work is aimed at the key area of fuel breakdown, the inter-relation of the fuel characteristics on the dissolution of the plutonium and a solvent extract cycle leading to a product suitable for a co-located fabrication plant. Because of the high activity of recycled fuel it is considered that fabrication must move to a fully remote operation as is already the case for reprocessing, and a gel precipitation process producing a vibro compacted fuel is under development for this purpose. The waste streams from the processing plants must be minimised, processed for recovery of plutonium where applicable and then conditioned so that the final products released from the processing cycle are acceptable for ultimate disposal. The safety aspects reviewed cover protection of operators, containment of radioactive materials, criticality and regulation of discharges to the environment

  16. Interim design report: fuel particle crushing

    International Nuclear Information System (INIS)

    Baer, J.W.; Strand, J.B.; Cook, E.J.; Miller, C.M.

    1977-11-01

    The double-roll fuel particle crusher was developed to fracture the silicon carbide coatings of Fort St. Vrain (FSV) fertile and fissile and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. The report details the design task for the fuel particle crusher, including historical test information on double-roll crushers for carbide-coated fuels and the design approach selected for the cold pilot plant crusher, and shows how the design addresses the equipment goals and operational objectives. Design calculations and considerations are included to support the selection of crusher drive and gearing, the materials chosen for crushing rolls and housing, and the bearing selection. The results of the initial testing for compliance with design objectives and operational capabilities are also presented. 8 figures, 4 tables

  17. Is it economically feasible for farmers to grow their own fuel? A study of Camelina sativa produced in the western United States as an on-farm biofuel

    International Nuclear Information System (INIS)

    Keske, Catherine M.H.; Hoag, Dana L.; Brandess, Andrew; Johnson, Jerry J.

    2013-01-01

    This paper models the economic feasibility of growing the oilseed crop Camelina sativa (“camelina”) in the western United States to produce value-added protein feed supplement and an SVO-based biofuel. Modeled in eastern Colorado, this study demonstrates that camelina can be grown profitably both as a commodity and as an energy biofuel. These findings, along with the stochastic crop rotation budget and profitability sensitivity analysis, reflect unique contributions to the literature. The study's stochastic break-even analysis demonstrates a 0.51 probability of growing camelina profitably when diesel prices reach 1.15 $ L −1 . Results also show that the sale of camelina meal has the greatest impact on profitability. Yet once the price of diesel fuel exceeds 0.90 $ L −1 , the farmer generates more revenue from the ability to offset diesel fuel purchases than the revenues generated from the sale of camelina meal. A risk analysis using second degree stochastic dominance demonstrates that a risk-averse farmer would choose to grow camelina if the price of diesel equals or exceeds 1.31 $ L −1 . The article concludes that camelina can offset on-farm diesel use, making it economically feasible for farmers to grow their own fuel. As a result, camelina production may increase farm income, diversify rural economic development, and contribute to the attainment of energy policy goals. -- Highlights: •This is a stochastic budget analysis of growing camelina as SVO-based biofuel. •Results demonstrate economic feasibility for producers to grow their own fuel. •Camelina production can diversify regional and national energy portfolios. •Camelina production can contribute to on-farm energy independence

  18. LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J

    2008-09-08

    The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

  19. THE EFFECT OF GASOLINE-LIKE FUEL PRODUCED FROM WASTE AUTOMOBILE TIRES ON EMISSIONS IN SPARK-IGNITION ENGINES

    OpenAIRE

    ÖZTOP, H. F.; VAROL, Y.; ALTUN, Ş.; FIRAT, M.

    2016-01-01

    In the present paper, the effect of Gasoline-Like Fuel (GLF) on emissions was investigated for direct injection spark-ignited engine. The GLF was obtained from waste automobile tires by using the pyrolysis. The tires are installed to oven without any procedure such as cutting, melding etc. Obtained GLF was then used in a four-cylinder, four-stroke, water-cooled and direct injection spark-ignited engine as blended with unleaded gasoline from 0% to 60% with an increment of 10%. Engine tests res...

  20. International safeguards relevant to geologic disposal of high-level wastes and spent fuels

    International Nuclear Information System (INIS)

    Pillay, K.K.S.; Picard, R.R.

    1989-01-01

    Spent fuels from once-through fuel cycles placed in underground repositories have the potential to become attractive targets for diversion and/or theft because of their valuable material content and decreasing radioactivity. The first geologic repository in the US, as currently designed, will contain approximately 500 Mt of plutonium, 60,000 Mt of uranium and a host of other fissile and strategically important elements. This paper identifies some of the international safeguards issues relevant to the various proposed scenarios for disposing of the spent fuel. In the context of the US program for geologic disposal of spent fuels, this paper highlights several issues that should be addressed in the near term by US industries, the Department of Energy, and the Nuclear Regulatory Commission before the geologic repositories for spent fuels become a reality. Based on US spent fuel discharges, an example is presented to illustrate the enormity of the problem of verifying spent fuel inventories. The geologic disposal scenario for high-level wastes originating from defense facilities produced a ''practicably irrecoverable'' waste form. Therefore, safeguards issues for geologic disposal of high-level waste now in the US are less pressing. 56 refs. , 2 figs