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Sample records for fuel reprocessing process

  1. Handbook on process and chemistry on nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Atsuyuki (ed.) [Tokyo Univ., Tokyo (Japan); Asakura, Toshihide; Adachi, Takeo (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO{sub 2} fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  2. Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Harold F. McFarlane; Terry Todd

    2013-11-01

    Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore. Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor

  3. Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Michael F. Simpson; Jack D. Law

    2010-02-01

    This is an a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. No formal abstract was required for the article. The full article will be attached.

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

    Energy Technology Data Exchange (ETDEWEB)

    Susan Stacy; Julie Braun

    2006-12-01

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

  5. Process behavior and environmental assessment of /sup 14/C releases from an HTGR fuel reprocessing facility

    Energy Technology Data Exchange (ETDEWEB)

    Snider, J.W.; Kaye, S.V.

    1976-01-01

    Large quantities of /sup 14/CO/sub 2/ will be evolved when graphite fuel blocks are burned during reprocessing of spent fuel from HTGR reactors. The possible release of some or all of this /sup 14/C to the environment is a matter of concern which is investigated in this paper. Various alternatives are considered in this study for decontaminating and releasing the process off-gas to the environment. Concomitant radiological analyses have been done for the waste process scenarios to supply the necessary feedbacks for process design.

  6. Advanced aqueous reprocessing in P and T strategies: process demonstrations on genuine fuels and targets

    Energy Technology Data Exchange (ETDEWEB)

    Satmark, B.; Apostolidis, C.; Courson, O.; Malmbeck, R.; Carlos, R.; Pagliosa, G.; Romer, K.; Glatz, J.P. [European Commission, DG-JRC, Institute for Transuranium Elements, Hot Cell Technology, Karlsruhe (Germany)

    2000-07-01

    In the present work the performance of several processes used for advanced reprocessing of commercial LWR fuels as well as transmutation targets is compared. As a first step uranium and plutonium were recovered by PUREX type reprocessing. The raffinate, containing fission products, lanthanides and the minor actinides (MA) were used as feed for the second step in which minor actinides and lanthanides were separated from the bulk of the fission products. The five different processes tested use CMPO, DIDPA, TRPO, Diamide and CYANEX 923 as extractant. In the third step MA are separated from lanthanides. Here three processes were tested, i.e. using CYANEX 301, the synergistic mixture of di-chloro substituted CYANEX 301 and TOPO, and BTP solvents. Column-, batch- and continuous counter-current extraction techniques were used for the tests. The different processes will be described and discussed in terms of performances and efficiencies for Am and Cm. Efficient separation of MA from different genuine fuel solutions could be demonstrated and thereby also the possibility of closing a future transmutation fuel cycle. The combination, Diamide and BTP was found to be the best among extractants tested to achieve an efficient MA recovery from spent fuel. (authors)

  7. Advanced aqueous reprocessing in P and T strategies: process demonstrations on genuine fuels and targets

    Energy Technology Data Exchange (ETDEWEB)

    Christiansen, B.; Apostolidis, C.; Carlos, R.; Courson, O.; Glatz, J.P.; Malmbeck, R.; Pagliosa, G.; Roemer, K.; Serrano-Purroy, D. [European Commission, JRC, Inst. for Transuranium Elements, Karlsruhe (Germany)

    2004-07-01

    In the present work the performance of several processes used for advanced reprocessing of commercial LWR fuels as well as transmutation targets is compared. As a first step uranium and plutonium were recovered by PUREX type reprocessing. The raffinate, containing fission products including lanthanides and the minor actinides (MA) was used as feed for the second step in which minor actinides and lanthanides were separated from the bulk of the fission products. The five different processes tested use CMPO, DIDPA, TRPO, diamide and CYANEX 923 as extractants. In the third step MA are separated from lanthanides. Here three processes were tested, i.e. using CYANEX 301, the synergistic mixture of di-chloro substituted CYANEX 301 and TOPO, and BTP solvents. Column-, batch- and continuous counter-current extraction techniques were used for the tests. The different processes will be described and discussed in terms of performances and efficiencies for Am and Cm separation. Efficient separation of MA from different genuine fuel solutions could be demonstrated and thereby also the possibility of closing a future transmutation fuel cycle. The combination of diamide and BTP seems to be the best, among extractants tested, to achieve an efficient MA recovery from spent fuel. (orig.)

  8. Basic research on separation control of long life nuclides in fuel reprocessing processes

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Atsuyuki; Usami, Go [Tokyo Univ. (Japan). Faculty of Engineering; Maeda, Mitsuru; Fujine, Sachio; Uchiyama, Gunzo; Kihara, Takehiro; Asakura, Toshihide; Hotoku, Shinobu

    1996-01-01

    The behavior of technetium (Tc) in nuclear fuel reprocessing processes has become the subject to be elucidated in the transition to distribution process by coextraction and the catalytic action in distribution process. In order to forecast or control the behavior of Tc in reprocessing processes, it is necessary to understand that at which valence Tc exists stably in respective processes. Tc is stable at 7 valence in nitric acid solution expected in reprocessing. In this research, the reaction speed of the oxidation and reduction reactions of rhenium (Re) which simulates Tc was measured by laser Raman spectroscopy which can do high speed analysis of valence. The experimental method is explained. The Raman spectra of Re in the experimental system of this research were measured in perchloric acid solution and nitric acid solution, and compared with the values in literatures. As the result, the validity of this research was assured. It was confirmed that Re(7) was not reduced by sulfamic acid and ascorbic acid. Re(7) was reduced by thiocyanic acid once, but was oxidized again by the reaction of thiocyanic acid and nitric acid. (K.I.)

  9. Experimental research subject and renovation of chemical processing facility (CPF) for advanced fast reactor fuel reprocessing technology development

    Energy Technology Data Exchange (ETDEWEB)

    Koyama, Tomozo; Shinozaki, Tadahiro; Nomura, Kazunori; Koma, Yoshikazu; Miyachi, Shigehiko; Ichige, Yoshiaki; Kobayashi, Tsuguyuki; Nemoto, Shin-ichi [Japan Nuclear Cycle Development Inst., Tokai Works, Tokai, Ibaraki (Japan)

    2002-12-01

    In order to enhance economical efficiency, environmental impact and nuclear nonproliferation resistance, the Advanced Reprocessing Technology, such as simplification and optimization of process, and applicability evaluation of the innovative technology that was not adopted up to now, has been developed for the reprocessing of the irradiated fuel taken out from a fast reactor. Renovation of the hot cell interior equipments, establishment and updating of glove boxes, installation of various analytical equipments, etc. in the Chemical Processing Facility (CPF) was done to utilize the CPF more positivity which is the center of the experimental field, where actual fuel can be used, for research and development towards establishment of the Advanced Reprocessing Technology development. The hot trials using the irradiated fuel pins of the experimental fast reactor 'JOYO' for studies on improved aqueous reprocessing technology, MA separation technology, dry process technology, etc. are scheduled to be carried out with these new equipments. (author)

  10. Review of the literature for dry reprocessing oxide, metal, and carbide fuel: The AIROX, RAHYD, and CARBOX pyrochemical processes

    Energy Technology Data Exchange (ETDEWEB)

    Hoyt, R.C.; Rhee, B.W. [Rockwell International Corp., Canoga Park, CA (United States). Energy Systems Group

    1979-09-30

    The state of the art of dry processing oxide, carbide, and metal fuel has been determined through an extensive literature review. Dry processing in one of the most proliferation resistant fuel reprocessing technologies available to date, and is one of the few which can be exported to other countries. Feasibility has been established for oxide, carbide, and metal fuel on a laboratory scale, and large-scale experiments on oxide and carbide fuel have shown viability of the dry processing concept. A complete dry processing cycle has been demonstrated by multicycle processing-refabrication-reirradiation experiments on oxide fuel. Additional experimental work is necessary to: (1) demonstrate the complete fuel cycle for carbide and metal fuel, (2) optimize dry processing conditions, and (3) establish fission product behavior. Dry process waste management is easier than for an aqueous processing facility since wastes are primarily solids and gases. Waste treatment can be accomplished by techniques which have been, or are being, developed for aqueous plants.

  11. Fuel reprocessing tank

    Energy Technology Data Exchange (ETDEWEB)

    Gonda, Sumitora

    1998-10-09

    A tank of the present invention for spent fuels comprises a stainless steel tank main body for storing a highly corrosive dissolving solution, a steam jet pump disposed to the inside of the tank main body for transferring the dissolving solution to the outside of the tank main body and pipelines connecting them. With such a constitution, abnormal abrasion and drag of mechanical parts are less caused. In addition, a cleaning nozzle and a cleaning liquid pipeline which eliminates clogging of a sucking port of the steam jet pump if clogging is caused by sludges are disposed thereby enabling to avoid possibility of clogging. (T.M.)

  12. Reprocessing method for spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hoshikawa, Tadahiro; Sawa, Toshio; Suzuoki, Akira [Hitachi Ltd., Tokyo (Japan); Takashima, Yoichi; Kumagai, Mikiro

    1998-09-29

    The present invention provides a method of reprocessing spent fuels to form MOX having a Pu/U ratio suitable to fuels of LWR or fast reactors and uranium oxides of fuels of an LWR reactor. In a brief separation step for uranium, carbonate is added to a nitric acid solution in which spent fuels are dissolved, to dissolve a portion of uranium in the nitric acid solution. The residual uranium, plutonium and fission products are made into complexes of carboxylic acid ions and precipitated. The precipitated complexes of carboxylic acid ions are brought into contact with a different nitric acid solution to recover the uranium, plutonium and fission products. The concentration of the carbonate in the nitric acid solution in which uranium is partially dissolved is determined in accordance with the plutonium/uranium ratio based on the relation between the saturation concentration of uranium to the concentration of carbonate in the nitric acid solution. (T.M.)

  13. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ilas, G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Spencer, B. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    In nuclear fuel reprocessing, various radioactive elements enter the gas phase from the unit operations found in the reprocessing facility. In previous reports, the pathways and required removal were discussed for four radionuclides known to be volatile, 14C, 3H, 129I, and 85Kr. Other, less volatile isotopes can also report to the off-gas streams in a reprocessing facility. These were reported to be isotopes of Cs, Cd, Ru, Sb, Tc, and Te. In this report, an effort is made to determine which, if any, of 24 semivolatile radionuclides could be released from a reprocessing plant and, if so, what would be the likely quantities released. As part of this study of semivolatile elements, the amount of each generated during fission is included as part of the assessment for the need to control their emission. Also included in this study is the assessment of the cooling time (time out of reactor) before the fuel is processed. This aspect is important for the short-lived isotopes shown in the list, especially for cooling times approaching 10 y. The approach taken in this study was to determine if semivolatile radionuclides need to be included in a list of gas-phase radionuclides that might need to be removed to meet Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. A list of possible elements was developed through a literature search and through knowledge and literature on the chemical processes in typical aqueous processing of nuclear fuels. A long list of possible radionuclides present in irradiated fuel was generated and then trimmed by considering isotope half-life and calculating the dose from each to a maximum exposed individual with the US EPA airborne radiological dispersion and risk assessment code CAP88 (Rosnick 1992) to yield a short list of elements that actually need to be considered for control because they require high decontamination factors to meet a reasonable fraction of the regulated release. Each of these elements is

  14. Lanthanides extraction processes in molten fluoride media. Application to nuclear spent fuel reprocessing

    OpenAIRE

    Taxil, Pierre; Massot, Laurent; Nourry, Christophe; Gibilaro, Mathieu; Chamelot, Pierre; Cassayre, Laurent

    2009-01-01

    This paper describes four techniques of extraction of lanthanides elements (Ln) from molten salts in the general frame of reprocessing nuclear wastes; One of them is chemical: the precipitation of Ln ions in insoluble compounds (oxides or oxifluorides); the others use electrochemical methodology in molten fluorides for extraction and measurement of the progress of the processes: first electrodeposition of pure Ln metals on an inert cathode material was proved to be incomplete and cause probl...

  15. Status of the nuclear measurement stations for the process control of spent fuel reprocessing at AREVA NC/La Hague

    Energy Technology Data Exchange (ETDEWEB)

    Eleon, Cyrille; Passard, Christian; Hupont, Nicolas; Estre, Nicolas [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 St Paul-lez-Durance (France); Battel, Benjamin; Doumerc, Philippe; Dupuy, Thierry; Batifol, Marc [AREVA NC, La Hague plant - Nuclear Measurement Team, F-50444 Beaumont-Hague (France); Grassi, Gabriele [AREVA NC, 1 place Jean-Millier, 92084 Paris-La-Defense cedex (France)

    2015-07-01

    Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectroscopy, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality and safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remained after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations no. 1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station no. 1 allows determining the burn-up of the irradiated fuel by gamma-ray spectroscopy with HP Ge (high purity germanium) detectors. The burn-up is correlated to the {sup 137}Cs and {sup 134}Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station no. 3 is dedicated to the control of the correct fuel dissolution, which is performed with a {sup 137}Cs gamma ray measurement with a HP Ge detector. Station no. 7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially in the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional {sup 3}He detectors. The measurement stations have been validated for the reprocessing of Uranium Oxide (UOX) fuels with a burn-up rate up to 60 GWd/t. This paper presents a brief overview of the current status of the nuclear measurement stations. (authors)

  16. Equipment specifications for an electrochemical fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Hemphill, Kevin P [Los Alamos National Laboratory

    2010-01-01

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

  17. Consolidated Fuel Reprocessing Program. Progress report, January 1 to March 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    Unger, W.E. (comp.)

    1979-06-01

    On Oct. 1, 1978, a transition phase was begun to concentrate all US fuel reprocessing research in one major program, the Consolidated Fuel Reprocessing Program (CFRP). The CFRP is organized into the following: process R and D, engineering research, engineering systems, technical support, HTGR fuel reprocessing, and pyrochemical and dry processing methods. Progress is reported in each area. (DLC)

  18. Storage and Reprocessing of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    Addressing the problem of waste, especially high-level waste (HLW), is a requirement of the nuclear fuel cycle that cannot be ignored. We explore the two options employed currently, long-term storage and reprocessing.

  19. Impact of the use of the ferritic/martensitic ODS steels cladding on the fuel reprocessing PUREX process

    Science.gov (United States)

    Gwinner, B.; Auroy, M.; Mas, D.; Saint-Jevin, A.; Pasquier-Tilliette, S.

    2012-09-01

    Some ferritic/martensitic oxide dispersed strengthened (F/M ODS) steels are presently developed at CEA for the fuel cladding of the next generation of sodium fast nuclear reactors. The objective of this work is to study if this change of cladding could have any consequences on the spent fuel reprocessing PUREX process. During the fuel dissolution stage the cladding can actually be corroded by nitric acid. But some process specifications impose not to exceed a limit concentration of the corrosion products such as iron and chromium in the dissolution medium. For that purpose the corrosion behavior of these F/M ODS steels is studied in hot and concentrated nitric acid. The influence of some metallurgical parameters such as the chromium content, the elaboration process and the presence of the yttrium oxides is first discussed. The influence of environmental parameters such as the nitric acid concentration, the temperature and the presence of oxidizing species coming from the fuel is then analyzed. The corrosion rate is characterized by mass loss measurements and electrochemical tests. Analyses of the corroded surface are carried out by X-ray photoelectron spectroscopy.

  20. Development of spent fuel reprocessing process based on selective sulfurization: Study on the Pu, Np and Am sulfurization

    Science.gov (United States)

    Kirishima, Akira; Amano, Yuuki; Nihei, Toshifumi; Mitsugashira, Toshiaki; Sato, Nobuaki

    2010-03-01

    For the recovery of fissile materials from spent nuclear fuel, we have proposed a novel reprocessing process based on selective sulfurization of fission products (FPs). The key concept of this process is utilization of unique chemical property of carbon disulfide (CS2), i.e., it works as a reductant for U3O8 but works as a sulfurizing agent for minor actinides and lanthanides. Sulfurized FPs and minor actinides (MA) are highly soluble to dilute nitric acid while UO2 and PuO2 are hardly soluble, therefore, FPs and MA can be removed from Uranium and Plutonium matrix by selective dissolution. As a feasibility study of this new concept, the sulfurization behaviours of U, Pu, Np, Am and Eu are investigated in this paper by the thermodynamical calculation, phase analysis of chemical analogue elements and tracer experiments.

  1. Process Description and Operating History for the CPP-601/-640/-627 Fuel Reprocessing Complex at the Idaho National Engineering and Environmental Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    E. P. Wagner

    1999-06-01

    The Fuel Reprocessing Complex (FRC) at the Idaho Nuclear Technology and Engineering Center at the Idaho National Engineering and Environmental Laboratory was used for reprocessing spent nuclear fuel from the early 1950's until 1992. The reprocessing facilities are now scheduled to be deactivated. As part of the deactivation process, three Resource Conservation and Recovery Act (RCRA) interim status units located in the complex must be closed. This document gathers the historical information necessary to provide a rational basis for the preparation of a comprehensive closure plan. Included are descriptions of process operations and the operating history of the FRC. A set of detailed tables record the service history and present status of the process vessels and transfer lines.

  2. Reprocessing of research reactor fuel the Dounreay option

    Energy Technology Data Exchange (ETDEWEB)

    Cartwright, P.

    1997-08-01

    Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storage tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.

  3. Classic Nuclear Fuel Reprocessing Flowsheet

    Energy Technology Data Exchange (ETDEWEB)

    Fallgren, Andrew James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-02-13

    This is a flowsheet as well as a series of subsheets to be used for discussion on the standard design of a reprocessing plant. This flowsheet consists of four main sections: offgas handling, separations, solvent wash, and acid recycle. As well as having the main flowsheet, subsections have been broken off into their own sheets to provide for larger font and ease of printing.

  4. Reprocessing of LEU U-Mo Dispersion and Monolithic Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Vandegrift, G.F.; Jerden, J.; Stepinski, D.C.; Figueroa, J.; Williamson, M.A.; Kleeck, M.A. Van; Blaskovitz, R.J.; Ziegler, A.J.; Maggos, L.E.; Swanson, J.; Fortner, J.; Bakel, A.J. [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States)

    2011-07-01

    For conversion of high-performance research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, a fuel material with a higher density than uranium aluminide is required. Development studies are underway to develop U-Mo dispersion and monolithic fuels for conversion of several high- performance reactors. For dispersion fuels, development is narrowing down to a composition of U-7Mo dispersed in an aluminium matrix containing {approx}5% silicon. For monolithic fuels to be used in high performance research reactors in the United States, a zirconium-bonded U-10Mo foil appears to be the fuel of choice. For conversion to be realized a back-end disposition path is required for both fuels; one disposition pathway is reprocessing. Argonne National Laboratory is developing a pyroprocess for reprocessing spent monolithic fuel. Pyroprocessing was chosen over conventional aqueous solvent extraction due to the necessity of adding fluoride to the fuel-dissolution solution in order to dissolve the zirconium bonding layer on the U-Mo fuel. The proposed flowsheet and development activities will be described. A literature survey points to the ability to reprocess U-Mo dispersion fuels by an aqueous process, but due to several special characteristics of the fuel, the solvent-extraction flowsheets will be a departure from that normally used for the reprocessing of power reactor fuel. Special concerns that must be addressed in reprocessing these fuels are, for example, the low solubilities of uranyl molybdate, molybdic acid, and silicic acid in nitric acid solutions. This paper will address these concerns and development activities required to overcome them. (author)

  5. Development of the GANEX process for the reprocessing of Gen IV spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Miguirditchian, M.; Chareyre, L.; Sorel, C.; Bisel, I.; Baron, P.; Masson, M. [CEA Marcoule - DEN/DRCP/SCPS - BP 17171, Bagnols-sur-Ceze, 30207 (France)

    2008-07-01

    The GANEX (group actinide extraction) process is composed of two extraction cycles following the dissolution of the spent fuel. In the first cycle, uranium(VI) is selectively extracted from the dissolution solution using a mono amide extractant DEHiBA (NN--di-(ethyl-2-hexyl)iso-butyr-amide) diluted in HTP (hydrogen tetra-propylene). Experimental data and modelling of uranium(VI) and nitric acid extractions are presented. A flowsheet was designed and was successfully tested in laboratory scale mixer-settlers on a surrogate uranium(VI)/HNO{sub 3} feed. For the group actinide separation in the second cycle, the DIAMEX-SANEX process was adjusted to separate neptunium and plutonium along with americium and curium. The data showed the possibility to extract all actinides together with good selectivities versus lanthanides. The flowsheets of the two GANEX cycles which will be tested on a high active feed at the end of 2008 in Atalante facility are presented. (authors)

  6. Consolidated fuel reprocessing program. Progress report, January 1-March 31, 1981

    Energy Technology Data Exchange (ETDEWEB)

    1981-06-01

    Progress and activities are reported on process development, laboratory R and D, engineering research, engineering systems, Integrated Equipment Test (IET) facility operations, and HTGR fuel reprocessing. (DLC)

  7. Power Reactor Fuel Reprocessing: Mechanical Phase

    Energy Technology Data Exchange (ETDEWEB)

    Klima, B. B.

    1959-07-01

    The major events in the mechanical phase of the Power Reactor fuels reprocessing program during June were: 1. Feasibility of shearing of fuel elements without disassembly has been demonstrated in tests using porcelain-loaded prototype fuel elements. 2. Further work with the Manco shear was not deemed tb be advisable since permission has been granted to use another shear for cutting UO{sub 2}-loaded fuel elements. 3. Necessity to strip the windows in Building 3048, to sandblast, and repaint them has seriously disrupted occupancy of the cell by July 1. Start of installation probably will not be before August 1. 4. A cold SRE element should be received during July which will permit a direct look a t the problems associated with processing of these irradiated fuel elements. 5. Concurrence with AEC, Atomics International, and ORNL people on the fabrication of a poisoned carrier was obtained and all criteria for the carrier were released and the design was completed. 6. A decision was made to install and use a 24-inch Ty-Sa-Man saw which is on hand and was originally purchased for use in the Segmenting Facility for the SRE reprocessing. This will be used instead of the multipurpose saw to allow more time to refine the design of that saw. The multipurpose saw will be installed for use in subsequent reprocessing programs. This report will chronicle the changes in status which occurred during the calendar month of June. A complete description of each item is not included and may be found in the parent report. The dates indicated on the schedule have slipped since the last report primarily due to increase in scope of the work and postponement on all phases of the work except for the SRE preparations. Twenty-four new items have been added to the schedule. The status of procurement is shown. A total of 93 purchase requests have been turned in to t% Purchasing Department. A total of $199,261.83 has been committed by purchase orders, and a total of 56 purchase orders have been

  8. Stability of Solvent Radiolysis in Purex Process of Spent Fuel Reprocessing

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>In Purex process the regradation of TBP-diluents due to the radiolysis may cause diffculty for the process. Three diluents such as n-dodecane, hydrogenation kerosene, special kerosene were used in Purex

  9. Integrated international safeguards concepts for fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, E.A.; Gutmacher, R.G.; Markin, J.T.; Shipley, J.P.; Whitty, W.J.; Camp, A.L.; Cameron, C.P.; Bleck, M.E.; Ellwein, L.B.

    1981-12-01

    This report is the fourth in a series of efforts by the Los Alamos National Laboratory and Sandia National Laboratories, Albuquerque, to identify problems and propose solutions for international safeguarding of light-water reactor spent-fuel reprocessing plants. Problem areas for international safeguards were identified in a previous Problem Statement (LA-7551-MS/SAND79-0108). Accounting concepts that could be verified internationally were presented in a subsequent study (LA-8042). Concepts for containment/surveillance were presented, conceptual designs were developed, and the effectiveness of these designs was evaluated in a companion study (SAND80-0160). The report discusses the coordination of nuclear materials accounting and containment/surveillance concepts in an effort to define an effective integrated safeguards system. The Allied-General Nuclear Services fuels reprocessing plant at Barnwell, South Carolina, was used as the reference facility.

  10. Toward a Greenish Nuclear Fuel Cycle: Ionic Liquids as Solvents for Spent Nuclear Fuel Reprocessing and Other Decontamination Processes for Contaminated Metal Waste

    Science.gov (United States)

    Straka, Martin

    2016-12-01

    The final disposition of spent nuclear fuel (SNF) is an area that requires innovative solutions. The use of ionic liquids (ILs) has been examined as one means to remediate SNF in a variety of different chemical environments and with different chemical starting materials. The effectiveness of various ILs for SNF reprocessing, as well as the reaction chemistry that occurs in them, is discussed.

  11. A step towards closing the CANDU fuel cycle: an innovative scheme for reprocessing used CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Collins, F.; Lister, D. [Univ. of New Brunswick, UNB Nuclear, Dept. of Chemical Engineering, Fredericton, New Brunswick (Canada)

    2011-07-01

    Disposal versus reprocessing costs for used CANDU fuel was recently discussed by Rozon and Lister in a report produced for the Nuclear Waste Management Organization (NWMO). Their study discussed the economic incentives for reprocessing, not for the recovery of fissile uranium but for the recovery of plutonium ash. A $370/kg break-even price of uranium was calculated, and their model was found to be very sensitive to the reprocessing costs of the chosen technology. Findings were consistent with earlier studies done by Harvard University. Various reprocessing technologies (most based on solvent extraction) have been in use for many decades, but there appears to be no conceptual engineering study available in the open literature for a spent fuel reprocessing facility - one that includes process flows, operating costs and economic analysis. A deeper engineering study of the design and economics of re-processing technologies has since been undertaken by the nuclear group at the University of New Brunswick. An improved fluorination process was developed and modeled using ASPEN process simulation software. This study examines the impact of chosen technology on the spent fuel re-processing costs. (author)

  12. Neutronic evaluation of thorium and reprocessed fuels by GANEX and UREX+ in ADS

    Energy Technology Data Exchange (ETDEWEB)

    Barros, Graiciany, E-mail: graiciany.barros@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Velasquez, Carlos E.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L., E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    A conceptual design of accelerator driven systems (ADS) that utilize thorium and reprocessed fuel in order to produce {sup 233}U and to transmute high radiotoxicity isotopes in spent nuclear fuel has been proposed. The use of thorium and reprocessed fuel in an ADS is one of the clean, safe, and economical solutions for the problem of nuclear waste. In this study, the aim was to compare the neutronic behavior of the core using spent fuel reprocessed by GANEX (Group ActiNide EXtraction) and UREX+ (Uranium Extraction), both spiked with thorium. The simulated design was a cylinder fuelled with a hexagonal lattice with 156 fuel rods. One of the studied fuels was a mixture based upon Pu-MA, removed from PWR-spent fuel, theoretically reprocessed by GANEX reprocessing and spiked with 82% of thorium. The other fuel was a reprocessed fuel obtained theoretically from UREX+ (Uranium Extraction) process and spiked with 82% of thorium. Monteburns 2.0 (MCNP5/ORIGEN 2.1) code was used to simulate the neutronic aspects of the fuels. The multiplication factors, the neutron spectra, and the nuclear fuel evolution were analyzed during 10 years of burn-up. The results allowed comparing the two reprocessing techniques, the {sup 233}U production and the reduction in the amount of high radiotoxicity isotopes of these fuels. (author)

  13. Process monitoring in international safeguards for reprocessing plants: A demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Ehinger, M.H.

    1989-01-01

    In the period 1985--1987, the Oak Ridge National Laboratory investigated the possible role of process monitoring for international safeguards applications in fuel reprocessing plants. This activity was conducted under Task C.59, ''Review of Process Monitoring Safeguards Technology for Reprocessing Facilities'' of the US program of Technical Assistance to the International Atomic Energy Agency (IAEA) Safeguards program. The final phase was a demonstration of process monitoring applied in a prototypical reprocessing plant test facility at ORNL. This report documents the demonstration and test results. 35 figs.

  14. Nuclear fuel reprocessing and high level waste disposal: informational hearings. Volume V. Reprocessing. Part 2

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-03-08

    Testimony was presented by a four member panel on the commercial future of reprocessing. Testimony was given on the status of nuclear fuel reprocessing in the United States. The supplemental testimony and materials submitted for the record are included in this report. (LK)

  15. Airborne effluent control for LMFBR fuel reprocessing plants

    Energy Technology Data Exchange (ETDEWEB)

    Yarbro, O.O.; Groenier, W.S.; Stephenson, M.J.

    1976-01-01

    A significant part of the LMFBR fuel reprocessing development program has been devoted to the development of efficient removal systems for the volatile fission products, including /sup 131/I, krypton, tritium, /sup 129/I, and most recently /sup 14/C. Flowsheet studies have indicated that very significant reductions of radioactive effluents can be achieved by integrating advanced effluent control systems with new concepts of containment and ventilation; however, the feasibility of such has not yet been established, nor have the economics been examined. This paper presents a flowsheet for the application of advanced containment systems to the processing of LMFBR fuels and summarizes the status and applicability of specific fission product removal systems.

  16. Consolidated fuel-reprocessing program. Progress report, April 1-June 30, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Burch, W D

    1982-09-01

    Highlights of progress accomplished during the quarter ending June 30, 1982 are summarized. Discussion is presented under the headings: Process development; Laboratory R and D; Engineering research; Engineering systems; Integrated equipment test facility operation; Instrument development; and HTGR fuel reprocessing.

  17. Proof of Concept Simulations of the Multi-Isotope Process Monitor: An Online, Nondestructive, Near-Real-Time Safeguards Monitor for Nuclear Fuel Reprocessing Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Orton, Christopher R.; Fraga, Carlos G.; Christensen, Richard; Schwantes, Jon M.

    2011-02-11

    The International Atomic Energy Agency (IAEA) will require the development of advanced technologies to effectively safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of nondestructive, near-real-time, autonomous process monitoring. This paper describes recent results from model simulations designed to test the Multi-Isotope Process (MIP) monitor, a novel approach to safeguarding reprocessing plants. The MIP monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in near-real-time. Three computer models including ORIGEN-ARP, AMUSE, and SYNTH were used in series to predict spent nuclear fuel composition, estimate element partitioning during separation, and simulate spectra from product and raffinate streams using a variety of gamma detectors, respectively. Simulations were generated for fuel with various irradiation histories and under a variety of plant operating conditions. Principal component analysis (PCA) was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup, and cooling time. Hierarchical cluster analysis (HCA) and partial least squares (PLS) were also used in the analysis. The MIP monitor was found to be sensitive to induced variations of several operating parameters including distinguishing ±2.5% variation from normal process acid concentrations. The ability of PLS to predict burnup levels from simulated spectra was also demonstrated to be within 3.5% of measured values.

  18. Proof of concept simulations of the Multi-Isotope Process monitor: An online, nondestructive, near-real-time safeguards monitor for nuclear fuel reprocessing facilities

    Science.gov (United States)

    Orton, Christopher R.; Fraga, Carlos G.; Christensen, Richard N.; Schwantes, Jon M.

    2011-02-01

    The International Atomic Energy Agency will require the development of advanced technologies to effectively safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of nondestructive, near-real-time, autonomous process monitoring. This paper describes recent results from model simulations designed to test the Multi-Isotope Process (MIP) monitor, a novel addition to a safeguards system for reprocessing facilities. The MIP monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in near-real-time. Three computer models including ORIGEN-ARP, AMUSE, and SYNTH were used in series to predict spent nuclear fuel composition, estimate element partitioning during separation, and simulate spectra from product and raffinate streams using a variety of gamma detectors, respectively. Simulations were generated for fuel with various irradiation histories and under a variety of plant operating conditions. Principal component analysis was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup, and cooling time. Hierarchical cluster analysis and partial least squares (PLS) were also used in the analysis. The MIP monitor was found to be sensitive to induced variations of several operating parameters including distinguishing ±2.5% variation from normal process acid concentrations. The ability of PLS to predict burnup levels from simulated spectra was also demonstrated to be within 3.5% of measured values.

  19. Corrosion studies in fuel element reprocessing environments containing nitric acid

    Energy Technology Data Exchange (ETDEWEB)

    Beavers, J A; White, R R; Berry, W E; Griess, J C

    1982-04-01

    Nitric acid is universally used in aqueous fuel element reprocessing plants; however, in the processing scheme being developed by the Consolidated Fuel Reprocessing Program, some of the equipment will be exposed to nitric acid under conditions not previously encountered in fuel element reprocessing plants. A previous report presented corrosion data obtained in hyperazeotropic nitric acid and in concentrated magnesium nitrate solutions used in its preparation. The results presented in this report are concerned with the following: (1) corrosion of titanium in nitric acid; (2) corrosion of nickel-base alloys in a nitric acid-hydrofluoric acid solution; (3) the formation of Cr(VI), which enhances corrosion, in nitric acid solutions; and (4) corrosion of mechanical pipe connectors in nitric acid. The results show that the corrosion rate of titanium increased with the refreshment rate of boiling nitric acid, but the effect diminished rapidly as the temperature decreased. The addition of iodic acid inhibited attack. Also, up to 200 ppM of fluoride in 70% HNO/sub 3/ had no major effect on the corrosion of either titanium or tantalum. In boiling 8 M HNO/sub 3/-0.05 M HF, Inconel 671 was more resistant than Inconel 690, but both alloys experienced end-grain attack. In the case of Inconel 671, heat treatment was very important; annealed and quenched material was much more resistant than furnace-cooled material.The rate of oxidation of Cr(III) to Cr(VI) increased significantly as the nitric acid concentration increased, and certain forms of ruthenium in the solution seemed to accelerate the rate of formation. Mechanical connectors of T-304L stainless steel experienced end-grain attack on the exposed pipe ends, and seal rings of both stainless steel and a titanium alloy (6% Al-4% V) underwent heavy attack in boiling 8 M HNO/sub 3/.

  20. Proof of concept experiments of the multi-isotope process monitor: An online, nondestructive, near real-time monitor for spent nuclear fuel reprocessing facilities

    Energy Technology Data Exchange (ETDEWEB)

    Orton, Christopher R., E-mail: christopher.orton@pnnl.gov [Pacific Northwest National Laboratory, 902 Battelle Boulevard, P.O. Box 999, Richland, WA 99354 (United States); Fraga, Carlos G., E-mail: carlos.fraga@pnnl.gov [Pacific Northwest National Laboratory, 902 Battelle Boulevard, P.O. Box 999, Richland, WA 99354 (United States); Christensen, Richard N., E-mail: christensen.3@osu.edu [The Ohio State University, 201W. 19th Avenue, Columbus, Ohio 43210 (United States); Schwantes, Jon M., E-mail: jon.schwantes@pnnl.gov [Pacific Northwest National Laboratory, 902 Battelle Boulevard, P.O. Box 999, Richland, WA 99354 (United States)

    2012-04-21

    Operators, national regulatory agencies and the IAEA will require the development of advanced technologies to efficiently control and safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of non-destructive, near real-time (NRT), autonomous process monitoring. This paper describes results from proof-of-principle experiments designed to test the multi-isotope process (MIP) monitor, a novel approach to monitoring and safeguarding reprocessing facilities. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in NRT. Commercial spent nuclear fuel of various irradiation histories was dissolved and separated using a PUREX-based batch solvent extraction. Extractions were performed at various nitric acid concentrations to mimic both normal and off-normal industrial plant operating conditions. Principal component analysis (PCA) was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup and cooling time. Partial least squares (PLS) regression was applied to attempt to quantify both the acid concentration and burnup of the dissolved spent fuel during the initial separation stage of recycle. The MIP Monitor demonstrated sensitivity to induced variations of acid concentration, including the distinction of {+-}1.3 M variation from normal process conditions by way of PCA. Acid concentration was predicted using measurements from the organic extract and PLS resulting in predictions with <0.7 M relative error. Quantification of burnup levels from dissolved fuel spectra using PLS was demonstrated to be within 2.5% of previously measured values.

  1. Proof of Concept Experiments of the Multi-Isotope Process Monitor: An Online, Nondestructive, Near Real-Time Monitor for Spent Nuclear Fuel Reprocessing Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Orton, Christopher R.; Fraga, Carlos G.; Christensen, Richard; Schwantes, Jon M.

    2012-04-21

    Operators, national regulatory agencies and the IAEA will require the development of advanced technologies to efficiently control and safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of non-destructive, near-real-time (NRT), autonomous process monitoring. This paper describes results from proof-of-principle experiments designed to test the Multi-Isotope Process (MIP) Monitor, a novel approach to safeguarding reprocessing facilities. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in NRT. Commercial spent nuclear fuel of various irradiation histories was dissolved and separated using a PUREX-based batch solvent extraction. Extractions were performed at various nitric acid concentrations to mimic both normal and off-normal industrial plant operating conditions. Principal Component Analysis (PCA) was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup and cooling time. Partial Least Squares (PLS) regression was applied to attempt to quantify both the acid concentration and burnup of the dissolved spent fuel during the initial separation stage of recycle. The MIP Monitor demonstrated sensitivity to induced variations of acid concentration, including the distinction of {+-} 1.3 M variation from normal process conditions by way of PCA. Acid concentration was predicted using measurements from the organic extract and PLS resulting in predictions with <0.7 M relative error. Quantification of burnup levels from dissolved fuel spectra using PLS was demonstrated to be within 2.5% of previously measured values.

  2. Proof of concept experiments of the multi-isotope process monitor: An online, nondestructive, near real-time monitor for spent nuclear fuel reprocessing facilities

    Science.gov (United States)

    Orton, Christopher R.; Fraga, Carlos G.; Christensen, Richard N.; Schwantes, Jon M.

    2012-04-01

    Operators, national regulatory agencies and the IAEA will require the development of advanced technologies to efficiently control and safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of non-destructive, near real-time (NRT), autonomous process monitoring. This paper describes results from proof-of-principle experiments designed to test the multi-isotope process (MIP) monitor, a novel approach to monitoring and safeguarding reprocessing facilities. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in NRT. Commercial spent nuclear fuel of various irradiation histories was dissolved and separated using a PUREX-based batch solvent extraction. Extractions were performed at various nitric acid concentrations to mimic both normal and off-normal industrial plant operating conditions. Principal component analysis (PCA) was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup and cooling time. Partial least squares (PLS) regression was applied to attempt to quantify both the acid concentration and burnup of the dissolved spent fuel during the initial separation stage of recycle. The MIP Monitor demonstrated sensitivity to induced variations of acid concentration, including the distinction of ±1.3 M variation from normal process conditions by way of PCA. Acid concentration was predicted using measurements from the organic extract and PLS resulting in predictions with <0.7 M relative error. Quantification of burnup levels from dissolved fuel spectra using PLS was demonstrated to be within 2.5% of previously measured values.

  3. Computer code system for the R and D of nuclear fuel cycle with fast reactor. 4. Development of an object-oriented analysis code for estimation of the material balance in the pyrochemical reprocessing process

    Energy Technology Data Exchange (ETDEWEB)

    Okamura, Nobuo; Sato, Koji [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2002-03-01

    An analysis code using the object-oriented software EX{center_dot}TD Ver.4 was developed for the estimation of material balance for the system design of the pyrochemical reprocessing plants consisting of batch processes. This code can also estimate the radioactivity balance, decay heat balance and holdup, and easily cope with the improvement of the process flow, and so on. An example of the material balance estimation under the consideration of the solvent (molten salt) recycling time is presented for the oxide electrowinning reprocessing system designed in the feasibility study of the FBR fuel cycle system. The results indicate the possibility of reduction of the vitrified waste form volume due to the extension of the recycling time of the solvent. This paper describes the outline of the code and estimation of the material balance in the oxide electrowinning reprocessing system under consideration of the solvent recycling time. (author)

  4. Study of an ADS Loaded with Thorium and Reprocessed Fuel

    Directory of Open Access Journals (Sweden)

    Graiciany de Paula Barros

    2012-01-01

    Full Text Available Accelerator-driven systems (ADSs are investigated for long-lived fission product transmutation and fuel regeneration. The aim of this paper is to investigate the nuclear fuel evolution and the neutronic parameters of a lead-cooled accelerator-driven system used for fuel breeding. The fuel used in some fuel rods was T232hO2 for U233 production. In the other fuel rods was used a mixture based upon Pu-MA, removed from PWR-spent fuel, reprocessed by GANEX, and finally spiked with thorium or depleted uranium. The use of reprocessed fuel ensured the use of T232hO2 without the initial requirement of U233 enrichment. In this paper was used the Monte Carlo code MCNPX 2.6.0 that presents the depletion/burnup capability, combining an ADS source and kcode-mode (for criticality calculations. The multiplication factor (keff evolution, the neutron energy spectra in the core at BOL, and the nuclear fuel evolution during the burnup were evaluated. The results indicated that the combined use of T232hO2 and reprocessed fuel allowed U233 production without the initial requirement of U233 enrichment.

  5. MICROBIAL TRANSFORMATIONS OF RADIONUCLIDES RELEASED FROM NUCLEAR FUEL REPROCESSING PLANTS.

    Energy Technology Data Exchange (ETDEWEB)

    FRANCIS,A.J.

    2006-10-18

    Microorganisms can affect the stability and mobility of the actinides U, Pu, Cm, Am, Np, and the fission products Tc, I, Cs, Sr, released from nuclear fuel reprocessing plants. Under appropriate conditions, microorganisms can alter the chemical speciation, solubility and sorption properties and thus could increase or decrease the concentrations of radionuclides in solution and the bioavailability. Dissolution or immobilization of radionuclides is brought about by direct enzymatic action or indirect non-enzymatic action of microorganisms. Although the physical, chemical, and geochemical processes affecting dissolution, precipitation, and mobilization of radionuclides have been investigated, we have only limited information on the effects of microbial processes. The mechanisms of microbial transformations of the major and minor actinides and the fission products under aerobic and anaerobic conditions in the presence of electron donors and acceptors are reviewed.

  6. Characteristics and behavior of emulsion at nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Gonda, K.; Nemoto, T.; Oka, K.

    1982-05-01

    The characteristics and behavior of the emulsion formed in mixer-settlers during nuclear fuel reprocessing were studied with the dissolver solution of spent fuel burned up to 28,000 MWd/MTU and a palladium colloidal solution, respectively. The emulsion was observed to be oil in water where nonsoluble residues of spent fuel were condensed as emulsifiers. Emulsion formed at interfaces in the settler showed electric conductivity due to continuity of the aqueous phase of the emulsion and viscosity due to the creamy state of the emulsion. The higher the palladium particle concentration was, the larger the amount of emulsion formed. This result agreed well with experience obtained in the Tokai Reprocessing Plant operation that both nonsoluble residues and emulsion formation increased remarkably on fuels in which burnup exceeded 20 000 MWd/MTU.

  7. Strategy and current state of research on enhanced iodine separation during spent fuel reprocessing by the Purex process

    Energy Technology Data Exchange (ETDEWEB)

    Devisme, F.; Juvenelle, A.; Touron, E. [CEA Valrho, Dir. de l' Energie Nucleaire, DEN/DRCP, 30 - Marcoule (France)

    2001-07-01

    An enhanced separation process designed to recover and purify molecular iodine desorbed during dissolution is described in the context of {sup 129}I management in the Purex process for transmutation or interim storage. It involves reducing acid scrubbing with hydroxyl-ammonium nitrate followed by oxidation with hydrogen peroxide to obtain selective desorption. The stoichiometry and kinetics are determined for each step and an experimental validation program is now in progress using a small pilot facility equipped with a scrubbing column. The technical feasibility of the process has already been demonstrated: room-temperature scrubbing with a HAN solution (0,5 mol.L{sup -1}) at a pH of about 5 results in 99% iodine trapping efficiency; the subsequent desorption yield is 99,5%. (author)

  8. Electrolysis cell for reprocessing plutonium reactor fuel

    Science.gov (United States)

    Miller, William E.; Steindler, Martin J.; Burris, Leslie

    1986-01-01

    An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals, the cell including a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket and the anode basket being extendable into the lower pool to dissolve at least some metallic contaminants, the anode basket containing the spent fuel acting as a second anode when in the electrolyte.

  9. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  10. Hybrid reprocessing technology of fluoride volatility and solvent extraction. New reprocessing technology, FLUOREX, for LWR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Fumio [Hitachi Ltd., Ibaraki (Japan)

    2002-11-01

    Hybrid Process of Fluoride Volatility and Solvent Extraction (FLUOREX) has been objected to develop a low cost reprocessing technology for collection of U and MOX (mixture U and Pu) in LWR fuel cycle. Outline, characteristics, technologies, problems and material balance of FLUOREX are explained. LWR spent fuel consists of about 96% U, 1% Pu and about 3% fission products (FP) and minor actinides (MA). FLUOREX method is hybrid system, which isolates about 90% U at high speed and refines by fluoride volatility process and residue about 10% U, Pu, MA and FP are processed by PUREX method after dissolution in acid. The special features are low cost by small type and lightweight, stable without gas Pu and stop of fluorine gas, reducing load of environment, resistance of nuclear proliferation, application of technologies demonstrated and flexible method for fast reactor. Three problems for development are selective fluoridation of U, transportation of oxides in the fluoride residue and dissolution of transported oxides. The preliminary examination of plan showed 800GWD/t processing volume, 200 day/year operation day, about 51 ten-thousand cubic meter volume of plant, about 1/3 Rokkasho reprocessing plant. (S.Y.)

  11. Radioactive Iodine and Krypton Control for Nuclear Fuel Reprocessing Facilities

    Directory of Open Access Journals (Sweden)

    Nick R. Soelberg

    2013-01-01

    Full Text Available The removal of volatile radionuclides generated during used nuclear fuel reprocessing in the US is almost certain to be necessary for the licensing of a reprocessing facility in the US. Various control technologies have been developed, tested, or used over the past 50 years for control of volatile radionuclide emissions from used fuel reprocessing plants. The US DOE has sponsored, since 2009, an Off-gas Sigma Team to perform research and development focused on the most pressing volatile radionuclide control and immobilization problems. In this paper, we focus on the control requirements and methodologies for 85Kr and 129I. Numerous candidate technologies have been studied and developed at laboratory and pilot-plant scales in an effort to meet the need for high iodine control efficiency and to advance alternatives to cryogenic separations for krypton control. Several of these show promising results. Iodine decontamination factors as high as 105, iodine loading capacities, and other adsorption parameters including adsorption rates have been demonstrated under some conditions for both silver zeolite (AgZ and Ag-functionalized aerogel. Sorbents, including an engineered form of AgZ and selected metal organic framework materials (MOFs, have been successfully demonstrated to capture Kr and Xe without the need for separations at cryogenic temperatures.

  12. Selected studies in HTGR reprocessing development. [KA2C process

    Energy Technology Data Exchange (ETDEWEB)

    Notz, K.J.

    1976-03-01

    Recent work at ORNL on hot cell studies, off-gas cleanup, and waste handling is reviewed. The work includes small-scale burning tests with irradiated fuels to study fission product release, development of the KALC process for the removal of /sup 85/Kr from a CO/sub 2/ stream, preliminary work on a nonfluidized bed burner, solvent extraction studies including computer modeling, characterization of reprocessing wastes, and initiation of a development program for the fixation of /sup 14/C as CaCO/sub 3/. (auth)

  13. Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Carlsen, Brett; Chiguer, Mustapha; Grahn, Per; Sampson, Michele; Wolff, Dietmar; Bevilaqua, Arturo; Wasinger, Karl; Saegusa, Toshiari; Seelev, Igor

    2016-09-01

    Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years until reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on

  14. Overview of reductants utilized in nuclear fuel reprocessing/recycling

    Energy Technology Data Exchange (ETDEWEB)

    Patricia Paviet-Hartmann; Catherine Riddle; Keri Campbell; Edward Mausolf

    2013-10-01

    Most of the aqueous processes developed, or under consideration worldwide for the recycling of used nuclear fuel (UNF) utilize the oxido-reduction properties of actinides to separate them from other radionuclides. Generally, after acid dissolution of the UNF, (essentially in nitric acid solution), actinides are separated from the raffinate by liquid-liquid extraction using specific solvents, associated along the process, with a particular reductant that will allow the separation to occur. For example, the industrial PUREX process utilizes hydroxylamine as a plutonium reductant. Hydroxylamine has numerous advantages: not only does it have the proper attributes to reduce Pu(IV) to Pu(III), but it is also a non-metallic chemical that is readily decomposed to innocuous products by heating. However, it has been observed that the presence of high nitric acid concentrations or impurities (such as metal ions) in hydroxylamine solutions increase the likelihood of the initiation of an autocatalytic reaction. Recently there has been some interest in the application of simple hydrophilic hydroxamic ligands such as acetohydroxamic acid (AHA) for the stripping of tetravalent actinides in the UREX process flowsheet. This approach is based on the high coordinating ability of hydroxamic acids with tetravalent actinides (Np and Pu) compared with hexavalent uranium. Thus, the use of AHA offers a route for controlling neptunium and plutonium in the UREX process by complexant based stripping of Np(IV) and Pu(IV) from the TBP solvent phase, while U(VI) ions are not affected by AHA and remain solvated in the TBP phase. In the European GANEX process, AHA is also used to form hydrophilic complexes with actinides and strip them from the organic phase into nitric acid. However, AHA does not decompose completely when treated with nitric acid and hampers nitric acid recycling. In lieu of using AHA in the UREX + process, formohydroxamic acid (FHA), although not commercially available, hold

  15. Consolidated fuel reprocessing. Program progress report, April 1-June 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-09-01

    This progress report is compiled from major contributions from three programs: (1) the Advanced Fuel Recycle Program at ORNL; (2) the Converter Fuel Reprocessing Program at Savannah River Laboratory; and (3) the reprocessing components of the HTGR Fuel Recycle Program, primarily at General Atomic and ORNL. The coverage is generally overview in nature; experimental details and data are limited.

  16. Krypton-85 health risk assessment for a nuclear fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Mellinger, P.J.; Brackenbush, L.W.; Tanner, J.E.; Gilbert, E.S.

    1984-08-01

    The risks involved in the routine release of /sup 85/Kr from nuclear fuel reprocessing operations to the environment were compared to those resulting from the capture and storage of /sup 85/Kr. Instead of releasing the /sup 85/Kr to the environment when fuel is reprocessed, it can be captured, immobilized and stored. Two alternative methods of capturing /sup 85/Kr (cryogenic distillation and fluorocarbon absorption) and one method of immobilizing the captured gas (ion implantation/sputtering) were theoretically incorporated into a representative fuel reprocessing plant, the Barnwell Nuclear Fuel Plant, even though there are no known plans to start up this facility. Given the uncertainties in the models used to generate lifetime risk numbers (0.02 to 0.027 radiation induced fatal cancers expected in the occupational workforce and 0.017 fatal cancers in the general population), the differences in total risks for the three situations, (i.e., no-capture and two-capture alternatives) cannot be considered meaningful. It is possible that no risks would occur from any of the three situations. There is certainly no reason to conclude that risks from /sup 85/Kr routinely released to the environment are greater than those that would result from the other two situations considered. Present regulations mandate recovery and disposal of /sup 85/Kr from the off gases of a facility reprocessing spent fuel from commercial sources. Because of the lack of a clear-cut indication that recovery woud be beneficial, it does not seem prudent to burden the facilities with a requirement for /sup 85/Kr recovery, at least until operating experience demonstrates the incentive. The probable high aging of the early fuel to be processed and the higher dose resulting from the release of the unregulated /sup 3/H and /sup 14/C also encourage delaying implementation of the /sup 85/Kr recovery in the early plants.

  17. Materials management in an internationally safeguarded fuels reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, E.A.; Baker, A.L.; Cobb, D.D.

    1980-04-01

    The following appendices are included: aqueous reprocessing and conversion technology, reference facilities, process design and operating features relevant to materials accounting, operator's safeguards system structure, design principles of dynamic materials accounting systems, modeling and simulation approach, optimization of measurement control, aspects of international verification problem, security and reliability of materials measurement and accounting system, estimation of in-process inventory in solvent-extraction contactors, conventional measurement techniques, near-real-time measurement techniques, isotopic correlation techniques, instrumentation available to IAEA inspectors, and integration of materials accounting and containment and surveillance. (DLC)

  18. Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

    Energy Technology Data Exchange (ETDEWEB)

    Van Hecke, K.; Goethals, P.

    2006-07-15

    The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

  19. 76 FR 34007 - Draft Regulatory Basis for a Potential Rulemaking on Spent Nuclear Fuel Reprocessing Facilities

    Science.gov (United States)

    2011-06-10

    ...: NUREG-1909, a white paper authored by the Advisory Committee on Nuclear Waste and Materials, titled... waste through developing more sophisticated reprocessing technologies. During the Bush Administration... reprocessing spent fuel and deploying fast reactors to burn long-lived actinides. In response to these...

  20. Consolidated Fuel-Reprocessing Program. Progress report, April 1 to June 30, 1983

    Energy Technology Data Exchange (ETDEWEB)

    1983-08-01

    All research and development on fuel reprocessing in the United States is managed under the Consolidated Fuel Reprocessing Program. Technical progress is reported in overview fashion. Conceptual studies for the proposed Breeder Reprocessing Engineering Test (BRET) have continued. Studies to date have confirmed the feasibility of modifying an existing DOE facility at Hanford, Washington. A study to measure the extent of plutonium polymerization during steam-jet transfers of nitric acid solutions indicated polymer would appear only after several successive transfers at temperatures of 75/sup 0/C or higher. Fast-Flux Test Facility fuel was processed for the first time in the Solvent Extraction Test Facility. Studies of krypton release from pulverized sputter-deposited Ni-Y-Kr matrices have shown that the release rate is inversely proportional to the particle radius at 200/sup 0/C. Preparation of the initial 500-g batch of mixed oxide gel-spheres was completed. Fabrication processing at HEDL of mixed oxide gel-spheres (DIPRES process) was initiated. Operational testing of both 8 packs of the centrifugal contactor has been completed. Fabrication of both the prototypical disassembly system and the prototypical shear system has been initiated. Planning for FY 1984 installation and modification work in the integrated equipment list facility was completed. Acceptance tests of the original Integrated Process Demonstration system have been completed. Instrumentation and controls work with the prototype multiwavelength uranium photometer was successful and has been expanded to continuously and simultaneously monitor three process streams (raffinate, aqueous feed, and organic strip) in the secondary extraction cycle. Major efforts of the environmental, safeguards, and waste management areas were directed toward providing data for BRET.

  1. Reprocessability of molybdenum and magnesia based inert matrix fuels

    Directory of Open Access Journals (Sweden)

    Ebert Elena L.

    2015-12-01

    Full Text Available This work focuses on the reprocessability of metallic 92Mo and ceramic MgO, which is under investigation for (Pu,MA-oxide (MA = minor actinide fuel within a metallic 92Mo matrix (CERMET and a ceramic MgO matrix (CERCER. Magnesium oxide and molybdenum reference samples have been fabricated by powder metallurgy. The dissolution of the matrices was studied as a function of HNO3 concentration (1-7 mol/L and temperature (25-90°C. The rate of dissolution of magnesium oxide and metallic molybdenum increased with temperature. While the MgO rate was independent of the acid concentration (1-7 mol/L, the rate of dissolution of Mo increased with acid concentration. However, the dissolution of Mo at high temperatures and nitric acid concentrations was accompanied by precipitation of MoO3. The extraction of uranium, americium, and europium in the presence of macro amounts of Mo and Mg was studied by three different extraction agents: tri-n-butylphosphate (TBP, N,Nʹ-dimethyl-N,Nʹ-dioctylhexylethoxymalonamide (DMDOHEMA, and N,N,N’,N’- -tetraoctyldiglycolamide (TODGA. With TBP no extraction of Mo and Mg occurred. Both matrix materials are partly extracted by DMDOHEMA. Magnesium is not extracted by TODGA (D < 0.1, but a weak extraction of Mo is observed at low Mo concentration.

  2. 乏燃料后处理溶解过程核临界安全初步分析%Nuclear Criticality Safety Analysis in Dissolving Process of Spent Fuel Reprocessing

    Institute of Scientific and Technical Information of China (English)

    刘颖瑜; 骆志文; 刘振华

    2013-01-01

    A rational space distribution model for spent fuel element dissolving at each stage of process in reprocessing plant was formulated .The nuclear criticality on safety issue was studied by calculating numerical model on the process of spent fuel dissolution in consideration of the given plant arrangement . An assessment of the influence of several main critical parameters to the plant safety was given .The calculation results show that the most dangerous status occurs at the initial stage of dissolution w hen fissile nuclide transforms under ideal conditions . A negative influence to the system is indicated by the increase of temperature and concentration of nitric acid ,and the effect is less than 4% .System safety can be improved greatly by the addition of neutron poison or the application of fuel burnup credit ,and the effect reaches 30% .%通过建立合理的空间分布模型,对后处理厂乏燃料溶解不同阶段的核临界安全问题进行分析,同时对重要的核临界安全参数给予影响评价。结果显示,在仅考虑易裂变核素形态转变的理想情况下,溶解初期为最危险状态;温度升高和硝酸浓度增大对系统的影响为负效应,影响均小于4%;可溶中子毒物的加入与燃耗信任制技术的应用能大幅提高系统的经济性,影响均可达到30%。

  3. Pyrochemical reprocessing of molten salt fast reactor fuel: focus on the reductive extraction step

    OpenAIRE

    Rodrigues Davide; Durán-Klie Gabriela; Delpech Sylvie

    2015-01-01

    The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile...

  4. Advanced Process Monitoring Techniques for Safeguarding Reprocessing Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Orton, Christopher R.; Bryan, Samuel A.; Schwantes, Jon M.; Levitskaia, Tatiana G.; Fraga, Carlos G.; Peper, Shane M.

    2010-11-30

    The International Atomic Energy Agency (IAEA) has established international safeguards standards for fissionable material at spent fuel reprocessing plants to ensure that significant quantities of weapons-grade nuclear material are not diverted from these facilities. For large throughput nuclear facilities, it is difficult to satisfy the IAEA safeguards accountancy goal for detection of abrupt diversion. Currently, methods to verify material control and accountancy (MC&A) at these facilities require time-consuming and resource-intensive destructive assay (DA). Leveraging new on-line non destructive assay (NDA) process monitoring techniques in conjunction with the traditional and highly precise DA methods may provide an additional measure to nuclear material accountancy which would potentially result in a more timely, cost-effective and resource efficient means for safeguards verification at such facilities. By monitoring process control measurements (e.g. flowrates, temperatures, or concentrations of reagents, products or wastes), abnormal plant operations can be detected. Pacific Northwest National Laboratory (PNNL) is developing on-line NDA process monitoring technologies, including both the Multi-Isotope Process (MIP) Monitor and a spectroscopy-based monitoring system, to potentially reduce the time and resource burden associated with current techniques. The MIP Monitor uses gamma spectroscopy and multivariate analysis to identify off-normal conditions in process streams. The spectroscopic monitor continuously measures chemical compositions of the process streams including actinide metal ions (U, Pu, Np), selected fission products, and major cold flowsheet chemicals using UV-Vis, Near IR and Raman spectroscopy. This paper will provide an overview of our methods and report our on-going efforts to develop and demonstrate the technologies.

  5. Do the Kepler AGN Light Curves Need Re-processing?

    CERN Document Server

    Kasliwal, Vishal P; Richards, Gordon T; Williams, Joshua; Carini, Michael T

    2015-01-01

    We gauge the impact of spacecraft-induced effects on the inferred variability properties of the light curve of the Seyfert 1 AGN Zw 229-15 observed by \\Kepler. We compare the light curve of Zw 229-15 obtained from the Kepler MAST database with a re-processed light curve constructed from raw pixel data (Williams & Carini, 2015). We use the first-order structure function, $SF(\\delta t)$, to fit both light curves to the damped power-law PSD of Kasliwal, Vogeley & Richards, 2015. On short timescales, we find a steeper log-PSD slope ($\\gamma = 2.90$ to within $10$ percent) for the re-processed light curve as compared to the light curve found on MAST ($\\gamma = 2.65$ to within $10$ percent)---both inconsistent with a damped random walk which requires $\\gamma = 2$. The log-PSD slope inferred for the re-processed light curve is consistent with previous results (Carini & Ryle, 2012, Williams & Carini, 2015) that study the same re-processed light curve. The turnover timescale is almost identical for bot...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-10-01

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

  7. Technical specifications on the welding in fuel reprocessing plants

    Energy Technology Data Exchange (ETDEWEB)

    Karino, Motonobu; Uryu, Mitsuru; Matsui, N.; Nakazawa, Fumio; Imanishi, Makoto; Koizumi; Kazuhiko; Sugawara, Junichi; Tanaka, Hideo

    1999-04-01

    The past specifications SGN of the welding in JNC was reexamined for the reprocessing plants in order to further promote the quality control. The specification first concerns the quality of raw materials, items of the quality tests, material management, and qualification standards of the welders. It extends over details of the welding techniques, welding design, welding testings, inspection and the judgment standards. (H. Baba)

  8. New approaches to reprocessing of oxide nuclear fuel.

    Science.gov (United States)

    Myasoedov, B F; Kulyako, Yu M

    Dissolution of UO2, U3O8, and solid solutions of actinides in UO2 in subacid aqueous solutions (pH 0.9-1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL(-1)). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.

  9. Transmutation Strategy Using Thorium-Reprocessed Fuel ADS for Future Reactors in Vietnam

    Directory of Open Access Journals (Sweden)

    Thanh Mai Vu

    2013-01-01

    Full Text Available Nuclear power is believed to be a key to the energy security for a developing country like Vietnam where the power demanding increases rapidly every year. Nevertheless, spent nuclear fuel from nuclear power plants is the source of radiotoxic and proliferation risk. A conceptual design of ADS utilizing thorium fuel as a based fuel and reprocessed fuel as a seed for nuclear waste transmutation and energy production is proposed as one of the clean, safe, and economical solutions for the problem. In the design, 96 seed assemblies and 84 blanket assemblies were inserted into the core to make a heterogeneous subcritical core configuration. Introducing thorium fuel into the core offers an effective way to transmute plutonium and minor actinide (MA and gain energy from this process. Transmutation rate as a function of burnup is estimated using MCNPX 2.7.0 code. Results show that by using the seed-blanket designed ADS, at 40 GWd/t burnup, 192 kg of plutonium and 156 kg of MA can be eliminated. Equivalently, 1  ADS can be able to transmute the transuranic (TRU waste from 2  LWRs. 14 units of ADS would be required to eliminate TRUs from the future reactors to be constructed in Vietnam.

  10. Adequacy of radioiodine control and monitoring at nuclear fuels reprocessing plants

    Energy Technology Data Exchange (ETDEWEB)

    Scheele, R.D.; Burger, L.L.; Soldat, J.K.

    1984-06-01

    The present backlog of irradiated reactor fuel leads to projections that no fuel out of the reactor less than 10 years need be reprocessed prior to the year 2000. The only radioiodine present in such aged fuel is /sup 129/I (half-life 1.6 x 10/sup 7/ y). The /sup 131/I initially present in the fuel decays to insignificance in the first few hundred days post-reactor. The /sup 129/I content of irradiated fuel is about 1 Ci per gigawatt-year of electricity generated (Ci/GW(e)-y). The US EPA has specified, in 40 CFR 190, a release limit for /sup 129/I of 5 mCi/GW(e)-y. Thus a retention factor (RF) of 200 for /sup 129/I at the fuel reprocessing plant (FRP) is required. Experience indicates that RF values obtained under actual FRP operating conditions can average as little as 10% of experimentally determined RF values. Therefore processes theoretically capable of achieving RF values of up to 10/sup 4/ have been investigated. The US EPA has also specified in 40 CFR 90 a thyroid dose limit of 75 mrem/y for a member of the general public. This dose limit could be readily met at a typical FRP site with an RF value of about 10 or less. Therefore, the limit of 5 mCi/GW(e)-y is more restrictive than the thyroid dose limit for /sup 129/I. The absence of /sup 131/I in effluents from processing of aged fuels makes analysis of /sup 129/I somewhat easier. However, in-line, real-time monitoring for /sup 129/I in FRP gas streams is currently not feasible. Moisture, chemicals, and other radioactive fission products interfere with in-plant measurements. Samples collected over several days must be taken to a laboratory for /sup 129/I analysis. Measurement techniques currently in use or under investigation include neutron activation analysis, scintillation counting, mass spectroscopy, and gas chromatography coupled with electron capture detection. 26 references, 3 figures, 7 tables.

  11. Symposium on the reprocessing of irradiated fuels. Book 3, Session V

    Energy Technology Data Exchange (ETDEWEB)

    None

    1958-12-31

    Book three of this conference has a single-focused session V entitled Engineering and Economics, with 16 papers. The session is concerned with several phases of chemical reprocessing of fuels which are of a general nature. Hot labs, radiochemical analytical facilities, and high level development cells are described. Dissolution equipment, contactors, flow generation, measurement, and control equipment, samplers, connectors, carriers, valves, filters, and hydroclones are described and discussed. Papers are included on: radiation safety, chemical safety, radiochemical plant operating experience in the U.S., and heavy element isotopic buildup. The general economics of solvent extraction processing is discussed, and capital and operating costs for several U. S. plants given. The Atomic Energy Commission's chemical processing programs and administration are evaluated and the services offered and charges therefore are listed.

  12. Reprocessing of spent nuclear fuels. Status and trends; Upparbetning av anvaent kaernbraensle. Laege och trender

    Energy Technology Data Exchange (ETDEWEB)

    Hultgren, Aa.

    1993-01-01

    The report gives a short review of the status for industrial reprocessing and recycling of Uranium/Plutonium. The following countries are covered: Belgium, France, Germany, Great Britain, India, Japan, Russia, USA. Different fuel cycle strategies are accounted for, and new developments outlined. 116 refs, 27 figs, 12 tabs.

  13. Reprocessed and combined thorium fuel cycles in a PER system with a micro heterogeneous approaches

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Fabiana B.A.; Castro, Victor F.; Faria, Rochkhudson B. de; Pereira, Claubia; Fortini, Angela, E-mail: fabianabeghini@yahoo.com.br, E-mail: victorfariacastro@gmail.com, E-mail: rochkdefaria@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br, E-mail: fortini@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2015-07-01

    A micro heterogeneous approaches were used to study the behavior of reprocessed fuel spiked with thorium in a PWR fuel element considering (TRU-Th) cycle. The goal is to achieve a higher burnup using three different configurations to model the fuel element using SCALE 6.0. The reprocessed fuels were obtained using the ORIGEN 2.1 code from a spent PWR standard fuel (33,000 MWd/tHM burned), with 3.1% of initial enrichment. The spent fuel remained in the cooling pool for five years and then reprocessed using the UREX+ technique. Three configurations of micro heterogeneous approaches were analyzed, and the k{sub inf} and plutonium evolution during the burnup were evaluated. The preliminary results show that the behavior of advanced fuel based on transuranic elements spiked with thorium, and micro heterogeneous approach are satisfactory in PWRs, and the configuration that use a combination of Th and TRU (configuration 1) seems to be the most promising once has higher values for k{sub inf} during the burnup, compared with other configurations. (author)

  14. Environmental survey of the reprocessing and waste management portions of the LWR fuel cycle: a task force report

    Energy Technology Data Exchange (ETDEWEB)

    Bishop, W.P.; Miraglia, F.J. Jr. (eds.)

    1976-10-01

    This Supplement deals with the reprocessing and waste management portions of the nuclear fuel cycle for uranium-fueled reactors. The scope of the report is limited to the illumination of fuel reprocessing and waste management activities, and examination of the environmental impacts caused by these activities on a per-reactor basis. The approach is to select one realistic reprocessing and waste management system and to treat it in enough depth to illuminate the issues involved, the technology available, and the relationships of these to the nuclear fuel cycle in general and its environmental impacts.

  15. THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Matthew Bunn; Steve Fetter; John P. Holdren; Bob van der Zwaan

    2003-07-01

    This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recycling to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices.

  16. Methods of Gas Phase Capture of Iodine from Fuel Reprocessing Off-Gas: A Literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Daryl Haefner

    2007-02-01

    A literature survey was conducted to collect information and summarize the methods available to capture iodine from fuel reprocessing off-gases. Techniques were categorized as either wet scrubbing or solid adsorbent methods, and each method was generally described as it might be used under reprocessing conditions. Decontamination factors are quoted only to give a rough indication of the effectiveness of the method. No attempt is made to identify a preferred capture method at this time, although activities are proposed that would provide a consistent baseline that would aid in evaluating technologies.

  17. NO/sub x/ emissions from Hanford nuclear fuels reprocessing plants

    Energy Technology Data Exchange (ETDEWEB)

    Pajunen, A. L.; Dirkes, R. L.

    1978-09-15

    Operation of the existing Hanford nuclear fuel reprocessing facilities will increase the release of nitrogen oxides (NO/sub x/) to the atmosphere over present emission rates. Stack emissions from two reprocessing facilities, one waste storage facility and two coal burning power plants will contain increased concentrations of NO/sub x/. The opacity of the reprocessing facilities' emissions is predicted to periodically exceed the State and local opacity limit of twenty percent. Past measurements failed to detect differences in the ambient air NO/sub x/ concentration with and without reprocessing plant operations. Since the facilities are not presently operating, increases in the non-occupational ambient air NO/sub x/ concentration were predicted from theoretical diffusion models. Based on the calculations, the annual average ambient air NO/sub x/ concentration will increase from the present level of less than 0.004 ppM to less than 0.006 ppM at the Hanford site boundaries. The national standard for the annual mean ambient air NO/sub 2/ concentration is 0.05 ppM. Therefore, the non-occupational ambient air NO/sub x/ concentration will not be increased to significant levels by reprocessing operations in the Hanford 200 Areas.

  18. Pyrochemical reprocessing of molten salt fast reactor fuel: focus on the reductive extraction step

    Directory of Open Access Journals (Sweden)

    Rodrigues Davide

    2015-12-01

    Full Text Available The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile material and to separate the minor actinides from fission products. The reprocessing involves several chemical steps based on redox and acido-basic properties of the various elements contained in the fuel salt. One challenge is to perform a selective extraction of actinides and lanthanides in spent liquid fuel. Extraction of actinides and lanthanides are successively performed by a reductive extraction in liquid bismuth pool containing metallic lithium as a reductive reagent. The objective of this paper is to give a description of the several steps of the reprocessing retained for the molten salt fast reactor (MSFR concept and to present the initial results obtained for the reductive extraction experiments realized in static conditions by contacting LiF-ThF4-UF4-NdF3 with a lab-made Bi-Li pool and for which extraction efficiencies of 0.7% for neodymium and 14.0% for uranium were measured. It was concluded that in static conditions, the extraction is governed by a kinetic limitation and not by the thermodynamic equilibrium.

  19. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

  20. 75 FR 45167 - Notice of Public Workshop on a Potential Rulemaking for Spent Nuclear Fuel Reprocessing Facilities

    Science.gov (United States)

    2010-08-02

    ... civilian nuclear power globally and close the nuclear fuel cycle through reprocessing spent fuel and... Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor... regulations in 10 CFR Part 171, ``Annual Fees for Reactor Licenses and Fuel Cycle Licenses and......

  1. Apparatus and method for reprocessing and separating spent nuclear fuels. [Patent application

    Science.gov (United States)

    Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.

    1982-01-19

    Spent nuclear fuels, including actinide fuels, volatile and non-volatile fission products, are reprocessed and separated in a molten metal solvent housed in the reaction region of a separation vessel which includes a reflux region positioned above the molten tin solvent. The reflux region minimizes loss of evaporated solvent during the separation of the actinide fuels from the volatile fission products. Additionally, inclusion of the reflux region permits the separation of the more volatile fission products (noncondensable) from the less volatile ones (condensable).

  2. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Jeong, C. J.; Yang, M. S

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs.

  3. Development of Online Spectroscopic pH Monitoring for Nuclear Fuel Reprocessing Plants: Weak Acid Schemes.

    Science.gov (United States)

    Casella, Amanda J; Ahlers, Laura R H; Campbell, Emily L; Levitskaia, Tatiana G; Peterson, James M; Smith, Frances N; Bryan, Samuel A

    2015-05-19

    In nuclear fuel reprocessing, separating trivalent minor actinides and lanthanide fission products is extremely challenging and often necessitates tight pH control in TALSPEAK (Trivalent Actinide-Lanthanide Separation by Phosphorus reagent Extraction from Aqueous Komplexes) separations. In TALSPEAK and similar advanced processes, aqueous pH is one of the most important factors governing the partitioning of lanthanides and actinides between an aqueous phase containing a polyaminopolycarboxylate complexing agent and a weak carboxylic acid buffer and an organic phase containing an acidic organophosphorus extractant. Real-time pH monitoring would significantly increase confidence in the separation performance. Our research is focused on developing a general method for online determination of the pH of aqueous solutions through chemometric analysis of Raman spectra. Spectroscopic process-monitoring capabilities, incorporated in a counter-current centrifugal contactor bank, provide a pathway for online, real-time measurement of solution pH. The spectroscopic techniques are process-friendly and can be easily configured for online applications, whereas classic potentiometric pH measurements require frequent calibration/maintenance and have poor long-term stability in aggressive chemical and radiation environments. Raman spectroscopy discriminates between the protonated and deprotonated forms of the carboxylic acid buffer, and the chemometric processing of the Raman spectral data with PLS (partial least-squares) regression provides a means to quantify their respective abundances and therefore determine the solution pH. Interpretive quantitative models have been developed and validated under a range of chemical composition and pH conditions using a lactic acid/lactate buffer system. The developed model was applied to new spectra obtained from online spectral measurements during a solvent extraction experiment using a counter-current centrifugal contactor bank. The model

  4. LMFBR operation in the nuclear cycle without fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, S.I. [Institute of Physics and Power Engineering, Kaluga (Russian Federation)

    1997-12-01

    Substantiation is given to expediency of investigation of nuclear power (NP) development with fast reactors cooled by lead-bismuth alloy operating during extended time in the open nuclear fuel cycle with slightly enriched or depleted uranium make-up. 9 refs., 1 fig., 6 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Clarysson A.M. da; Pereira, Claubia; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Gual, Maritza R., E-mail: clarysson@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: maritzargual@gmail.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2013-07-01

    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)

  6. Available reprocessing and recycling services for research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tozser, Sandor; Marshall, Frances M.; Adelfang, Pablo; Bradley, Edward [International Atomic Energy Agency, Vienna (Austria); Budu, Madalina Elena [SOSNY Research and Development Company, Moscow (Russian Federation); Chiguer, Mustapha [AREVA, Paris La Defense (France)

    2016-03-15

    International activities in the back end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. In the future inventories of LEU SNF will continue to be created and the back end solution of RR SNF remains a critical issue. The IAEA, based on the experience gained during the decade of international cooperation in supporting the objectives of the HEU take-back programmes, drew up a report presenting available reprocessing and recycling services for RR SNF. This paper gives an overview of the report, which will address all aspects of reprocessing and recycling services for RR SNF.

  7. Thorium utilization program progress report for January 1, 1974--June 30, 1975. [Reprocessing; refabrication; recycle fuel irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Kasten, P.R.

    1976-05-01

    Work was carried out on the following: HTGR reprocessing development and pilot plant, refabrication development and pilot plant, recycle fuel irradiations, engineering and economic studies, and conceptual design of a commercial recycle plant. (DLC)

  8. Conservatism in effective dose calculations for accident events involving fuel reprocessing waste tanks.

    Science.gov (United States)

    Bevelacqua, J J

    2011-07-01

    Conservatism in the calculation of the effective dose following an airborne release from an accident involving a fuel reprocessing waste tank is examined. Within the regulatory constraints at the Hanford Site, deterministic effective dose calculations are conservative by at least an order of magnitude. Deterministic calculations should be used with caution in reaching decisions associated with required safety systems and mitigation philosophy related to the accidental release of airborne radioactive material to the environment.

  9. Contaminants of the bismuth phosphate process as signifiers of nuclear reprocessing history.

    Energy Technology Data Exchange (ETDEWEB)

    Schwantes, Jon M.; Sweet, Lucas E.

    2012-10-01

    Reagents used in spent nuclear fuel recycling impart unique contaminant patterns into the product stream of the process. Efforts are underway at Pacific Northwest National Laboratory to characterize and understand the relationship between these patterns and the process that created them. A main challenge to this effort, recycling processes that were employed at the Hanford site from 1944-1989 have been retired for decades. This precludes direct measurements of the contaminant patterns that propagate within product streams of these facilities. In the absence of any operating recycling facilities at Hanford, we have taken a multipronged approach to cataloging contaminants of U.S. reprocessing activities using: (1) historical records summarizing contaminants within the final Pu metal button product of these facilities; (2) samples of opportunity that represent intermediate products of these processes; and (3) lab-scale experiments and model simulations designed to replicate contaminant patterns at each stage of nuclear fuel reprocessing. This report provides a summary of the progress and results from Fiscal Year (April 1, 2010-September 30) 2011.

  10. Technology, safety, and costs of decommissioning a reference nuclear fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Jenkins, C.E.; Rhoads, R.E.

    1977-09-01

    Safety and cost information were developed for the conceptual decommissioning of a fuel reprocessing plant with characteristics similar to the Barnwell Nuclear Fuel Plant. The main process building, spent fuel receiving and storage station, liquid radioactive waste storage tank system, and a conceptual high-level waste-solidification facility were postulated to be decommissioned. The plant was conceptually decommissioned to three decommissioning states or modes; layaway, protective storage, and dismantlement. Assuming favorable work performance, the elapsed time required to perform the decommissioning work in each mode following plant shutdown was estimated to be 2.4 years for layaway, 2.7 years for protective storage, and 5.2 years for dismantlement. In addition to these times, approximately 2 years of planning and preparation are required before plant shutdown. Costs, in constant 1975 dollars, for decommissioning were estimated to be $18 million for layaway, $19 million for protective storage and $58 million for dismantlement. Maintenance and surveillance costs were estimated to be $680,000 per year after layaway and $140,000 per year after protective storage. The combination mode of protective storage followed by dismantlement deferred for 10, 30, and 100 years was estimated to cost $64 million, $67 million and $77 million, respectively, in nondiscounted total 1975 dollars. Present values of these costs give reduced costs as dismantlement is deferred. Safety analyses indicate that radiological and nonradiological safety impacts from decommissioning activities should be small. The 50-year radiation dose commitment to the members of the public from airborne releases from normal decommissioning activities were estimated to be less than 11 man-rem.

  11. Development of On-Line Spectroscopic pH Monitoring for Nuclear Fuel Reprocessing Plants: Weak Acid Schemes

    Energy Technology Data Exchange (ETDEWEB)

    Casella, Amanda J.; Hylden, Laura R.; Campbell, Emily L.; Levitskaia, Tatiana G.; Peterson, James M.; Smith, Frances N.; Bryan, Samuel A.

    2015-05-19

    Knowledge of real-time solution properties and composition is a necessity for any spent nuclear fuel reprocessing method. Metal-ligand speciation in aqueous solutions derived from the dissolved commercial spent fuel is highly dependent upon the acid concentration/pH, which influences extraction efficiency and the resulting speciation in the organic phase. Spectroscopic process monitoring capabilities, incorporated in a counter current centrifugal contactor bank, provide a pathway for on-line real-time measurement of solution pH. The spectroscopic techniques are process-friendly and can be easily configured for on-line applications, while classic potentiometric pH measurements require frequent calibration/maintenance and have poor long-term stability in aggressive chemical and radiation environments. Our research is focused on developing a general method for on-line determination of pH of aqueous solutions through chemometric analysis of Raman spectra. Interpretive quantitative models have been developed and validated under the range of chemical composition and pH using a lactic acid/lactate buffer system. The developed model was applied to spectra obtained on-line during solvent extractions performed in a centrifugal contactor bank. The model predicted the pH within 11% for pH > 2, thus demonstrating that this technique could provide the capability of monitoring pH on-line in applications such as nuclear fuel reprocessing.

  12. Thoria-based nuclear fuels thermophysical and thermodynamic properties, fabrication, reprocessing, and waste management

    CERN Document Server

    Bharadwaj, S R

    2013-01-01

    This book presents the state of the art on thermophysical and thermochemical properties, fabrication methodologies, irradiation behaviours, fuel reprocessing procedures, and aspects of waste management for oxide fuels in general and for thoria-based fuels in particular. The book covers all the essential features involved in the development of and working with nuclear technology. With the help of key databases, many of which were created by the authors, information is presented in the form of tables, figures, schematic diagrams and flow sheets, and photographs. This information will be useful for scientists and engineers working in the nuclear field, particularly for design and simulation, and for establishing the technology. One special feature is the inclusion of the latest information on thoria-based fuels, especially on the use of thorium in power generation, as it has less proliferation potential for nuclear weapons. Given its natural abundance, thorium offers a future alternative to uranium fuels in nuc...

  13. Thermal decomposition of organic solvent with nitric acid in nuclear fuel reprocessing plants

    Energy Technology Data Exchange (ETDEWEB)

    Koike, Tadao; Nishio, Gunji; Takada, Junichi; Tukamoto, Michio; Watanabe, Kouji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Miyata, Sadaichirou

    1995-02-01

    Since a thermal decomposition of organic solvent containing TBP (tributyl phosphate) with nitric acid and heavy metal nitrates is an exothermic reaction, it is possible to cause an explosive decomposition of TBP-complex materials formed by a nitration between the solvent and nitric acid, if the solvent involving TBP-complex is heated upto a thermal limit in an evaporator to concentrate a fuel liquid solution from the extraction process in the reprocessing plant. In JAERI, the demonstration test for explosive decomposition of TBP-complex by the nitration was performed to elucidate the safety margin of the evaporator in the event of hypothetical explosion under auspices of the Science and Technology Agency. The demonstration test was carried out by heating TBP/n-dodecane solvent mixed with nitric acid and uranium nitrate. In the test, the thermal decomposition behavior of the solvent was examined, and also a kinematic reaction constant and a heat formation of the TBP-complex decomposition were measured by the test. In the paper, a safety analysis of a model evaporator was conducted during accidental conditions under the explosive decomposition of the solvent. (author).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-02-01

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

  15. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  16. Novel Sorbent Development and Evaluation for the Capture of Krypton and Xenon from Nuclear Fuel Reprocessing Off-Gas Streams

    Energy Technology Data Exchange (ETDEWEB)

    Troy G. Garn; Mitchell R. Greenhalgh; Jack D. Law

    2013-10-01

    The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbent development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, INL sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up.

  17. Novel Sorbent Development and Evaluation for the Capture of Krypton and Xenon from Nuclear Fuel Reprocessing Off-Gas Streams

    Energy Technology Data Exchange (ETDEWEB)

    Troy G. Garn; Mitchell R. Greenhalgh; Jack D. Law

    2013-09-01

    The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbent development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, INL sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up.

  18. New approaches to reprocessing of oxide nuclear fuel

    OpenAIRE

    Myasoedov, B. F.; Kulyako, Yu. M.

    2012-01-01

    Dissolution of UO2, U3O8, and solid solutions of actinides in UO2 in subacid aqueous solutions (pH 0.9–1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL−1...

  19. Glass ceramics containment matrix for insoluble residues coming from spent fuel reprocessing

    Science.gov (United States)

    Pinet, O.; Boën, R.

    2014-04-01

    Spent fuel reprocessing by hydrometallurgical process generates insoluble residues waste streams called fines solution. Considering their radioactivity, fines solution could be considered as Intermediate Level Waste. This waste stream is usually mixed with fission products stream before vitrification. Thus fines are incorporated in glass matrix designed for High Level Waste. The withdrawal of fines from high level glass could decrease the volume of high level waste after conditioning. It could also decrease the reaction time between high level waste and additives to obtain a homogeneous melt and then increase the vitrification process capacity. Separated conditioning of fines in glass matrices has been tested. The fines content targeted value is 16 wt%. To achieve this objective, two types of glass ceramic formulations have been tested. 700 g of the two selected glass ceramics have been prepared using simulated fines. Additives used were ground glass. Melting is achieved at 1100 °C. According to the type of glass ceramic, reducing or oxidizing conditions have been performed during melting. Due to their composition and the melting redox conditions, different phases have been observed. These crystalline phases are typically RuO2, metallic Ru, metallic Pd, MoO2 and CaMoO4. In view of melting these matrices in an in can process the corrosiveness of one of the most oxidizing borosilicate glass ceramic formulation has been tested. This one has been remelted at 1100 °C in inconel 601 pot for 3 days. The oxygen fugacity measurement performed in the remelted glass leads to an oxidizing value, indicating that no significant reaction occurred between the inconel pot and the glass melt had occurred.

  20. Assessment of sensitivity of neutron-physical parameters of fast neutron reactor to purification of reprocessed fuel from minor actinides

    Science.gov (United States)

    Cherny, V. A.; Kochetkov, L. A.; Nevinitsa, A. I.

    2013-12-01

    The work is devoted to computational investigation of the dependence of basic physical parameters of fast neutron reactors on the degree of purification of plutonium from minor actinides obtained as a result of pyroelectrochemical reprocessing of spent nuclear fuel and used for manufacturing MOX fuel to be reloaded into the reactors mentioned. The investigations have shown that, in order to preserve such important parameters of a BN-800 type reactor as the criticality, the sodium void reactivity effect, the Doppler effect, and the efficiency of safety rods, it is possible to use the reprocessed fuel without separation of minor actinides for refueling (recharging) the core.

  1. Atmospheric dispersal of [sup 129]iodine from nuclear fuel reprocessing facilities

    Energy Technology Data Exchange (ETDEWEB)

    Moran, J.E.; Schink, D.R. (Texas A and M Univ., College Station, TX (United States). Dept. of Oceanography); Oktay, S.; Santschi, P.H. (Texas A and M Univ., Galveston, TX (United States). Dept. of Oceanography)

    1999-08-01

    [sup 129]I/[sup 127]I ratios measured in meteoric water and epiphytes from the continental United States are higher than those measured in coastal seawater or surface freshwater and suggest long-range atmospheric transport of [sup 129]I from the main source for the earth's surface inventory, viz., nuclear fuel reprocessing facilities. The median ratio for 14 meteoric water samples is 2100 [times] 10[sup [minus]12], corresponding to a [sup 129]I concentration of 2.5 [times] 10[sup 7] atoms/L, whereas 9 epiphyte samples have a median ratio of 1800 [times] 10[sup [minus]12]. Calculated deposition rates of [sup 129]I in the continental United States reveal that a small but significant fraction of the atmospheric releases from the nuclear fuel reprocessing facilities at Sellafield, England, and Cap de La Hague, France, is deposited after distribution by long-range transport. The inferred dominant mode of transport is easterly, within the troposphere, mainly in the form of the organic gas methyl iodide.

  2. Calibration of burnup monitor of spent nuclear fuel installed at Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Matoba, Masaru; Wakabayashi, Genichiro [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering; Naito, Hirofumi; Hirota, Masanari [Nuclear Fuel Industries Ltd., Tokyo (Japan); Morizaki, Hidetoshi; Kumanomido, Hironori; Natsume, Koichiro [Toshiba Corp., Tokyo (Japan)

    2001-05-01

    The spent nuclear fuel storage pool of Rokkasho reprocessing plant adopts the burnup credit' conception. Spent fuel assemblies are measured every one by one, by burnup monitors, and stored to a storage rack which is designed with specified residual enrichment. For nuclear criticality control, it is necessary for the burnup monitor that the measured value includes a kind of margin, which consists of errors of the monitor. In this paper, we describe the error of the burnup monitors, and the way of taking of the margin. From the result of calibration of the burnup monitor carried out from July through November, 1999, we describe that the way of taking of the margin is validated. And comments about possibility of error reduction are remarked. (author)

  3. Fast and Simultaneous Determination of Pu(Ⅳ) and Nitric Acid in Spent Nuclear Fuel Reprocessing Sample by Near Infrared Spectroscopy

    Institute of Scientific and Technical Information of China (English)

    LI; Ding-ming; ZHANG; Li-hua; WANG; Ling; GONG; Yan-ping; FAN; De-jun; YI; Bao-shan; CHEN; Qiang; JI; Yong-chao; WU; Ji-zong

    2013-01-01

    Determination of Pu(Ⅳ)and nitric acid plays significant role in nuclear fuel reprocessing plant to control process accurately and timely.Coupling C-T fixed-type grating with InGaAs detector,a new novel analytical system for simultaneous measurement of nitric acid and Pu(Ⅳ)was developed by our working group.After obtaining near infrared absorptive spectra by the spectroscopic instrument,the spectra data

  4. Status of nuclear fuel reprocessing, spent fuel storage, and high-level waste disposal. Overview and summary

    Energy Technology Data Exchange (ETDEWEB)

    Varanini, E.E. III; Maullin, R.L.

    1978-01-11

    With regard to the specific question embodied in California's nuclear statutes about the demonstrated and approved permanent terminal disposal of nuclear waste (assuming that the reprocessing question is now most for legislative purposes), the finding of the Energy Commission is that such a technology has not been demonstrated and that it is even questionable to assume that one will be demonstrated before the mid 1980s. Following upon this finding and addressing the broader question of continued implementation of the policy expressed by the nuclear fuel cycle statutes, the evidence indicates that it is not prudent to continue siting nuclear powerplants based on an optimistic assumption that waste management technologies to handle nuclear waste will be developed and scientifically demonstrated. The California Legislature has questioned that optimistic assumption by placing the burden of proof on the developers of a demonstrated, scientifically tested process for the permanent and terminal disposal of nuclear wastes. Such a process does not exist at this time. There are many who are optimistic that the development of such a technology will become a reality in the near future. This overview and the supporting report indicate that this optimism is not warranted. Weapons proliferation and degradation of the biosphere by radioactive waste have proved to be unanticipated, difficult and possibly intractable problems in spite of an overriding confidence that nuclear technology would not present such problems. On the basis of the evidence received by this Commission, there are substantial scientific gaps which preclude proceeding on the basis of faith that all the attendant risks and issues will be resolved.

  5. Initiating events study of the first extraction cycle process in a model reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Renze; Zhang, Jian Gang; Zhuang, Dajie; Feng, Zong Yang [China Institute for Radiation Protection, Taiyuan (China)

    2016-06-15

    Definition and grouping of initiating events (IEs) are important basics for probabilistic safety assessment (PSA). An IE in a spent fuel reprocessing plant (SFRP) is an event that probably leads to the release of dangerous material to jeopardize workers, public and environment. The main difference between SFRPs and nuclear power plants (NPPs) is that hazard materials spread diffusely in a SFRP and radioactive material is just one kind of hazard material. Since the research on IEs for NPPs is in-depth around the world, there are several general methods to identify IEs: reference of lists in existence, review of experience feedback, qualitative analysis method, and deductive analysis method. While failure mode and effect analysis (FMEA) is an important qualitative analysis method, master logic diagram (MLD) method is the deductive analysis method. IE identification in SFRPs should be consulted with the experience of NPPs, however the differences between SFRPs and NPPs should be considered seriously. The plutonium uranium reduction extraction (Purex) process is adopted by the technics in a model reprocessing plant. The first extraction cycle (FEC) is the pivotal process in the Purex process. Whether the FEC can function safely and steadily would directly influence the production process of the whole plant-production quality. Important facilities of the FEC are installed in the equipment cells (ECs). In this work, IEs in the FEC process were identified and categorized by FMEA and MLD two methods, based on the fact that ECs are containments in the plant. The results show that only two ECs in the FEC do not need to be concerned particularly with safety problems, and criticality, fire and red oil explosion are IEs which should be emphatically analyzed. The results are accordant with the references.

  6. Task 5c: measurement and instrumentation under subsystem design of the LLL safeguard material control program. [For fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    1976-12-31

    A product survey was conducted of all security products currently available on the market. Documentation is presented of the survey and a printout of the data is included. A general description is given of new but recommended instrumentation and security devices for application to fuel reprocessing plants. Security systems and hardware recommended for development, assembly, and testing are discussed briefly. (DLC)

  7. Review Of Decommissioning Experience In Spent Fuel Reprocessing Facilities at Marcoule

    Energy Technology Data Exchange (ETDEWEB)

    Guiberteau, Ph.; Vendroux, M. [CODEM GIE, BP 21004, 30201 Bagnols sur Ceze cedex (France); Berlan, C. [COGEMA Reprocessing Business Unit, 2, rue Paul Dautier - BP.4, 78141 Velizy Cedex (France)

    2003-07-01

    Final shutdown and decontamination, dismantling, and legacy waste retrieval programs are currently in progress at the Marcoule nuclear fuel reprocessing plant in southern France. They began in 1998 and will continue until about 2040. CODEM is the funding, decision-making and inspection organization for these decommissioning operations, COGEMA is the nuclear operator and the industrial contractor. After an overview of the facilities, the project and the participants, most significant operations are discussed in greater detail. High activity levels and the presence of large quantities of {alpha}-emitters complicate operating and waste treatment conditions. The major issues impacting cost-effectiveness-scenario, waste removal and project organization will be highlighted in the conclusion.

  8. 高温氧化挥发法--一种先进乏燃料后处理的首端工艺技术%An Advanced Head-end Process for Reprocessing of Spent Fuel by High Temperature Vol-oxidation Treatment

    Institute of Scientific and Technical Information of China (English)

    李辉波; 何辉; 叶国安; 苏哲

    2015-01-01

    High temperature vol‐oxidation treatment technology is a dry head‐end process used for decladding ,oxidation of spent fuel ,and the removal of 3 H ,85 Kr/Xe , 14C ,129I and Cs from fuel prior to main spent fuel treatment process which would remove most of the volatile nuclides before fuel dissolution and cause rapid fuel dissolu‐tion in HNO3 .If effective ,this process would be an effective way for sharp reduction in the volume of liquid radioactive waste and localization fission nuclides (such as tritium and iodine) manage by vol‐oxidation in the head‐end process .In this paper ,the key influence factors (such as temperature ,oxidizing atmosphere and so on ) and applica‐tions characteristics in the head‐end process for reprocessing of spent fuel by vol‐oxidation treatment are expounded synthetically .%高温氧化挥发处理技术是乏燃料后处理的干法首端过程,其目的是在乏燃料后处理分离工艺前实现包壳与燃料芯块分离,燃料氧化和裂变产物3 H、85 Kr/Xe、14 C、129 I、Cs的去除。此过程既有利于乏燃料元件的溶解,又有利于在乏燃料元件进入溶解工艺之前实现氚碘等裂变元素去除,是实现整个乏燃料后处理流程过程废液最小化和氚碘等裂变产物集中管理的最有效方法之一。本文针对氧化挥发技术在乏燃料后处理首端中的应用特点以及氧化温度、气氛等关键影响因素进行了综合分析和阐述。

  9. Model of iodine transport and reaction kinetics in a nuclear fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Davis, W. Jr.

    1977-05-01

    A model is presented to describe the time-dependent flow and retention of stable iodine isotopes and the decay of /sup 131/I in a nuclear fuel reprocessing plant. The plant consists of 16 units of equipment such as a voloxidizer or graphite burner, fuel dissolver, solvent extractors, storage tanks, vaporizers, primary iodine sorbers, and silver zeolite. The rate of accumulation of bulk and radioactive iodine in these units and in the environment is described using 19 differential equations. Reasonable time-dependence of iodine retention factors (RFs) by the plant were calculated. RFs for a new plant in excess of 10/sup 6/ for stable iodine and /sup 129/I decrease to the range of 10/sup 3/ to 10/sup 2/ as plant operating times exceed 50 to 100 days. The RFs for /sup 131/I also decrease initially, for a period of approximately 10 days, but then increase by several orders of magnitude due to radioactive decay and isotopic exchange. Generally, the RFs for /sup 131/I exceed those for stable iodine by factors of 10/sup 4/ or more. 19 references, 13 figures, 2 tables. (DLC)

  10. The multi-isotope process monitor: Non-destructive, near-real-time nuclear safeguards monitoring at a reprocessing facility

    Science.gov (United States)

    Orton, Christopher Robert

    The IAEA will require advanced technologies to effectively safeguard nuclear material at envisioned large scale nuclear reprocessing plants. This dissertation describes results from simulations and experiments designed to test the Multi-Isotope Process (MIP) Monitor, a novel safeguards approach for process monitoring in reprocessing plants. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams, nondestructively and in near-real time (NRT). Three different models were used to predict spent nuclear fuel composition, estimate chemical distribution during separation, and simulate spectra from a variety of gamma detectors in product and raffinate streams for processed fuel. This was done for fuel with various irradiation histories and under a variety of plant operating conditions. Experiments were performed to validate the results from the model. Three segments of commercial spent nuclear fuel with variations in burnup and cooling time were dissolved and subjected to a batch PUREX method to separate the uranium and plutonium from fission and activation products. Gamma spectra were recorded by high purity germanium (HPGe) and cadmium zinc telluride (CZT) detectors. Hierarchal Cluster Analysis (HCA) and Principal Component Analysis (PCA) were applied to spectra from both model and experiment to investigate spectral variations as a function of acid concentration, burnup level and cooling time. Partial Least Squares was utilized to extract quantitative information about process variables, such as acid concentration or burnup. The MIP Monitor was found to be sensitive to the induced variations of the process and was capable of extracting quantitative process information from the analyzed spectra.

  11. Materials accounting in a fast-breeder-reactor fuels-reprocessing facility: optimal allocation of measurement uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Dayem, H.A.; Ostenak, C.A.; Gutmacher, R.G.; Kern, E.A.; Markin, J.T.; Martinez, D.P.; Thomas, C.C. Jr.

    1982-07-01

    This report describes the conceptual design of a materials accounting system for the feed preparation and chemical separations processes of a fast breeder reactor spent-fuel reprocessing facility. For the proposed accounting system, optimization techniques are used to calculate instrument measurement uncertainties that meet four different accounting performance goals while minimizing the total development cost of instrument systems. We identify instruments that require development to meet performance goals and measurement uncertainty components that dominate the materials balance variance. Materials accounting in the feed preparation process is complicated by large in-process inventories and spent-fuel assembly inputs that are difficult to measure. To meet 8 kg of plutonium abrupt and 40 kg of plutonium protracted loss-detection goals, materials accounting in the chemical separations process requires: process tank volume and concentration measurements having a precision less than or equal to 1%; accountability and plutonium sample tank volume measurements having a precision less than or equal to 0.3%, a shortterm correlated error less than or equal to 0.04%, and a long-term correlated error less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having a precision less than or equal to 0.4%, a short-term correlated error less than or equal to 0.1%, and a long-term correlated error less than or equal to 0.05%. The effects of process design on materials accounting are identified. Major areas of concern include the voloxidizer, the continuous dissolver, and the accountability tank.

  12. Materials accounting in a fast-breeder-reactor fuels-reprocessing facility: optimal allocation of measurement uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Dayem, H.A.; Ostenak, C.A.; Gutmacher, R.G.; Kern, E.A.; Markin, J.T.; Martinez, D.P.; Thomas, C.C. Jr.

    1982-07-01

    This report describes the conceptual design of a materials accounting system for the feed preparation and chemical separations processes of a fast breeder reactor spent-fuel reprocessing facility. For the proposed accounting system, optimization techniques are used to calculate instrument measurement uncertainties that meet four different accounting performance goals while minimizing the total development cost of instrument systems. We identify instruments that require development to meet performance goals and measurement uncertainty components that dominate the materials balance variance. Materials accounting in the feed preparation process is complicated by large in-process inventories and spent-fuel assembly inputs that are difficult to measure. To meet 8 kg of plutonium abrupt and 40 kg of plutonium protracted loss-detection goals, materials accounting in the chemical separations process requires: process tank volume and concentration measurements having a precision less than or equal to 1%; accountability and plutonium sample tank volume measurements having a precision less than or equal to 0.3%, a shortterm correlated error less than or equal to 0.04%, and a long-term correlated error less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having a precision less than or equal to 0.4%, a short-term correlated error less than or equal to 0.1%, and a long-term correlated error less than or equal to 0.05%. The effects of process design on materials accounting are identified. Major areas of concern include the voloxidizer, the continuous dissolver, and the accountability tank.

  13. Costs of head-end incineration with respect to Kr separation in the reprocessing of HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Barnert-Wiemer, H.; Boehnert, R.

    1976-07-15

    The C-incinerations and the Kr-separations during head-end incineration in the reprocessing of HTR fuel elements are described. The costs for constructing an operating a head-end incineration of reprocessing capacities with 5,000 to 50,000 MW(e)-HTR power have been determined. The cost estimates are divided into investment and operating costs, further after the fraction of the N/sub 2/-content in the incineration exhaust gas, which strongly affects costs. It appears that, in the case of Kr-separation from the incineration exhaust gas, the investment costs as well as the operating costs of the head-end for N/sub 2/-containing exhaust gas are considerably greater than those for gas without N/sub 2/. The C-incineration of the graphite of the HTR fuel elements should therefore only be performed with influx gas that is free of N/sub 2/.

  14. Monitoring, Controlling and Safeguarding Radiochemical Streams at Spent Fuel Reprocessing Facilities, Part 2: Gamma-Ray Spectroscopic Methods

    Energy Technology Data Exchange (ETDEWEB)

    Schwantes, Jon M.; Bryan, Samuel A.; Orton, Christopher R.; Levitskaia, Tatiana G.; Fraga, Carlos G.

    2012-02-10

    The International Atomic Energy Agency (IAEA) has established international safeguards standards for fissionable material at spent fuel reprocessing plants to ensure that significant quantities of weapons-useable nuclear material are not diverted from these facilities. For large throughput nuclear facilities, it is difficult to satisfy the IAEA safeguards accountancy goal for detection of abrupt diversion. Currently, methods to verify material control and accountancy (MC&A) at these facilities require time-consuming and resource-intensive destructive assay (DA). Leveraging new on-line non-destructive assay (NDA) process monitoring techniques in conjunction with the traditional and highly precise DA methods may provide an additional measure to nuclear material accountancy which would potentially result in a more timely, cost-effective and resource efficient means for safeguards verification at such facilities. By monitoring process control measurements (e.g. flowrates, temperatures, or concentrations of reagents, products or wastes), abnormal plant operations can be detected. Pacific Northwest National Laboratory (PNNL) is developing on-line NDA process monitoring technologies based upon gamma-ray and optical spectroscopic measurements to potentially reduce the time and resource burden associated with current techniques. The Multi-Isotope Process (MIP) Monitor uses gamma spectroscopy and multivariate analysis to identify off-normal conditions in process streams. The spectroscopic monitor continuously measures chemical compositions of the process streams including actinide metal ions (U, Pu, Np), selected fission products, and major stable flowsheet reagents using UV-Vis, Near IR and Raman spectroscopy. Multi-variate analysis is also applied to the optical measurements in order to quantify concentrations of analytes of interest within a complex array of radiochemical streams. This paper will provide an overview of these methods and reports on-going efforts to develop

  15. Seismic analysis of the Nuclear Fuel Service Reprocessing Plant at West Valley, New York: documentation

    Energy Technology Data Exchange (ETDEWEB)

    Murray, R.C.; Nelson, T.A.; Davito, A.M.

    1977-04-26

    This material was generated as part of a seismic case review of the NFS Reprocessing Plant. This study is documented in UCRL-52266. The material is divided into two parts: mathematical model information, and ultimate load calculations and comparisons. (DLC)

  16. Specialist MTR reprocessing at Dounreay

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, A. J.; Skea, D. C. J. (UKAEA, Dounreay (United Kingdom))

    1999-12-15

    A summary is provided of the facilities at Dounreay and goes on to describe the plans to adapt an existing facility to reprocess irradiated TRIGA fuel. These facilities will provide a treatment for the fuel, thus enabling reactor operators to pursue their programme of decommissioning. The main features of the processing route are receipt, storage, dismantling and chemical treatment by solvent extraction. Solvent extraction will be on a small scale using improved plant containment and replaceable modular equipment. An outline process flowsheet is described. Wastes produced by the process will pass through established routes, with medium active liquor being stored in the short term and ultimately cemented. The modifications to the facilities will allow the reprocessing of other 'exotic' fuel types to produce waste forms suitable for disposal. (orig.)

  17. Monitoring, Controlling and Safeguarding Radiochemical Streams at Spent Fuel Reprocessing Facilities with Optical and Gamma-Ray Spectroscopic Methods

    Energy Technology Data Exchange (ETDEWEB)

    Schwantes, Jon M.; Bryan, Samuel A.; Orton, Christopher R.; Levitskaia, Tatiana G.; Fraga, Carlos G.

    2012-11-06

    The International Atomic Energy Agency (IAEA) has established international safeguards standards for fissionable material at spent fuel reprocessing plants to ensure that significant quantities of weapons-useable nuclear material are not diverted from these facilities. For large throughput nuclear facilities, it is difficult to satisfy the IAEA safeguards accountancy goal for detection of abrupt diversion. Currently, methods to verify material control and accountancy (MC&A) at these facilities require time-consuming and resourceintensive destructive assay (DA). Leveraging new on-line non-destructive assay (NDA) process monitoring techniques in conjunction with the traditional and highly precise DA methods may provide an additional measure to nuclear material accountancy which would potentially result in a more timely, cost-effective and resource efficient means for safeguards verification at such facilities. By monitoring process control measurements (e.g. flowrates, temperatures, or concentrations of reagents, products or wastes), abnormal plant operations can be detected. Pacific Northwest National Laboratory (PNNL) is developing on-line NDA process monitoring technologies based upon gamma-ray and optical spectroscopic measurements to potentially reduce the time and resource burden associated with current techniques. The Multi-Isotope Process (MIP) Monitor uses gamma spectroscopy and multivariate analysis to identify offnormal conditions in process streams. The spectroscopic monitor continuously measures chemical compositions of the process streams including actinide metal ions (U, Pu, Np), selected fission products, and major stable flowsheet reagents using UV-Vis, Near IR and Raman spectroscopy. Multi-variate analysis is also applied to the optical measurements in order to quantify concentrations of analytes of interest within a complex array of radiochemical streams. This paper will provide an overview of these methods and reports on-going efforts to develop

  18. Waste management system alternatives for treatment of wastes from spent fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    McKee, R.W.; Swanson, J.L.; Daling, P.M.; Clark, L.L.; Craig, R.A.; Nesbitt, J.F.; McCarthy, D.; Franklin, A.L.; Hazelton, R.F.; Lundgren, R.A.

    1986-09-01

    This study was performed to help identify a preferred TRU waste treatment alternative for reprocessing wastes with respect to waste form performance in a geologic repository, near-term waste management system risks, and minimum waste management system costs. The results were intended for use in developing TRU waste acceptance requirements that may be needed to meet regulatory requirements for disposal of TRU wastes in a geologic repository. The waste management system components included in this analysis are waste treatment and packaging, transportation, and disposal. The major features of the TRU waste treatment alternatives examined here include: (1) packaging (as-produced) without treatment (PWOT); (2) compaction of hulls and other compactable wastes; (3) incineration of combustibles with cementation of the ash plus compaction of hulls and filters; (4) melting of hulls and failed equipment plus incineration of combustibles with vitrification of the ash along with the HLW; (5a) decontamination of hulls and failed equipment to produce LLW plus incineration and incorporation of ash and other inert wastes into HLW glass; and (5b) variation of this fifth treatment alternative in which the incineration ash is incorporated into a separate TRU waste glass. The six alternative processing system concepts provide progressively increasing levels of TRU waste consolidation and TRU waste form integrity. Vitrification of HLW and intermediate-level liquid wastes (ILLW) was assumed in all cases.

  19. Safety demonstration tests on pressure rise in ventilation system and blower integrity of a fuel-reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Junichi; Suzuki, Motoe; Tsukamoto, Michio; Koike, Tadao; Nishio, Gunji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-12-01

    In JAERI, the demonstration test was carried out as a part of safety researches of the fuel-reprocessing plant using a large-scale facility consist of cells, ducts, dumpers, HEPA filters and a blower, when an explosive burning due to a rapid reaction of thermal decomposition for solvent/nitric acid occurs in a cell of the reprocessing plant. In the demonstration test, pressure response propagating through the facility was measured under a blowing of air from a pressurized tank into the cell in the facility to elucidate an influence of pressure rise in the ventilation system. Consequently, effective pressure decrease in the facility was given by a configuration of cells and ducts in the facility. In the test, transient responses of HEPA filters and the blower by the blowing of air were also measured to confirm the integrity. So that, it is confirmed that HEPA filters and the blower under pressure loading were sufficient to maintain the integrity. The content described in this report will contribute to safety assessment of the ventilation system in the event of explosive burning in the reprocessing plant. (author)

  20. Nondestructive, energy-dispersive, x-ray fluorescence analysis of actinide stream concentrations from reprocessed nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Camp, D.C.; Ruhter, W.D.

    1979-06-27

    In one plan for reprocessing LWR spent fuel, after separation from fission products and transplutonics, part of the U and all of the Pu in a nitrate solution will form a coprocessed stream which is then evaporated and sent to a hold tank for accounting. The remaining U fraction will be purified and sent to a separate storage tank. These two streams can be monitored using x-ray fluorescence analysis. This report discusses equipment, spectra, cell calibration, and dynamic concentration measurements. 7 figures. (DLC)

  1. Characterization and simulation of soft gamma-ray mirrors for their use with spent fuel rods at reprocessing facilities.

    Science.gov (United States)

    Ruz, J; Descalle, M A; Alameda, J B; Brejnholt, N F; Chichester, D L; Decker, T A; Fernandez-Perea, M; Hill, R M; Kisner, R A; Melin, A M; Patton, B W; Soufli, R; Trellue, H; Watson, S M; Ziock, K P; Pivovaroff, M J

    2016-06-01

    The use of a grazing incidence optic to selectively reflect K-shell fluorescence emission and isotope-specific lines from special nuclear materials is a highly desirable nondestructive analysis method for use in reprocessing fuel environments. Preliminary measurements have been performed, and a simulation suite has been developed to give insight into the design of the x ray optics system as a function of the source emission, multilayer coating characteristics, and general experimental configurations. The experimental results are compared to the predictions from our simulation toolkit to illustrate the ray-tracing capability and explore the effect of modified optics in future measurement campaigns.

  2. PRELIMINARY STUDY OF CERAMICS FOR IMMOBILIZATION OF ADVANCED FUEL CYCLE REPROCESSING WASTES

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.; Billings, A.; Brinkman, K.; Marra, J.

    2010-09-22

    The Savannah River National Laboratory (SRNL) developed a series of ceramic waste forms for the immobilization of Cesium/Lanthanide (CS/LN) and Cesium/Lanthanide/Transition Metal (CS/LN/TM) waste streams anticipated to result from nuclear fuel reprocessing. Simple raw materials, including Al{sub 2}O{sub 3}, CaO, and TiO{sub 2} were combined with simulated waste components to produce multiphase ceramics containing hollandite-type phases, perovskites (particularly BaTiO{sub 3}), pyrochlores, zirconolite, and other minor metal titanate phases. Identification of excess Al{sub 2}O{sub 3} via X-ray Diffraction (XRD) and Scanning Electron Microscopy with Energy Dispersive Spectroscopy (SEM/EDS) in the first series of compositions led to a Phase II study, with significantly reduced Al{sub 2}O{sub 3} concentrations and increased waste loadings. Three fabrication methodologies were used, including melting and crystallizing, pressing and sintering, and Spark Plasma Sintering (SPS), with the intent of studying phase evolution under various sintering conditions. XRD and SEM/EDS results showed that the partitioning of the waste elements in the sintered materials was very similar, despite varying stoichiometry of the phases formed. The Phase II compositions generally contained a reduced amount of unreacted Al{sub 2}O{sub 3} as identified by XRD, and had phase assemblages that were closer to the initial targets. Chemical composition measurements showed no significant issues with meeting the target compositions. However, volatilization of Cs and Mo was identified, particularly during melting, since sintering of the pressed pellets and SPS were performed at lower temperatures. Partitioning of some of the waste components was difficult to determine via XRD. SEM/EDS mapping showed that those elements, which were generally present in small concentrations, were well distributed throughout the waste forms. Initial studies of radiation damage tolerance using ion beam irradiation at Los

  3. Fuels Processing Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — NETL’s Fuels Processing Laboratory in Morgantown, WV, provides researchers with the equipment they need to thoroughly explore the catalytic issues associated with...

  4. Glutarimidedioxime. A complexing and reducing reagent for plutonium recovery from spent nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Xian, Liang [China Institute of Atomic Energy, Beijing (China). Radiochemistry Dept.; Tian, Guoxin [China Institute of Atomic Energy, Beijing (China). Radiochemistry Dept.; Lawrence Berkeley National Laboratory, Berkeley, CA (United States). Chemical Sciences Div.; Beavers, Christine M.; Teat, Simon J. [Lawrence Berkeley National Laboratory, Berkeley, CA (United States). Advanced Light Source; Shuh, David K. [Lawrence Berkeley National Laboratory, Berkeley, CA (United States). Chemical Sciences Div.

    2016-04-04

    Efficient separation processes for recovering uranium and plutonium from spent nuclear fuel are essential to the development of advanced nuclear fuel cycles. The performance characteristics of a new salt-free complexing and reducing reagent, glutarimidedioxime (H{sub 2}A), are reported for recovering plutonium in a PUREX process. With a phase ratio of organic to aqueous of up to 10:1, plutonium can be effectively stripped from 30 % tributyl phosphate (TBP) in kerosene into 1M HNO{sub 3} with H{sub 2}A. The complexation-reduction mechanism is illustrated with the combination of UV/Vis absorption spectra and the crystal structure of a Pu{sup IV} complex with the reagent. The fast stripping rate and the high efficiency for stripping Pu{sup IV}, through the complexation-reduction mechanism, is suitable for use in centrifugal contactors with very short contact/resident times, thereby offering significant advantages over conventional processes.

  5. Considerations affecting deep-well disposal of tritium-bearing low-level aqueous waste from nuclear fuel reprocessing plants

    Energy Technology Data Exchange (ETDEWEB)

    Trevorrow, L. E.; Warner, D. L.; Steindler, M. J.

    1977-03-01

    Present concepts of disposal of low-level aqueous wastes (LLAW) that contain much of the fission-product tritium from light water reactors involve dispersal to the atmosphere or to surface streams at fuel reprocessing plants. These concepts have been challenged in recent years. Deep-well injection of low-level aqueous wastes, an alternative to biospheric dispersal, is the subject of this presentation. Many factors must be considered in assessing its feasibility, including technology, costs, environmental impact, legal and regulatory constraints, and siting. Examination of these factors indicates that the technology of deep-well injection, extensively developed for other industrial wastes, would require little innovation before application to low-level aqueous wastes. Costs would be low, of the order of magnitude of 10/sup -4/ mill/kWh. The environmental impact of normal deep-well disposal would be small, compared with dispersal to the atmosphere or to surface streams; abnormal operation would not be expected to produce catastrophic results. Geologically suitable sites are abundant in the U.S., but a well would best be co-located with the fuel-reprocessing plant where the LLAW is produced. Legal and regulatory constraints now being developed will be the most important determinants of the feasibility of applying the method.

  6. Audit of fuel processing restoration property

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-01

    In April, 1992, due to a diminished need for reprocessed uranium, the Secretary of Energy terminated the Fuel Processing Restoration (FPR) project. The termination left management and operating (M&O) contractors at the Idaho National Engineering Laboratory (Laboratory) with over $54 million in tools, equipment and material to be retained, utilized or disposed of. The objectives of the audit were to determine whether FPR property was adequately accounted for and whether the property was properly redistributed or excessed when the FPR project was terminated.

  7. Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

    2017-11-01

    This paper provides an overview of research evaluating the use of tellurite glass as a waste form for salt wastes from electrochemical processing. The capacities to immobilize different salts were evaluated including: a LiCl-Li2O oxide reduction salt (for oxide fuel) containing fission products, a LiCl-KCl eutectic salt (for metallic fuel) containing fission products, and SrCl2. Physical and chemical properties of the glasses were characterized by using X-ray diffraction, bulk density measurements, chemical durability tests, scanning electron microscopy, and energy dispersive X-ray emission spectroscopy. These glasses were found to accommodate high concentrations of halide salts and have high densities. However, improvements are needed to meet chemical durability requirements.

  8. World-wide redistribution of 129Iodine from nuclear fuel reprocessing facilities:results from meteoric, river, and seawater tracer studies

    Energy Technology Data Exchange (ETDEWEB)

    Fehn, U; Moran, J E; Oktay, S; Santschi, P H; Schink, D R; Snyder, G

    1998-10-02

    Releases of the long-lived radioisotope of iodine, 129I from commercial nuclear fuel reprocessing facilities in England and France have surpassed natural, and even bomb test inventories. 129I/127I ratios measured in a variety of environmental matrices from Europe, North America and the southern hemisphere show the influence of fuel reprocessing-derived 129I, which is transported globally via the atmosphere. Transport and cycling of I and 129I in the hydrosphere and in soils are described based on a spatial survey of 129I in freshwater.

  9. Technological study of electrochemical uranium fuel reprocessing in fused chloride bath; Estudo tecnologico do reprocessamento eletroquimico de combustiveis de uranio em meio de cloretos fundidos

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, Damaris

    2002-07-01

    This study is applied to metallic fuels recycling, concerning advanced reactor concept, which was proposed and tested in LMR type reactors. Conditions for electrochemical non-irradiated uranium fuel reprocessing in fused chloride bath in laboratory scale were established. Experimental procedures and parameters for dehydration treatment of LiCl-KCl eutectic mixture and for electrochemical study of U{sup 3+}/U system in LiCl-KCl were developed and optimized. In the voltammetric studies many working electrodes were tested. As auxiliary electrodes, graphite and stainless steels crucibles were verified, with no significant impurities inclusions in the system. Ag/AgCl in Al{sub 2}O{sub 3} with 1 w% in AgCl were used as reference electrode. The experimental set up developed for electrolyte treatment as well as for the study of the system U{sup 3+}/U in LiCl-KCl showed to be adequate and efficient. Thermogravimetric Techniques, Scanning Electron Microscopy with Energy Dispersive X-Ray Spectrometry and cyclic voltametry showed an efficient dehydration method by using HCl gas and than argon flux for 12 h. Scanning Electron Microscopy, with Energy Dispersive X-Ray Spectrometry and Inductively Coupled Plasma Emission Spectrometry and DC Arc Emission Spectrometry detected the presence of uranium in the cadmium phase. X-ray Diffraction and also Inductively Coupled Plasma Emission Spectrometry and DC Arc Emission Spectrometry were used for uranium detection in the salt phase. The obtained results for the system U{sup 3+}/U in LiCl-KCl showed the viability of the electrochemical reprocessing process based on the IFR advanced fuel cycle. (author)

  10. Potential applications of sonochemistry in spent nuclear fuel reprocessing: a short review.

    Science.gov (United States)

    Nikitenko, S I; Venault, L; Pflieger, R; Chave, T; Bisel, I; Moisy, P

    2010-08-01

    The industrial treatment of spent nuclear fuel is based upon a hydrometallurgical process in nitric acid medium. In order to minimize the volume of radioactive waste it seems interesting to generate the reactive species in situ in such solutions using ultrasonic irradiation without addition of salt-forming reagents. This review summarizes for the first time the versatile sonochemical processes with uranium, neptunium and plutonium in homogeneous nitric acid solutions and heterogeneous systems. The dissolution of refractory solids, ultrasonically driven liquid-liquid extraction and the sonochemical degradation of the volatile products of organic solvent radiolysis issued from PUREX process are considered. Also the guidelines for required further work to ensure successful application of the studied processes at industrial scale are discussed.

  11. Potential applications of sono-chemistry in spent nuclear fuel reprocessing: A short review

    Energy Technology Data Exchange (ETDEWEB)

    Nikitenko, S. I.; Pflieger, R.; Chave, T. [CEA Marcoule, CNRS, UMII, ICSM, ENSCM, UMR 5257, F-30207 Bagnols Sur Ceze (France); Venault, L.; Bisel, I.; Moisy, P. [CEA Marcoule, DEN DRCP, F-30207 Bagnols Sur Ceze (France)

    2010-07-01

    The industrial treatment of spent nuclear fuel is based upon a hydrometallurgical process in nitric acid medium. In order to minimize the volume of radioactive waste it scorns interesting to generate the reactive species in situ in such solutions using ultrasonic irradiation without addition of salt-forming reagents. This review summarizes for the first time the versatile sono-chemical processes with uranium, neptunium and plutonium in homogeneous nitric acid solutions and heterogeneous systems. The dissolution of refractory solids, ultrasonically driven liquid-liquid extraction and the sono-chemical degradation of the volatile products of organic solvent radiolysis issued from PUREX process are considered. Also the guidelines for required further work to ensure successful application of the studied processes at industrial scale are discussed. (authors)

  12. Symposium on the reprocessing of irradiated fuels. Book 2, Session IV

    Energy Technology Data Exchange (ETDEWEB)

    None

    1958-12-31

    Book two of this conference has a single-focused session IV entitled Nonaqueous Processing, with 8 papers. The session deals with fluoride volatility processes and pyrometallurgical or pyrochemical processes. The latter involves either an oxide drossing or molten metal extraction or fused salt extraction technique and results in only partial decontamination. Fluoride volatility processes appear to be especially favorable for recovery of enriched uranium and decontamination factors of 10/sup 7/ to 10/sup 8/ would be achieved by simpler means than those employed in solvent extraction. Data from lab research on the BrF/sub 3/ process and the ClF/sub 3/ process are given and discussed and pilot plant experience is described, all in connection with natural uranium or slightly enriched uranium processing. Fluoride volatility processes for enriched or high alloy fuels are described step by step. The economic and engineering considerations of both types of nonaqueous processing are treated separately and as fully as present knowledge allows. A comprehensive review of the chemistry of pyrometallurgical processes is included.

  13. Glutarimidedioxime: a complexing and reducing reagent for plutonium recovery from spent nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Xian, Liang [Radiochemistry Department, China Institute of Atomic Energy, Beijing (China); Tian, Guoxin [Radiochemistry Department, China Institute of Atomic Energy, Beijing (China); Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Beavers, Christine M.; Teat, Simon J. [Advanced Light Source, Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Shuh, David K. [Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, CA (United States)

    2016-04-04

    Efficient separation processes for recovering uranium and plutonium from spent nuclear fuel are essential to the development of advanced nuclear fuel cycles. The performance characteristics of a new salt-free complexing and reducing reagent, glutarimidedioxime (H{sub 2}A), are reported for recovering plutonium in a PUREX process. With a phase ratio of organic to aqueous of up to 10:1, plutonium can be effectively stripped from 30 % tributyl phosphate (TBP) in kerosene into 1 m HNO{sub 3} with H{sub 2}A. The complexation-reduction mechanism is illustrated with the combination of UV/Vis absorption spectra and the crystal structure of a Pu{sup IV} complex with the reagent. The fast stripping rate and the high efficiency for stripping Pu{sup IV}, through the complexation-reduction mechanism, is suitable for use in centrifugal contactors with very short contact/resident times, thereby offering significant advantages over conventional processes. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  14. Improving ATLAS reprocessing software

    CERN Document Server

    Novak, Tadej

    2014-01-01

    For my CERN Summer Student programme I have been working with ATLAS reprocessing group. Data taken at ATLAS experiment is not only processed after being taken, but is also reprocessed multiple times afterwards. This allows applying new alignments, calibration of detector and using improved or faster algorithms. Reprocessing is usually done in campaigns for different periods of data or for different interest groups. The idea of my project was to simplify the definition of tasks and monitoring of their progress. I created a LIST configuration files generator script in Python and a monitoring webpage for tracking current reprocessing tasks.

  15. Effect of reprocessing cycles on the degradation of polypropylene copolymer filled with talc or montmorillonite during injection molding process

    Energy Technology Data Exchange (ETDEWEB)

    Demori, R.; Mauler, R. S., E-mail: raquel.mauler@ufrgs.br [Chemistry Institute, Federal University of Rio Grande do Sul, UFRGS, Av. Bento Gonçalves, 9500, Porto Alegre, 91501-970 (Brazil); Ashton, E.; Weschenfelder, V. F.; Cândido, L. H. A.; Kindlein, W. [Laboratory of Design LDSM, Federal University of Rio Grande do Sul, UFRGS (Brazil)

    2015-05-22

    Mechanical recycling of polymeric materials is a favorable technique resulting in economic and environmental benefits, especially in the case of polymers with a high production volume as the polypropylene copolymer (PP). However, recycling by reprocessing techniques can lead to thermal, mechanical or thermo-oxidative degradation that can affect the structure of the polymer and subsequently the material properties. PP filled with montmorillonite (MMT) or talc are widely produced and studied, however, its degradation reactions by reprocessing cycles are poorly studied so far. In this study, the effects of reprocessing cycles in the structure and in the properties of the PP/MMT and PP/Talc were evaluated. The samples were mixed with 5% talc or MMT Cloisite C15A in a twin-screw extrusion. After extrusion, this filled material was submitted to five reprocessing cycles through an injection molding process. In order to evaluate the changes induced by reprocessing techniques, the samples were characterized by DSC, FT-IR, Izod impact and tensile strength tests. The study showed that Young modulus, elongation at brake and Izod impact were not affected by reprocessing cycles, except when using talc. In this case, the elongation at brake reduced until the fourth cycle, showing rigidity increase. The DSC results showed that melting and crystallization temperature were not affected. A comparison of FT-IR spectra of the reprocessed indicated that in both samples, between the first and the fifth cycle, no noticeable change has occurred. Thus, there is no evidence of thermo oxidative degradation. In general, these results suggest that PP reprocessing cycles using MMT or talc does not change the material properties until the fifth cycle.

  16. Thorium utilization program. Quarterly progress report for the period ending November 30, 1975. [Fuel element crushing, solids handling, fluidized bed combustion, aqueous separations, solvent extraction, systems design and drafting, alternative head-end reprocessing, and fuel recycle systems analysis

    Energy Technology Data Exchange (ETDEWEB)

    1975-12-31

    The development program for HTGR fuel reprocessing continues to emphasize the design and construction of a prototype head-end line. Design work on the multistage crushing system, the primary and secondary fluidized bed burners, the pneumatic transfer systems, and the ancillary fixtures for semiremote assembly and disassembly is essentially complete. Fabrication and receipt of all major components is under way, and auxiliary instrumentation and support systems are being installed. Studies of flow characteristics of granular solids in pneumatic transfer systems are continuing and data are being collected for use in design of systems for solids handling. Experimental work on the 20-cm primary fluidized bed burner verified the fines recycle operating mode in runs of greater than 24 hr. Twelve leaching runs were performed during the quarter using crushed, burned-back TRISO coated ThC/sub 2/ particles and burned-back BISO coated sol gel ThO/sub 2/ particles to examine the effect of varying the Thorex-to-thoria ratio to give product solutions ranging from 0.25M to 1M in thorium. Only minor effects were observed and reference values for facility operations were specified. Two-stage leaching runs with burned-back ThC/sub 2/ indicate there are no measurable differences in total dissolution time as compared to single-stage leaching. Bench-scale tests on oxidation of HTGR fuel boron carbide at 900/sup 0/C indicates that most if not all of the carbide will be converted to boron oxide in the fluidized bed burner. Eight solvent extraction runs were completed during the quarter. These runs represented the first cycle and second uranium cycle of the acid-Thorex flowsheet. A detailed calculation of spent fuel compositions by fuel block and particle type is being performed for better definition of process streams in a fuel reprocessing facility.

  17. Estimation of 85Kr dispersion from the spent nuclear fuel reprocessing plant in Rokkasho, Japan, using an atmospheric dispersion model.

    Science.gov (United States)

    Abe, K; Iyogi, T; Kawabata, H; Chiang, J H; Suwa, H; Hisamatsu, S

    2015-11-01

    The spent nuclear fuel reprocessing plant of Japan Nuclear Fuel Limited (JNFL) located in Rokkasho, Japan, discharged small amounts of (85)Kr into the atmosphere during final tests of the plant with actual spent fuel from 31 March 2006 to October 2008. During this period, the gamma-ray dose rates due to discharged (85)Kr were higher than the background rates measured at the Institute for Environmental Sciences and at seven monitoring stations of the Aomori prefectural government and JNFL. The dispersion of (85)Kr was simulated by means of the fifth-generation Penn State/NCAR Mesoscale Model and the CG-MATHEW/ADPIC models (ver. 5.0) with a vertical terrain-following height coordinate. Although the simulated gamma-ray dose rates due to discharged (85)Kr agreed fairly well with measured rates, the agreement between the estimated monthly mean (85)Kr concentrations and the observed concentrations was poor. Improvement of the vertical flow of air may lead to better estimation of (85)Kr dispersion.

  18. Public comments and Task Force responses regarding the environmental survey of the reprocessing and waste management portions of the LWR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    1977-03-01

    This document contains responses by the NRC Task Force to comments received on the report ''Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle'' (NUREG-0116). These responses are directed at all comments, inclding those received after the close of the comment period. Additional information on the environmental impacts of reprocessing and waste management which has either become available since the publication of NUREG-0116 or which adds requested clarification to the information in that document.

  19. Using OPUS to Perform HST On-The-Fly Re-Processing (OTFR)

    Science.gov (United States)

    Swam, M. S.; Hopkins, E.; Swade, D. A.

    The Hubble Space Telescope (HST) OPUS implementation of On-The-Fly Calibration (OTFC) processing currently provides the benefit of applying the most current calibration algorithms, reference files, and repairs of errant header keyword values (aperture, shutter, etc.) to Level-1b datasets as they are retrieved from the HST archive. While OTFC has performed well, a number of concerns about maintenance and flexibility have resulted in the evolution towards an On-The-Fly Re-Processing (OTFR) System. Also based on the OPUS pipeline architecture, OTFR carries further the notion of creating products for archive users at the time of their request, by completely regenerating calibrated (Level-2) data products for an exposure from the base telemetry files sent from the HST (Level-1a data). By starting processing at this earlier state, and taking advantage of the changes to the data processing software that are made as that software matures, improved, more consistent calibrated (Level-2) data products are produced. Use of OPUS distributed multi-processing, the relatively small size of HST datasets, and the efficiency of the data processing and calibration software results in a very small impact on the overall time it takes to complete an archive retrieval. There could be an impact to archive research, however, since the archive catalog meta-data will not completely reflect the reprocessed products as they would be delivered to the archive user. This problem will be addressed by performing catalog updates for any major discrepancies. This paper will describe the concerns raised about OTFC, the design of the OTFR pipeline system, and the benefits of using the OPUS architecture in this design.

  20. Evaluation technology for burnup and generated amount of plutonium by measurement of xenon isotopic ratio in dissolver off-gas at reprocessing facility (Joint research)

    OpenAIRE

    岡野 正紀; 久野 剛彦; 高橋 一朗; 白水 秀知; Charlton, W. S.; Wells, C. A.; Hemberger, P. H.; 山田 敬二; 酒井 敏雄

    2006-01-01

    The amount of Pu in the spent fuel was evaluated from Xe isotopic ratio in off-gas in reprocessing facility, is related to burnup. Six batches of dissolver off-gas at spent fuel dissolution process were sampled from the main stack in Tokai Reprocessing Plant during BWR fuel reprocessing campaign. Xenon isotopic ratio was determined with GC/MS. Burnup and generated amount of Pu were evaluated with Noble Gas Environmental Monitoring Application code (NOVA), developed by Los Alamos National Labo...

  1. Metal-organic frameworks for removal of Xe and Kr from nuclear fuel reprocessing plants.

    Science.gov (United States)

    Liu, Jian; Thallapally, Praveen K; Strachan, Denis

    2012-08-01

    Removal of xenon (Xe) and krypton (Kr) from process off-gases containing 400 ppm Xe, 40 ppm Kr, 78% N(2), 21% O(2), 0.9% Ar, 0.03% CO(2), and so forth using adsorption was demonstrated for the first time. Two well-known metal-organic frameworks (MOFs), HKUST-1 and Ni/DOBDC, which both have unsaturated metal centers but different pore morphologies, were selected as novel adsorbents. Results of an activated carbon were also included for comparison. The Ni/DOBDC has higher Xe/Kr selectivities than those of the activated carbon and the HKUST-1. In addition, results show that the Ni/DOBDC and HKUST-1 can adsorb substantial amounts of Xe and Kr even when they are mixed in air. Moreover, the Ni/DOBDC can successfully separate 400 ppm Xe from 40 ppm Kr and air containing O(2), N(2), and CO(2) with a Xe/Ke selectivity of 7.3 as indicated by our breakthrough results. This shows a promising future for MOFs in radioactive nuclide separations from spent fuels.

  2. NOISE CHARACTERISTIC AND SEASONAL SIGNALS IN THE RE-PROCESSED EUREF PERMANENT NETWORK COORDINATE TIME SERIES

    Science.gov (United States)

    Kenyeres, A.; Williams, S. D.; Figurski, M.; van Dam, T. M.; Szafranek, K.

    2009-12-01

    Previous analyses of periodic signals present in continuous GPS time series showed that the amplitude and phase of the derived seasonal term mostly disagree with surface mass loading models. The CGPS results appeared to over-estimate the amplitude of the seasonal term and the estimated amplitudes and/or phases were poorly coherent with the loading models, especially at sites close to coastal areas. The studies concluded that the GPS results are distorted by analysis artifacts (such as ocean tide loading, aliasing, and antenna phase centre variation models), monument thermal effects, and multipath. In addition, the actual CGPS time series were inhomogeneous in terms of processing strategy, applied models and reference frame alignment. With the introduction of absolute antenna phase centre variation models an effort, within the EUREF Permanent Network, was initiated to produce a complete GPS re-analysis from global to local levels. A test re-processing of all EPN observations from 1996 to 2007 has already been completed by the Military University of Technology (MUT), Warsaw, Poland and cumulative EPN solutions, from the daily SINEX files, have been created using the CATREF software. We used a combination of Weighted Least Squares, Maximum Likelihood Estimation (MLE), Empirical Orthogonal Functions (EOF’s) and Wavelets to analyze the data for their spatial and temporal noise characteristics and investigate the periodic signals. We find that the noise levels in the re-processed daily solutions is reduced compared to past solutions, but the noise spectra is still represented by a combination of flicker noise and white noise. The amplitudes of the seasonal term have generally decreased and the spatial distribution of the phase lag appears to be more uniform. Comparisons of the estimated annual variations with combined loading models (NCEP + LaD - World - Fraser + ECCO) and the vertical displacement model of the GRACE R4 gravity fields show an improved agreement

  3. 76 FR 13605 - Notice of Availability of Draft Waste Incidental to Reprocessing Evaluation for the Vitrification...

    Science.gov (United States)

    2011-03-14

    ... waste from reprocessing of spent nuclear fuel and certain treatment material) at the West Valley... a solid glass waste form. DOE used the vitrification melter as part of this process, specifically to melt glass frit (material used in making glass) together with reprocessing waste sludge and...

  4. Application of ionic liquids in nuclear fuel reprocessing%离子液体在核燃料后处理中的应用

    Institute of Scientific and Technical Information of China (English)

    袁立永; 石伟群; 蓝建慧; 柴之芳

    2012-01-01

    核燃料后处理是核燃料循环的核心,对于核环境安全和核能的可持续发展意义重大.离子液体作为“新一代绿色溶剂”在核燃料后处理中具有广阔的潜在应用前景.离子液体可以替代易挥发的有机溶剂用于水法后处理萃取分离放射性核素,也可以替代强腐蚀性的高温熔盐用于干法后处理电解回收金属离子.本文在作者工作基础上总结了近年来离子液体用于核燃料水法和干法后处理的基础研究成果,归纳和分析了其中的关键科学问题.此外,由于核燃料后处理涉及强辐射应用环境,离子液体的辐射稳定性是其实际应用的前提和关键,因此本文还综述了国内外有关离子液体辐射效应的研究进展,评估了离子液体用于核燃料后处理的辐射化学可行性.最后,基于当前的研究现状和研究水平展望了离子液体在核燃料后处理应用方面的研究前景.%The nuclear fuel reprocessing, as the essential part of nuclear fuel cycle, is of great significance from the point of view of both nuclear safety and sustainable development of nuclear energy. Room temperature ionic liquids (RTILs) regarded as "new generation green solvents" have recently received an ever-increasing amount of interest in nuclear fuel reprocessing due to their unique physical and chemical properties. They can be used in aqueous reprocessing as environmentally benign alternatives to volatile organic solvents for traditional liquid-liquid extraction of high level radioactive nuclides. They are also applicable in non-aqueous reprocessing" by substituting caustic molten salts for electro-deposition of metal ions. Herein, we reviewed the recent basic researches on the utility of RTILs in nuclear fuel reprocessing, from which the key scientific issues on their practical application were summarized. In addition, it is well known that nuclear fuel reprocessing involves high-level radioactive matter, and full

  5. Comparison of radiation hazard of HLW in several spent nuclear fuel reprocessing scenarios

    Directory of Open Access Journals (Sweden)

    Gladilov D.

    2012-10-01

    Full Text Available Radiation hazard of radionuclide has been calculated as a product of Aε where A is an activity of radionuclide and ε is a dose coefficient through ingestion. The values Aε of 18 radionuclide in spent fuel of WWER-440 are calculated. Because the full division of americium and curium from HLW is very complicated a separation americium from curium is considered. It is shown that a separation of americium in a special fraction allows decreasing the radiation hazard of HLW by 97.6% after 1000 years.

  6. Thorium utilization program. Quarterly progress report for the period ending May 31, 1976. [Fuel element crushing, solids handling, fluidized-bed combustion, aqueous separations, solvent extraction, off-gas studies, semiremote handling systems, alternative head-end processing, and fuel recycle design

    Energy Technology Data Exchange (ETDEWEB)

    1976-06-30

    The work reported includes the development of unit processes and equipment for reprocessing of High-Temperature Gas-Cooled Reactor (HTGR) fuel and the design and development of an integrated line to demonstrate the head end of HTGR reprocessing using unirradiated fuel materials. Work is also described on trade-off studies concerning the required design of recycle facilities for the large-scale recycle of HTGR fuels in order to guide the development activities for HTGR fuel recycle.

  7. NASADEM Initial Production Processing Results: Shuttle Radar Topography Mission (SRTM) Reprocessing with Improvements

    Science.gov (United States)

    Buckley, S.; Agram, P. S.; Belz, J. E.; Crippen, R. E.; Gurrola, E. M.; Hensley, S.; Kobrick, M.; Lavalle, M.; Martin, J. M.; Neumann, M.; Nguyen, Q.; Rosen, P. A.; Shimada, J.; Simard, M.; Tung, W.

    2016-12-01

    NASADEM is a significant modernization of SRTM digital elevation model (DEM) data supported by the NASA MEaSUREs program. We are reprocessing the raw radar signal data using improved algorithms and incorporating ICESat and DEM data unavailable during the original processing. The NASADEM products will be freely-available through the Land Processes Distributed Active Archive Center (LPDAAC) at one-arcsecond spacing and delivered by continent: North America, South America, Australia, Eurasia, Africa, and Island Groups. We are in the production phase of the project. This involves radar interferometry (InSAR) processing on thousands of radar datatakes. New phase unwrapping and height ripple error correction (HREC) procedures are applied to the data. The resulting strip DEMs and ancillary information are passed to a back-end processor to create DEM mosaics and new geocoded single-swath products. Manual data quality assessment (QA) and fixes are performed at several steps in the processing chain. Post-production DEM void-filling is described in a companion AGU Fall Meeting presentation. The team completed the InSAR processing for all continents and the manual QA of the strip DEMs for more than half the world. North America strip DEM void areas are reduced by more than 50%. The ICESat data is used for height ripple error correction and as control for continent-scale adjustment of the strip DEMs. These ripples are due to uncompensated mast motion most pronounced after Shuttle roll angle adjustment maneuvers. After an initial assessment of the NASADEM production processing for the Americas, we further refined the selection of ICESat data for control by excluded data over glaciers, snow cover, forest clear cuts, and sloped areas. The HREC algorithm reduces the North America ICESat-SRTM bias from 80 cm to 3 cm and the RMS from 5m to 4m.

  8. Noble gas atmospheric monitoring at reprocessing facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nakhleh, C.W.; Perry, R.T. Jr.; Poths, J.; Stanbro, W.D.; Wilson, W.B.; Fearey, B.L.

    1997-05-01

    The discovery in Iraq after the Gulf War of the existence of a large clandestine nuclear-weapon program has led to an across-the-board international effort, dubbed Programme 93+2, to improve the effectiveness and efficiency of International Atomic Energy Agency (IAEA) safeguards. One particularly significant potential change is the introduction of environmental monitoring (EM) techniques as an adjunct to traditional safeguards methods. Monitoring of stable noble gas (Kr, Xe) isotopic abundances at reprocessing plant stacks appears to be able to yield information on the burnup and type of the fuel being processed. To estimate the size of these signals, model calculations of the production of stable Kr, Xe nuclides in reactor fuel and the subsequent dilution of these nuclides in the plant stack are carried out for two case studies: reprocessing of PWR fuel with a burnup of 35 GWd/tU, and reprocessing of CAND fuel with a burnup of 1 GWd/tU. For each case, a maximum-likelihood analysis is used to determine the fuel burnup and type from the isotopic data.

  9. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: reprocessing light-water reactor fuel. [Radiation dose commitment to human populations from radioactive effluents released to environment

    Energy Technology Data Exchange (ETDEWEB)

    Finney, B.C.; Blanco, R.E.; Dahlman, R.C.; Hill, G.S.; Kitts, F.G.; Moore, R.E.; Witherspoon, J.P.

    1976-10-01

    A cost/benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model nuclear fuel reprocessing plant which processes light-water reactor (LWR) fuels, and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term as low as reasonably achievable in relation to limiting the release of radioactive materials from nuclear facilities. The base case model plant is representative of current plant technology and has an annual capacity of 1500 metric tons of LWR fuel. Additional radwaste treatment systems are added to the base case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The cost for the added waste treatment operations and the corresponding dose commitments are calculated for each case. In the final analysis, radiological dose is plotted vs the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Much of the technology used in the advanced cases is in an early stage of development and is not suitable for immediate use. The methodology used in estimating the costs, and the radiological doses, detailed calculations, and tabulations are presented in Appendix A and ORNL-4992. This report is a revision of the original study (ORNL/TM-4901).

  10. Behavior of silicon in nitric media. Application to uranium silicides fuels reprocessing; Comportement du silicium en milieu nitrique. Application au retraitement des combustibles siliciures d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Cheroux, L

    2001-07-01

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

  11. Tritium concentrations in the atmospheric environment at Rokkasho, Japan before the final testing of the spent nuclear fuel reprocessing plant.

    Science.gov (United States)

    Akata, Naofumi; Kakiuchi, Hideki; Shima, Nagayoshi; Iyogi, Takashi; Momoshima, Noriyuki; Hisamatsu, Shun'ichi

    2011-09-01

    This study aimed at obtaining background tritium concentrations in precipitation and air at Rokkasho where the first commercial spent nuclear fuel reprocessing plant in Japan has been under construction. Tritium concentration in monthly precipitation during fiscal years 2001-2005 had a seasonal variation pattern which was high in spring and low in summer. The tritium concentration was higher than that observed at Chiba City as a whole. The seasonal peak concentration at Rokkasho was generally higher than that at Chiba City, while the baseline concentrations of both were similar. The reason for the difference may be the effect of air mass from the Asian continent which is considered to have high tritium concentration. Atmospheric tritium was operationally separated into HTO, HT and hydrocarbon (CH(3)T) fractions, and the samples collected every 3 d-14 d during fiscal year 2005 were analyzed for these fractions. The HTO concentration as radioactivity in water correlated well with that in the precipitation samples. The HT concentration was the highest among the chemical forms analyzed, followed by the HTO and CH(3)T concentrations. The HT and CH(3)T concentrations did not have clear seasonal variation patterns. The HT concentration followed the decline previously reported by Mason and Östlund with an apparent half-life of 4.8 y. The apparent and environmental half-lives of CH(3)T were estimated as 9.2 y and 36.5 y, respectively, by combining the present data with literature data. The Intergovernmental Panel on Climate Change used the atmospheric lifetime of 12 y for CH(4) to estimate global warming in its 2007 report. The longer environmental half-life of CH(3)T suggested its supply from other sources than past nuclear weapon testing in the atmosphere. Copyright © 2011 Elsevier Ltd. All rights reserved.

  12. The formation of hydrophilic Np(IV) complexes and their potential application in nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    May, I.; Taylor, R.J.; Brown, G. [British Nuclear Fuels Ltd., Sellafield (United Kingdom)

    1998-06-12

    A series of organic ligands have been screened for their effectiveness as complexants for Np(IV) in a neptunium rejection stage of an advanced PUREX process. Four of these species, formohydroxamic acid, acetohydroxamic acid, glycolic acid and pyruvic acid, readily form hydrophilic complexes with Np(IV) and can strip the actinide from 30% TBP/OK (30% tributylphosphate in odourless kerosene) into nitric acid. Near infra-red spectroscopy has been used to monitor Np(IV) complexation in nitric acid. Distribution experiments have been undertaken between nitric acid and 30% TBP/OK to examine the effect of ligand and nitric acid concentration on Np(IV) stripping. Finally, it has been shown that the extractability of U(VI) is unaffected by the presence of these ligands and all can be used to selectively strip Np(IV) from a U(VI) product stream in an advanced PUREX process. (orig.) 11 refs.

  13. Consolidated fuel reprocessing program. Progress report, October 1--December 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Burch, W.D.; Feldman, M.J.; Groenier, W.S.; Vondra, B.L.; Unger, W.E.

    1979-03-01

    The status of the following studies is reported: plutonium (IV) polymer reaction in aqueous solutions; plutonium reductive stripping studies; plutonium--uranium--thorium coprocessing studies; plutonium losses due to solution instability and solids formation; solvent cleanup; nitrogen compound chemistry; fission product chemistry; electrochemical methods evaluation; evaluation of alternate extractants; hot-cell development; solvent extraction; product conversion; analytical chemistry development; voloxidation; dissolution; feed preparation; off-gas processing; and engineering systems. (LK)

  14. Used nuclear fuel separations process simulation and testing

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D. [Argonne National Laboratory: 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2013-07-01

    Recent efforts in separations process simulation at Argonne have expanded from the traditional focus on solvent extraction flowsheet design in order to capture process dynamics and to simulate other components, processing and systems of a used nuclear fuel reprocessing plant. For example, the Argonne Model for Universal Solvent Extraction (AMUSE) code has been enhanced to make it both more portable and more readily extensible. Moving away from a spreadsheet environment makes the addition of new species and processes simpler for the expert user, which should enable more rapid implementation of chemical models that simulate evolving processes. The dyAMUSE (dynamic AMUSE) version allows the simulation of transient behavior across an extractor. Electrochemical separations have now been modeled using spreadsheet codes that simulate the electrochemical recycle of fast reactor fuel. The user can follow the evolution of the salt, products, and waste compositions in the electro-refiner, cathode processors, and drawdown as a function of fuel batches treated. To further expand capabilities in integrating multiple unit operations, a platform for linking mathematical models representing the different operations that comprise a reprocessing facility was adapted to enable systems-level analysis and optimization of facility functions. (authors)

  15. Effect of Cognitive Processing Therapy and Holographic Reprocessing on Reduction of Posttraumatic Cognitions in Students Exposed to Trauma

    Directory of Open Access Journals (Sweden)

    Parviz molavi

    2011-10-01

    Full Text Available "nObjective: This research was conducted to examine the effect of cognitive processing therapy and holographic reprocessing on the reduction of posttraumatic cognitions in students exposed to trauma. "nMethod: This was an experimental study with spread pretest-posttest randomized groups design. Statistical society of this research consisted of male freshman, junior and senior high school students of Uremia (N=10286. Utilizing Traumatic Events Screening Inventory, and SCL-90 R on 1000 randomly selected high school students, 129 students were recognized as having experienced traumatic events. Of the subjects, 60 were selected randomly. Then, clinical interview was conducted, and the selected sample was randomly assigned in to three groups of cognitive processing therapy, holographic reprocessing and control. These groups responded to Posttraumatic Cognitions Inventory in pretest and post test. Differences of pre-post test scores were analyzed using one way ANOVA and Scheffe test. "nResults: The results demonstrated significant differences between the three groups in total score of the Posttraumatic Cognition Inventory. Difference was also observed in negative cognitions on self and self-blame dimensions. Furthermore, these two therapeutic methods were equally effective in the reduction of posttraumatic cognitions.   "nConclusion: It appears that cognitive processing therapy and holographic reprocessing which had been originally developed and tested for sexually assaulted females, can also be applied for the victims of other traumatic events, particularly  adolescents.

  16. DEVELOPMENT OF CRYSTALLINE CERAMICS FOR IMMOBILIZATION OF ADVANCED FUEL CYCLE REPROCESSING WASTES

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.; Brinkman, K.

    2011-09-22

    The Savannah River National Laboratory (SRNL) is developing crystalline ceramic waste forms to incorporate CS/LN/TM high Mo waste streams consisting of perovskite, hollandite, pyrochlore, zirconolite, and powellite phase assemblages. Simple raw materials, including Al{sub 2}O{sub 3}, CaO, and TiO{sub 2} were combined with simulated waste components to produce multiphase crystalline ceramics. Fiscal Year 2011 (FY11) activities included (i) expanding the compositional range by varying waste loading and fabrication of compositions rich in TiO{sub 2}, (ii) exploring the processing parameters of ceramics produced by the melt and crystallize process, (iii) synthesis and characterization of select individual phases of powellite and hollandite that are the target hosts for radionuclides of Mo, Cs, and Rb, and (iv) evaluating the durability and radiation stability of single and multi-phase ceramic waste forms. Two fabrication methods, including melting and crystallizing, and pressing and sintering, were used with the intent of studying phase evolution under various sintering conditions. An analysis of the XRD and SEM/EDS results indicates that the targeted crystalline phases of the FY11 compositions consisting of pyrochlore, perovskite, hollandite, zirconolite, and powellite were formed by both press and sinter and melt and crystallize processing methods. An evaluation of crystalline phase formation versus melt processing conditions revealed that hollandite, perovskite, zirconolite, and residual TiO{sub 2} phases formed regardless of cooling rate, demonstrating the robust nature of this process for crystalline phase development. The multiphase ceramic composition CSLNTM-06 demonstrated good resistance to proton beam irradiation. Electron irradiation studies on the single phase CaMoO{sub 4} (a component of the multiphase waste form) suggested that this material exhibits stability to 1000 years at anticipated self-irradiation doses (2 x 10{sup 10}-2 x 10{sup 11} Gy), but that

  17. The Eye Movement Desensitization and Reprocessing Procedure Prevents Defensive Processing in Health Persuasion

    NARCIS (Netherlands)

    Dijkstra, Arie; van Asten, Regine

    2014-01-01

    In the present study, the method of eye movement desensitization and reprocessing (EMDR) is studied to understand and prevent defensive reactions with regard to a negatively framed message advocating fruit and vegetable consumption. EMDR has been shown to tax the working memory. Participants from a

  18. A novel technique towards deployment of hydrostatic pressure based level sensor in nuclear fuel reprocessing facility.

    Science.gov (United States)

    Praveen, K; Rajiniganth, M P; Arun, A D; Sahoo, P; Murty, S A V Satya

    2016-02-01

    A novel approach towards deployment of a hydrostatic pressure based level monitoring device is presented for continuous monitoring of liquid level in a reservoir with high resolution and precision. Some of the major drawbacks such as spurious information of measured level due to change in ambient temperature, requirement of high resolution pressure sensor, and bubbling effect by passing air or any gaseous fluid into the liquid are overcome by using such a newly designed hydrostatic pressure based level monitoring device. The technique involves precise measurement of hydrostatic pressure exerted by the process liquid using a high sensitive pulsating-type differential pressure sensor (capacitive type differential pressure sensor using a specially designed oil manometer) and correlating it to the liquid level. In order to avoid strong influence of temperature on liquid level, a temperature compensation methodology is derived and used in the system. A wireless data acquisition feature has also been provided in the level monitoring device in order to work in a remote area such as a radioactive environment. At the outset, a prototype level measurement system for a 1 m tank is constructed and its test performance has been well studied. The precision, accuracy, resolution, uncertainty, sensitivity, and response time of the prototype level measurement system are found to be less than 1.1 mm in the entire range, 1%, 3 mm, <1%, 10 Hz/mm, and ∼4 s, respectively.

  19. Conversion of uranium nuclear fuel into U 3O 8 at the head end of HTR reprocessing

    Science.gov (United States)

    Hoogen, N.; Aschhoff, H. G.; Staib, G.

    1984-04-01

    Corresponding to the reference procedure for the head-end treatment of HTR fuel elements, separation of the moderator graphite from the materials uranium and plutonium is envisaged by combustion in the fluidized bed. Due to the defective silicon carbide layers of the uranium fuel particles a chemical conversion of the UO 2 kernel into U 3O 8 takes place in the oxidizing atmosphere of the combustion process. This reaction proceeds spontaneously and quantitatively, and causes a disintegration of the heavy metal kernel. It is observed that the degree of hardness of the kernel fragments is clearly dependent on the heat-up rate. In the commercial design of the head-end process step, attention must be paid to the cross-over of fuel from the stationary fluidized bed into the dust discharge.

  20. Monolithic Fuel Fabrication Process Development

    Energy Technology Data Exchange (ETDEWEB)

    C. R. Clark; N. P. Hallinan; J. F. Jue; D. D. Keiser; J. M. Wight

    2006-05-01

    The pursuit of a high uranium density research reactor fuel plate has led to monolithic fuel, which possesses the greatest possible uranium density in the fuel region. Process developments in fabrication development include friction stir welding tool geometry and cooling improvements and a reduction in the length of time required to complete the transient liquid phase bonding process. Annealing effects on the microstructures of the U-10Mo foil and friction stir welded aluminum 6061 cladding are also examined.

  1. Estimation of tritium dispersion from the spent nuclear fuel reprocessing plant in Rokkasho using an atmospheric dispersion model

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Koichi; Kakiuchi, Hideki; Iyogi, Takashi; Hisamatsu, Shun' ichi [Institute for Environmental Sciences, Rokkasho, Aomori 039-3212 (Japan); Akata, Naofumi [Institute for Environmental Sciences, Rokkasho, Aomori 039-3212 (Japan); National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Chiang, Jing-Hsien; Suwa, Hiroji [Japan NUS Co., Ltd., Tokyo 160-0023 (Japan)

    2014-07-01

    Japan's first large-scale commercial plant for reprocessing spent nuclear fuel was constructed in Rokkasho, Japan, by Japan Nuclear Fuel Limited (JNFL). Final tests of plant operation carried out with spent fuels since 31 March 2006 have indicated that small amounts of radionuclides (mainly {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I) are discharged into the atmosphere from the main stack of the plant. To estimate the atmospheric dispersion of {sup 3}H discharged from the plant, we used a combination of the Fifth-Generation Penn State/NCAR Mesoscale Model (MM5) and the CG-MATHEW/ADPIC models, Version 5.0 (ARAC-2). Simulation results were validated with atmospheric {sup 3}H concentrations and wet deposition rates measured at the Institute for Environmental Sciences (IES), located 2.6 km east from the stack. Biweekly atmospheric HTO, HT, and CH3T samples and monthly precipitation samples were collected at IES from April 2006 to February 2009 (the test period). Concentrations of {sup 3}H in the samples were measured with a low-background liquid scintillation counter (LSC-LB5, Hitachi Aloka Medical, Ltd., Tokyo, Japan). To simulate the dispersion of {sup 3}H from the stack, a meteorological field was calculated by MM5 and used as input to ARAC-2, which consists of a mass-consistent wind model and a particle-tracing-type dispersion model. The simulation areas were 315 km x 315 km for MM5 and 50 km x 50 km for ARAC-2. The following meteorological data were input to MM5: grid point data derived from the Mesoscale Model of the Japan Meteorological Agency (JMA), data from JMA's Automated Meteorological Data Acquisition System (AMeDAS), and wind speed and direction at IES and JNFL measured every 10 min. The weekly discharge rates of {sup 3}H disclosed by JNFL were used as the source term for ARAC-2. The concentrations of {sup 3}H in atmospheric moisture and precipitation samples increased from their background values during the test period. As an index of

  2. Standard model for the safety analysis report of nuclear fuel reprocessing plants; Modelo padrao para relatorio de analise de seguranca de usinas de reprocessamento de combustiveis nucleares

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-02-15

    This norm establishes the Standard Model for the Safety Analysis Report of Nuclear Fuel Reprocessing Plants, comprehending the presentation format, the detailing level of the minimum information required by the CNEN for evaluation the requests of Construction License or Operation Authorization, in accordance with the legislation in force. This regulation applies to the following basic reports: Preliminary Safety Analysis Report - PSAR, integrating part of the requirement of Construction License; and Final Safety Analysis Report (FSAR) which is the integrating part of the requirement for Operation Authorization.

  3. Biodegradation of radioactive organic liquid waste from spent fuel reprocessing; Biodegradacao de rejeitos radioativos liquidos organicos provenientes do reprocessamento do combustivel nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Rafael Vicente de Padua

    2008-07-01

    The research and development program in reprocessing of low burn-up spent fuel elements began in Brazil in 70's, originating the lab-scale hot cell, known as Celeste located at Nuclear and Energy Research Institute, IPEN - CNEN/SP. The program was ended at the beginning of 90's, and the laboratory was closed down. Part of the radioactive waste generated mainly from the analytical laboratories is stored waiting for treatment at the Waste Management Laboratory, and it is constituted by mixture of aqueous and organic phases. The most widely used technique for the treatment of radioactive liquid wastes is the solidification in cement matrix, due to the low processing costs and compatibility with a wide variety of wastes. However, organics are generally incompatible with cement, interfering with the hydration and setting processes, and requiring pre -treatment with special additives to stabilize or destroy them. The objective of this work can be divided in three parts: organic compounds characterization in the radioactive liquid waste; the occurrence of bacterial consortia from Pocos de Caldas uranium mine soil and Sao Sebastiao estuary sediments that are able to degrade organic compounds; and the development of a methodology to biodegrade organic compounds from the radioactive liquid waste aiming the cementation. From the characterization analysis, TBP and ethyl acetate were chosen to be degraded. The results showed that selected bacterial consortia were efficient for the organic liquid wastes degradation. At the end of the experiments the biodegradation level were 66% for ethyl acetate and 70% for the TBP. (author)

  4. Analytical Methods of Dry Reprocessing Technology for Spent Nuclear Fuel%乏燃料干法后处理研究中的分析方法

    Institute of Scientific and Technical Information of China (English)

    白雪; 常志远

    2016-01-01

    综述了几种典型的乏燃料干法后处理方法,并对其中使用的分析方法进行了总结。详细论述了干法后处理研究中的在线分析方法,包括电化学分析方法、紫外可见吸收光谱法、X射线衍射法、拉曼原位分析、EXAFS原位分析、NMR原位分析等。在线分析方法有助于对工艺料液中物质的形态及结构进行实时监测。此外,离线分析方法可作为在线方法的有效补充,根据研究对象的形态(气态、液态、固态)对一些典型的离线分析方法进行了论述。%Some typical technical routes of dry reprocessing for spent nuclear fuel were reviewed and the analytical methods used in the processes were summarized .Several methods for on‐line monitoring were demonstrated in detail ,including electroanalytical methods ,UV‐Vis absorption spectrometry ,X‐ray diffraction analysis ,in‐situ Raman spectrometry ,in‐situ EXAFS analysis and in‐situ NMR analysis .On‐line analytical methods can give real‐time information of the morphologies and structures .As the effective complementarity of on‐line analytical methods ,off‐line analytical methods were discussed based on different states of the study object ,such as the one in a gas ,solid or liquid state .

  5. The reprocessing of reactor core materials

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jing, E-mail: wang-jing@nuaa.edu.cn [State Key Laboratory of Mechanics and Control of Mechanial Structures, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Liu, Bing; Shao, Youlin; Lu, Zhenming; Liu, Malin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2014-05-01

    Generation IV high temperature gas-cooled reactors (HTGR) are preferentially fueled by spherical fuel elements, which are composed of a fuel zone of triso-coated uranium oxide (UO{sub 2}) particles and a matrix graphite layer. Unqualified coated particles and spherical fuel elements unavoidablely occur during the processing of coating UO{sub 2} kernels and embedding the coated particles into the graphite matrix. So it is necessary to reprocess the UO{sub 2} in the unqualified coated particles and spherical fuel elements to maximize the use of the reactor core materials. In this work, we present several methods to: (1) separate the coated particles from the graphite matrix and, (2) expose and recover the UO{sub 2} kernels from the coated particles. The comparison of different methods shows that the thermal oxidation of graphite by a fixed bed burner and the jet grinding of the unqualified coated particles are prosing in practice for the separation of coated particles from the graphite matrix and recovering the uranium dioxide kernels, respectively. Some other methods, such as etching the SiC layer with the active fluorine species in plasma generated by the dielectric barrier discharge (DBD) under the atmosphere also show their great potential values in the reprocessing of reactor core materials, especially for the activated and contaminated fuels.

  6. Application of curium measurements for safeguarding at reprocessing plants. Study 1: High-level liquid waste and Study 2: Spent fuel assemblies and leached hulls

    Energy Technology Data Exchange (ETDEWEB)

    Rinard, P.M.; Menlove, H.O.

    1996-03-01

    In large-scale reprocessing plants for spent fuel assemblies, the quantity of plutonium in the waste streams each year is large enough to be important for nuclear safeguards. The wastes are drums of leached hulls and cylinders of vitrified high-level liquid waste. The plutonium amounts in these wastes cannot be measured directly by a nondestructive assay (NDA) technique because the gamma rays emitted by plutonium are obscured by gamma rays from fission products, and the neutrons from spontaneous fissions are obscured by those from curium. The most practical NDA signal from the waste is the neutron emission from curium. A diversion of waste for its plutonium would also take a detectable amount of curium, so if the amount of curium in a waste stream is reduced, it can be inferred that there is also a reduced amount of plutonium. This report studies the feasibility of tracking the curium through a reprocessing plant with neutron measurements at key locations: spent fuel assemblies prior to shearing, the accountability tank after dissolution, drums of leached hulls after dissolution, and canisters of vitrified high-level waste after separation. Existing pertinent measurement techniques are reviewed, improvements are suggested, and new measurements are proposed. The authors integrate these curium measurements into a safeguards system.

  7. Noble Gas Measurement and Analysis Technique for Monitoring Reprocessing Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Charlton, William S [Univ. of California, Berkeley, CA (United States)

    1999-09-01

    An environmental monitoring technique using analysis of stable noble gas isotopic ratios on-stack at a reprocessing facility was developed. This technique integrates existing technologies to strengthen safeguards at reprocessing facilities. The isotopic ratios are measured using a mass spectrometry system and are compared to a database of calculated isotopic ratios using a Bayesian data analysis method to determine specific fuel parameters (e.g., burnup, fuel type, fuel age, etc.). These inferred parameters can be used by investigators to verify operator declarations. A user-friendly software application (named NOVA) was developed for the application of this technique. NOVA included a Visual Basic user interface coupling a Bayesian data analysis procedure to a reactor physics database (calculated using the Monteburns 3.01 code system). The integrated system (mass spectrometry, reactor modeling, and data analysis) was validated using on-stack measurements during the reprocessing of target fuel from a U.S. production reactor and gas samples from the processing of EBR-II fast breeder reactor driver fuel. These measurements led to an inferred burnup that matched the declared burnup with sufficient accuracy and consistency for most safeguards applications. The NOVA code was also tested using numerous light water reactor measurements from the literature. NOVA was capable of accurately determining spent fuel type, burnup, and fuel age for these experimental results. Work should continue to demonstrate the robustness of this system for production, power, and research reactor fuels.

  8. Controllability of plutonium concentration for FBR fuel at a solvent extraction process in the PUREX process

    Energy Technology Data Exchange (ETDEWEB)

    Enokida, Youichi; Kitano, Motoki; Sawada, Kayo [Nagoya University, 1 Furo-cho, Chikusa-ku, Nagoya-shi, Aichi-ken, 4630052 (Japan)

    2013-07-01

    Typical Purex solvent extraction systems for the reprocessing of spent nuclear fuel have a feed material containing dilute, 1% in weight, plutonium, along with uranium and fission products. Current reprocessing proposals call for no separation of the pure plutonium. The work described in this paper studied, by computer simulation, the fundamental feasibility of preparing a 20% concentrated plutonium product solution from the 1% feed by adjusting only the feed rates and acid concentrations of the incoming streams and without the addition of redox reagents for the plutonium. A set of process design flowsheets has been developed to realize a concentrated plutonium solution of a 20% stream from the dilute plutonium feed without using redox reagents. (authors)

  9. Improvement in soil-plant-atmosphere modelling of {sup 14}C dynamics and the application of two models to data from a nuclear fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Limer, Laura M.C. [Quintessa Limited, 633/635 Birchwood Boulevard, WA3 7QU, Warrington (United Kingdom); Le Dizes-Maurel, Severine; Maro, Denis [Institut de Radioprotection et de Surete Nucleaire (IRSN), PRP-ENV, SERIS, LM2E, Cadarache, Saint-Paul Lez Durance (France); Klos, Ryk [Aleksandria Sciences Limited, S7 2DD, Sheffield (United Kingdom); Norden, Maria [Swedish Radiation Safety Authority, SE-171 16, Stockholm (Sweden)

    2014-07-01

    The need to address radiological impacts from {sup 14}C released to the biosphere has been recognised for some time. However, because of its role in biological processes and its ecological cycling, the standard methods employed to model long-term radionuclide transport and accumulation in the biosphere cannot be used satisfactorily for {sup 14}C. The degree of complexity in any {sup 14}C model used must be balanced against the availability of supporting data and the assessment context. In 2011, the model SSPAM14C was developed on behalf of the Swedish Radiation Safety Authority (SSM), with the intention to usage in both long-term and short-term release assessments (Limer et al., 2013). As part of the model testing it was applied to data collected during laboratory experiments performed by Imperial College London in the 1990's (Tucker and Shaw, 1997). Independently, IRSN has also been developing its own {sup 14}C model, TOCATTA (Le Dizes et al., 2012), and has previously tested it against field data collected by IRSN, between 2006 and 2008, in the vicinity of the La Hague nuclear fuel reprocessing plant in France (Aulagnier et al., 2012). The main conclusion drawn from these comparisons highlighted the need to develop an hourly time step model of {sup 14}C transfer based more thoroughly on knowledge arising from plant physiology, soil science and meteorology (Farquhar and von Caemmerer, 1982). These models have undergone further development, and have been applied here to the La Hague field data as it represents a medium term data set with both short term variation and a sizeable time series of measurements against which to compare the models. By increasing the temporal resolution of the IRSN model, a new version called TOCATTA-ccan simulate the impact of intermittent {sup 14}C releases occurring either the day or night (Aulagnier et al., 2013). Simplification of the soil sub-model in SSPAM14C is also shown to be justified for application to operational release

  10. Final report, Task 3: possible uses of the Nuclear Fuel Services, Inc. reprocessing plant at West Valley, New York. [For research on alternative fuel cycles, spiking, coprocessing, waste solidification, and recovery of radioactive gases

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-06-14

    The West Valley Plant could readily be used for work on reprocessing of alternative fuels, spiking, coprocessing (including CIVEX), waste solidification, and the recovery of radioactive gases. The plant could be easily modified for any scale between small-scale experimental work to production-scale demonstration, involving virtually any combination of fissile/fertile fuel materials that might be used in the future. The use of this plant for the contemplated experimental work would involve lower capital costs than the use of other facilities at DOE sites, except possibly for spiking of recovered products; the operating costs would be no greater than at other sites. The work on reprocessing of alternative fuels and coprocessing could commence within about one year; on recovery of radioactive gases, in 3 to 5 years; on spiking, in 4 years; and on waste solidification demonstration, in about 5 years. The contemplated work could be begun at this plant at least as early as at Barnwell, although work on spiking of recovered products could probably be started in existing hot cells earlier than at West Valley. (DLC)

  11. Consolidation of the EXAm process: towards the reprocessing of a concentrated PUREX raffinate

    Energy Technology Data Exchange (ETDEWEB)

    Vanel, V.; Bollesteros, M.J.; Marie, C.; Montuir, M.; Pacary, V.; Antegnard, F.; Costenoble, S.; Boyer-Deslys, V. [CEA Marcoule, Nuclear Energy Division, Radiochemistry and Processes Department, Bagnols-sur-Ceze, F-30207 (France)

    2016-07-01

    Recycling americium alone from the spent fuel is an important issue currently studied for the future nuclear cycle (Generation IV systems) as Am is one of the main contributors to the long-term radiotoxicity and heat power of final waste. The solvent extraction process called EXAm has been developed by the CEA to enable the recovery of Am alone from a PUREX raffinate (with U, Np and Pu already removed). A mixture of DMDOHEMA and HDEHP diluted in TPH is used as the solvent and the Am/Cm selectivity is improved using TEDGA as a selective complexing agent to maintain Cm and the heavier lanthanides in the acidic aqueous phase (HNO{sub 3} 5-6 M). Americium is then selectively stripped from the light lanthanides at low acidity (pH 2.5-3) with a poly-aminocarboxylic acid (DTPA). An additional step is necessary before Am recovery, in order to strip molybdenum which would otherwise be complexed by DTPA and contaminate the Am raffinate. In order to make the process and its associated future plant more compact, the objective is now to adapt the EXAm process to a concentrated raffinate. With a concentrated PUREX raffinate, the process operates under conditions close to saturation both for the solvent and complexing agent TEDGA during the Am extraction step. Consequently, some changes were needed to adapt the flowsheet to higher concentrations of cations. Before the test on a real PUREX raffinate in the CBP shielded line at ATALANTE (at the end of 2015), the EXAm flowsheet had to be consolidated and achievable target performances ensured. A series of experiments and tests was performed: on laboratory scale (batch experiments), to identify the good operating conditions and to simulate the main phenomena involved (2010-2014); first on an inactive surrogate feed solution at G1 facility (2011-2013), and then on a surrogate feed solution with trace amounts of americium and curium (spiked test) in the C17 shielded line at ATALANTE (2014). (authors)

  12. The eye movement desensitization and reprocessing procedure prevents defensive processing in health persuasion.

    Science.gov (United States)

    Dijkstra, Arie; van Asten, Regine

    2014-01-01

    In the present study, the method of eye movement desensitization and reprocessing (EMDR) is studied to understand and prevent defensive reactions with regard to a negatively framed message advocating fruit and vegetable consumption. EMDR has been shown to tax the working memory. Participants from a university sample (n = 124) listened to the persuasive message in a randomized laboratory experiment. In the EMDR condition, they were also instructed to follow with their eyes a dot on the computer screen. The dot constantly moved from one side of the screen to the other in 2 seconds. In addition, a self-affirmation procedure was applied in half of the participants. EMDR led to a significant increase in persuasion, only in recipients in whom the persuasive message could be expected to activate defensive self-regulation (in participants with a moderate health value and in participants with low self-esteem). In those with a moderate health value, EMDR increased persuasion, but only when recipients were not affirmed. In addition, EMDR increased persuasion only in recipients with low self-esteem, not in those with high self-esteem. These results showed that EMDR influenced persuasion and in some way lowered defensive reactions. The similarities and differences in effects of EMDR and self-affirmation further increased our insight into the psychology of defensiveness.

  13. Spent nuclear fuel reprocessing and international law. Germany`s obligations under international law in matters of spent fuel reprocessing and the relevant contracts concluded with France and the United Kingdom; Wiederaufarbeitung und Voelkerrecht. Die voelkerrechtlichen Verpflichtungen der Bundesrepublik Deutschland gegenueber der Franzoesischen Republik und dem Vereinigten Koenigreich auf dem Gebiet der Wiederaufarbeitung

    Energy Technology Data Exchange (ETDEWEB)

    Heintschel v. Heinegg, W. [Augsburg Univ. (Germany). Juristische Fakultaet

    1999-01-01

    The review presented is an excerpt from an expert opinion written by the author in December last year, in response to changes in nuclear energy policy announced by the new German government. The reprocessing of spent nuclear fuels from German power reactors in the reprocessing facilities of France (La Hague) and the UK (Sellafield) is not only based on contracts concluded by the German electric utilities and the French COGEMA or the British BNFL, but has been agreed as well by an exchange of diplomatic notes between the French Ministry of Foreign Affairs and the German ambassador in Paris, the German Foreign Ministry and the French ambassador as well as the British ambassador in Bonn. The article therefore first examines from the angle of international law the legal obligations binding the states involved, and Germany in particular, in matters of spent fuel reprocessing contracts. The next question arising in this context and discussed by the article is that of whether and how much indemnification can be demanded by the reprocessing companies, or their governments, resp., if Germany should discontinue spent fuel resprocessing and thus might be made liable for breach of the bilateral agreements. (orig/CB) [Deutsch] Der Beitrag enthaelt eine gekuerzte Zusammenfassung eines Gutachtens, das der Verfasser im Dezember 1998 erstellte. Anlass war die Ankuendigung der neuen deutschen Regierung, die Wiederaufarbeitung abgebrannter Kernbrennstoffe bald beenden zu wollen zugunsten der Zwischenlagerung und spaeteren Entsorgung. Die Wiederaufarbeitung deutscher Brennelemente im franzoesischen La Hague und im englischen Sellafield ist Gegenstand nicht allein der Vereinbarungen zwischen den deutschen Stromversorgern und der COGEMA sowie der BNFL, sondern auch von Notenwechseln zwischen dem franzoesischen Ministerium fuer Auswaertige Angelegenheiten und dem deutschen Botschafter in Paris, dem Auswaertigen Amt und dem franzoesischen Botschafter in Bonn, sowie dem Staatssekretaer im

  14. Plasma coal reprocessing

    Science.gov (United States)

    Messerle, V. E.; Ustimenko, A. B.

    2013-12-01

    Results of many years of investigations of plasma-chemical technologies for pyrolysis, hydrogenation, thermochemical preparation for combustion, gasification, and complex reprocessing of solid fuels and hydrocarbon gas cracking are represented. Application of these technologies for obtaining the desired products (hydrogen, industrial carbon, synthesis gas, valuable components of the mineral mass of coal) corresponds to modern ecological and economical requirements to the power engineering, metallurgy, and chemical industry. Plasma fuel utilization technologies are characterized by the short-term residence of reagents within a reactor and the high degree of the conversion of source substances into the desired products without catalyst application. The thermochemical preparation of the fuel to combustion is realized in a plasma-fuel system presenting a reaction chamber with a plasmatron; and the remaining plasma fuel utilization technologies, in a combined plasma-chemical reactor with a nominal power of 100 kW, whose zone of the heat release from an electric arc is joined with the chemical reaction zone.

  15. Microbial fuel cell treatment of fuel process wastewater

    Science.gov (United States)

    Borole, Abhijeet P; Tsouris, Constantino

    2013-12-03

    The present invention is directed to a method for cleansing fuel processing effluent containing carbonaceous compounds and inorganic salts, the method comprising contacting the fuel processing effluent with an anode of a microbial fuel ell, the anode containing microbes thereon which oxidatively degrade one or more of the carbonaceous compounds while producing electrical energy from the oxidative degradation, and directing the produced electrical energy to drive an electrosorption mechanism that operates to reduce the concentration of one or more inorganic salts in the fuel processing effluent, wherein the anode is in electrical communication with a cathode of the microbial fuel cell. The invention is also directed to an apparatus for practicing the method.

  16. The conceptual analysis of MBA and KMP for advanced spent fuel management process

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yoon; Ko, Won Il; Ha, Jang Ho; Kim, Ho Dong; Koo, Dae Seo [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    This report describes the concept of dry reprocessing of molten salt which is proposed as nuclear fuel cycle with nuclear proliferation resistance. These basic researches in Japan, U. S., Russia are in progress, and Republic of Korea is performing basic research of metallic conversion fabrication of molten salt of uranium dioxide fuels through nuclear research project. In this report, we have performed conceptual analysis and establishment of MBA and KMP for nuclear material safeguards in order to accomplish metallic conversion research of molten salt of uranium dioxide fuels. This report will contribute to the implementation of nuclear material safeguards of advanced spent fuel management process, and also the usage of basic data of nuclear material safeguards for spent fuel recycling process of native country. 11 refs., 17 figs., 8 tabs. (Author)

  17. Remote maintenance lessons learned'' on prototypical reprocessing equipment

    Energy Technology Data Exchange (ETDEWEB)

    Kring, C.T.; Schrock, S.L.

    1990-01-01

    Hardware representative of essentially every major equipment item necessary for reprocessing breeder reactor nuclear fuel has been installed and tested for remote maintainability. This testing took place in a cold mock-up of a remotely maintained hot cell operated by the Consolidated Fuel Reprocessing Program (CFRP) within the Fuel Recycle Division at Oak Ridge National Laboratory (ORNL). The reprocessing equipment tested included a Disassembly System, a Shear System, a Dissolver System, an Automated Sampler System, removable Equipment Racks on which various chemical process equipment items were mounted, and an advanced servomanipulator (ASM). These equipment items were disassembled and reassembled remotely by using the remote handling systems that are available within the cold mock-up area. This paper summarizes the lessons learned'' as a result of the numerous maintenance activities associated with each of these equipment items. 4 refs., 3 figs., 1 tab.

  18. Study of non aqueous reprocessing methods. Final progress report. [Container materials for pyrochemical processes

    Energy Technology Data Exchange (ETDEWEB)

    Teitel, R. J.; Luderer, J. E.; Henderson, T. M.

    1978-11-17

    The problems associated with container materials for selected pyrochemical processes and process containment conditions are reviewed. A rationale for container materials selection is developed. Candidate process container materials are presented, and areas warranting further development are identified. 14 tables.

  19. General Atomic Reprocessing Pilot Plant: engineering-scale dissolution system description

    Energy Technology Data Exchange (ETDEWEB)

    Yip, H.H.

    1979-04-01

    In February 1978, a dissolver-centrifuge system was added to the cold reprocessing pilot plant at General Atomic Company, which completed the installation of an HTGR fuel head-end reprocessing pilot plant. This report describes the engineering-scale equipment in the pilot plant and summarizes the design features derived from development work performed in the last few years. The dissolver operating cycles for both thorium containing BISO and uranium containinng WAR fissile fuels are included. A continuous vertical centrifuge is used to clarify the resultant dissolver product solution. Process instrumentation and controls for the system reflect design philosophy suitable for remote operation.

  20. A secondary fuel removal process: plasma processing

    Energy Technology Data Exchange (ETDEWEB)

    Min, J. Y.; Kim, Y. S. [Hanyang Univ., Seoul (Korea, Republic of); Bae, K. K.; Yang, M. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    Plasma etching process of UO{sub 2} by using fluorine containing gas plasma is studied as a secondary fuel removal process for DUPIC (Direct Use of PWR spent fuel Into Candu) process which is taken into consideration for potential future fuel cycle in Korea. CF{sub 4}/O{sub 2} gas mixture is chosen for reactant gas and the etching rates of UO{sub 2} by the gas plasma are investigated as functions of CF{sub 4}/O{sub 2} ratio, plasma power, substrate temperature, and plasma gas pressure. It is found that the optimum CF{sub 4}/O{sub 2} ratio is around 4:1 at all temperatures up to 400 deg C and the etching rate increases with increasing r.f. power and substrate temperature. Under 150W r.f. power the etching rate reaches 1100 monolayers/min at 400 deg C, which is equivalent to about 0.5mm/min. (author).

  1. Neural processing of emotions in traumatized children treated with Eye Movement Desensitization and Reprocessing therapy: a hdEEG study.

    Science.gov (United States)

    Trentini, Cristina; Pagani, Marco; Fania, Piercarlo; Speranza, Anna Maria; Nicolais, Giampaolo; Sibilia, Alessandra; Inguscio, Lucio; Verardo, Anna Rita; Fernandez, Isabel; Ammaniti, Massimo

    2015-01-01

    Eye Movement Desensitization and Reprocessing (EMDR) therapy has been proven efficacious in restoring affective regulation in post-traumatic stress disorder (PTSD) patients. However, its effectiveness on emotion processing in children with complex trauma has yet to be explored. High density electroencephalography (hdEEG) was used to investigate the effects of EMDR on brain responses to adults' emotions on children with histories of early maltreatment. Ten school-aged children were examined before (T0) and within one month after the conclusion of EMDR (T1). hdEEGs were recorded while children passively viewed angry, afraid, happy, and neutral faces. Clinical scales were administered at the same time. Correlation analyses were performed to detect brain regions whose activity was linked to children's traumatic symptom-related and emotional-adaptive problem scores. In all four conditions, hdEEG showed similar significantly higher activity on the right medial prefrontal and fronto-temporal limbic regions at T0, shifting toward the left medial and superior temporal regions at T1. Moreover, significant correlations were found between clinical scales and the same regions whose activity significantly differed between pre- and post-treatment. These preliminary results demonstrate that, after EMDR, children suffering from complex trauma show increased activity in areas implicated in high-order cognitive processing when passively viewing pictures of emotional expressions. These changes are associated with the decrease of depressive and traumatic symptoms, and with the improvement of emotional-adaptive functioning over time.

  2. Neural processing of emotions in traumatized children treated with Eye Movement Desensitization and Reprocessing therapy: A hdEEG study

    Directory of Open Access Journals (Sweden)

    Cristina eTrentini

    2015-11-01

    Full Text Available Eye Movement Desensitization and Reprocessing (EMDR therapy has been proven efficacious in restoring affective regulation in Post–Traumatic Stress Disorder (PTSD patients. However, its effectiveness on emotion processing in children with complex trauma has yet to be explored. High density Electroencephalography (hdEEG was used to investigate the effects of EMDR on brain responses to adults’ emotions on children with histories of early maltreatment. Ten school–aged children were examined before (T0 and within one month after the conclusion of EMDR (T1. hdEEGs were recorded while children passively viewed angry, afraid, happy, and neutral faces. Clinical scales were administered at the same time. Correlation analyses were performed to detect brain regions whose activity was linked to children’s traumatic symptom–related and emotional–adaptive problem scores. In all four conditions, hdEEG showed similar significantly higher activity on the right medial prefrontal and fronto–temporal limbic regions at T0, shifting towards the left medial and superior temporal regions at T1. Moreover, significant correlations were found between clinical scales and the same regions whose activity significantly differed between pre– and post–treatment. These preliminary results demonstrate that, after EMDR, children suffering from complex trauma show increased activity in areas implicated in high–order cognitive processing when passively viewing pictures of emotional expressions. These changes are associated with the decrease of depressive and traumatic symptoms, and with the improvement of emotional–adaptive functioning over time.

  3. 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

  4. Determination of Zr{sup 93} and Mo{sup 93} in reprocessing effluents of spent fuels; Determination des radionucleides zirconium 93 et molybdene 93 dans des effluents de retraitement des combustibles irradies

    Energy Technology Data Exchange (ETDEWEB)

    Puech, P.; Bienvenu, Ph. [CEA Cadarache, Dept. d' Entreposage et de Stockage des Dechets, 13 - Saint-Paul-lez-Durance (France)

    2001-07-01

    In this work is presented the approach undertaken and the results obtained within the context of a study carried out with COGEMA on the quantification of two long lived isotopes: {sup 93}Zr and {sup 93}Mo contained in spent fuel reprocessing units effluents. The quantity of long lived radionuclides contained in nuclear wastes is indeed a very important parameter for surface storages and a determining one for underground storages. (O.M.)

  5. Transfer of conservative and non-conservative radionuclides from the Sellafield nuclear fuel reprocessing plant to the coastal waters of Ireland

    Energy Technology Data Exchange (ETDEWEB)

    Mcmahon, C.A.; Fegan, M.; Wong, J.; Long, S.C.; Mckittrick, L.; Thomas, K.; Rafferty, B. [Radiological Protection Institute of Ireland, Dublin (Ireland)

    2004-07-01

    The Radiological Protection Institute of Ireland has monitored levels of anthropogenic radionuclides in the Irish marine environment for over 20 years. While the primary objective of the monitoring programme is to assess the exposure of the Irish population resulting from the presence of these radionuclides in the marine environment, the programme also aims to assess the geographical distribution and temporal variations of the radionuclides. The programme involves the routine sampling of and testing for radioactivity in fish, shellfish, seaweed, sediments and seawater. The data generated in the course of this programme, as well as in a separate study of changing plutonium isotopic ratios in Fucus vesiculosus from the west coast of Ireland, are used in this paper to estimate transport times from the Sellafield nuclear fuel reprocessing plant to the western Irish Sea and from the Irish Sea to the west coast of Ireland. The results obtained are discussed in the paper and the transfer times estimated for particle-reactive radionuclides (plutonium isotopes) compared with those obtained for more conservative radionuclides ({sup 137}Cs and {sup 99}Tc). Transfer factors (calculated as the ratio between observed concentrations in the environment and an average discharge rate {tau} years earlier, where {tau} is the transport time) are also presented. (author)

  6. Fuel salt reprocessing influence on the MSFR behavior and on its associated reprocessing unit; Influence du retraitement physico-chimique du sel combustible sur le comportement du MSFR et sur le dimensionnement de son unite de retraitement

    Energy Technology Data Exchange (ETDEWEB)

    Doligez, X.

    2010-10-15

    In order to face with the growing of the energy demand, the nuclear industry has to reach the fourth generation technology. Among those concept, molten salt reactor, and especially the fast neutron spectrum configuration, seems very promising: indeed breeding is achievable while the feedback coefficient are still negative. However, the reprocessing salt scheme is not totally set down yet. A lot of uncertainties remain on chemical properties of the salt. Thanks to numerical simulation we studied the behavior of the molten Salt Fast Reactor coupled to a nominal reprocessing unit. We are now able to determine heat transfer and radiation in each elementary step of the unit and, by this way determine those that need special study for radioprotection. We also studied which elements are fundamental to extract for the reactor operation. Finally, we present a sensibility analysis of the chemical uncertainties to few relevant properties of the reactor behavior. (author)

  7. Evaluation of the Use of Synroc to Solidify the Cesium and Strontium Separations Product from Advanced Aqueous Reprocessing of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Julia Tripp; Vince Maio

    2006-03-01

    This report is a literature evaluation on the Synroc process for determining the potential for application to solidification of the Cs/Sr strip product from advanced aqueous fuel separations activities.

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

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

  9. Hydroxylamine as a potential reagent for dissolution off gas scrubbing in spent fuel reprocessing: kinetics of the iodine reduction. An example of similarity between the studies on the chemistry of iodine in reactor safety and in spent fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Cau Dit Coumes, C.; Devisme, F. [Commissariat a l`Energie Atomique, CE/VRH, Bagnols-sur-Ceze (France); Vargas, S.; Chopin-Dumas, J. [Laboratoire d`Electrochimie Inorganique, ENSSPICAM, Marseille (France)

    1996-12-01

    Iodine, which can be released inside the containment building when an accident occurs, can be traced, in normal operating conditions, at the back end of the fuel cycle. Hydroxylamine has been selected as a reagent of potential interest to trap iodine in the dissolution off gas treatment. The kinetics of the reaction between hydroxylamine and iodine has been studied in a narrow range of pH (1-2), with hydroxylamine in excess (ratios of hydroxylamine to iodine initial concentrations varying from 2 to 40), at constant temperature (30{sup o}C) and ionic strength (0.1 mol/L). Spectrophotometry and voltametry have been coupled for analytical investigation. The problem of the rapid mixing of the reactants has been solved using a continuous reactor. Triiodide has been shown non reactive towards hydroxylamine. An initial rate law has been proposed, pointing out the first order of the reaction with respect to hydroxylamine and iodine, and the inhibitory effect of iodide and hydrogen ions. Nitrous acid has been identified as a transitory product. Nitrous oxide and nitrogen monoxide have been detected by gas chromatography, the ratio of the amounts of products formed depending on acidity. The complexity of the overall reaction has been ascribed to the competition of four reactions as previously proposed in the literature. (author) 8 figs., 1 tab., 13 refs.

  10. Dry Process Fuel Performance Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

    2005-04-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

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

    Energy Technology Data Exchange (ETDEWEB)

    Davis, W. Jr.; Blanco, R.E.; Finney, B.C.; Hill, G.S.; Moore, R.E.; Witherspoon, J.P.

    1976-05-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model high-temperature gas-cooled reactor (HTGR) fuel reprocessing plant and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist the U. S. Nuclear Regulatory Commission in defining the term as low as reasonably achievable as it applies to this nuclear facility. The base case is representative of conceptual, developing technology of head-end graphite-burning operations and of extensions of solvent-extraction technology of current designs for light-water-reactor (LWR) fuel reprocessing plants. The model plant has an annual capacity of 450 metric tons of heavy metal (MTHM, where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base-case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods used in the case studies is discussed.

  12. Size Design of CdZnTe Detector Shield for Measuring Burnup of Spent Fuel

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>It is important to measure the burnup of spent fuel for nuclear safeguards, burnup credit and critical safety in spent-fuel reprocessing process. The purpose of this work is designing a portable device to

  13. Dry process fuel performance technology development

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kweon Ho; Kim, K. W.; Kim, B. K. (and others)

    2006-06-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

  14. Accident Safety Analysis Method Study for Spent Fuel Reprocessing Plant%乏燃料后处理厂事故安全分析方法的探讨

    Institute of Scientific and Technical Information of China (English)

    李锐柔; 徐云起

    2012-01-01

    According to some relative documents (like NRC, DOE and IAEA documents et al. ), and considering the experience and practical technology level of the safety analysis of spent fuel reprocess-ing plant in China, this article suggested that the risk assessment, which combining both deterministic ap-proach and probabilistic safety analysis, could be applied to the accident safety analysis of spent fuel repro-cessing plant in China. Meanwhile, the coordinated working procedure was also proposed.%参考了NRC、DOE、IAEA等相关文件,结合我国乏燃料后处理厂安全分析的经验和实际技术水平,建议我国乏燃料后处理厂在事故安全分析中可采用确定论和概率安全分析相结合的风险评价方法,并提出了相应的工作流程。

  15. Hydroxylamine a potential reagent for dissolution off gas scrubbing in nuclear spent fuel reprocessing: kinetics of the iodine reduction

    Energy Technology Data Exchange (ETDEWEB)

    Cau Dit Coumes, C.; Devisme, F. [CEA Centre d`Etudes de la Vallee du Rhone, 30 - Marcoule (France). Dept. d`Exploitation du Retraitement et de Demantelement; Chopin, J.; Vargas, S.

    1996-12-31

    Iodine, which can be released inside the containment buildings when accident occurs, can be traced, in normal operating conditions, at the back end of the fuel cycle. Hydroxylamine has been selected as a regent of potential interest to trap iodine in the dissolution off gas treatment. The kinetics of the reaction between hydroxylamine and iodine has been studied in a narrow range of pH(1-2), with hydroxylamine in excess (ratios of hydroxylamine to iodine initial concentrations varying from 2 to 40), at constant temperature (30 deg. C) and ionic strength (0.1 mol/l). Spectrophotometry and voltametry have been coupled for analytical solved using a investigation. The problem of the rapid mixing of the reactants has been solved using a continuous reactor. Tri-iodine has been shown non reactive towards hydroxylamine. An initial rate law have been proposed, pointing out the first order of the reaction with respect to hydroxylamine and iodine, and the inhibitory effect and hydrogen ions. Nitrous acid has been identified as a transitory product. Nitrous oxide and nitrogen monoxide have been detected by gas chromatography, the ratio of the amounts of products formed depending on acidity. The complexity of the overall reaction has been ascribed to the competition of hour reactions (I{sub 2} + I <=> I{sub 3}{sup -}NH{sub 3}OH{sup +} + 2 I{sub 2} + H{sub 2}O ->HNO{sub 2} + 4 I{sup -} + 5 H{sup +}; NH{sub 3}OH{sup +} + HNO{sub 2} -> N{sub 2}O + 2 H{sub 2}O + H-+ 2HNO{sub 2} + 2 I{sup -} + 2H-+ -> 2 NO + I{sub 2} + H{sub 2}O). (authors). 12 refs.

  16. Data validation and security for reprocessing.

    Energy Technology Data Exchange (ETDEWEB)

    Tolk, Keith Michael; Merkle, Peter Benedict; DurÔan, Felicia Angelica; Cipiti, Benjamin B.

    2008-10-01

    Next generation nuclear fuel cycle facilities will face strict requirements on security and safeguards of nuclear material. These requirements can result in expensive facilities. The purpose of this project was to investigate how to incorporate safeguards and security into one plant monitoring system early in the design process to take better advantage of all plant process data, to improve confidence in the operation of the plant, and to optimize costs. An existing reprocessing plant materials accountancy model was examined for use in evaluating integration of safeguards (both domestic and international) and security. International safeguards require independent, secure, and authenticated measurements for materials accountability--it may be best to design stand-alone systems in addition to domestic safeguards instrumentation to minimize impact on operations. In some cases, joint-use equipment may be appropriate. Existing domestic materials accountancy instrumentation can be used in conjunction with other monitoring equipment for plant security as well as through the use of material assurance indicators, a new metric for material control that is under development. Future efforts will take the results of this work to demonstrate integration on the reprocessing plant model.

  17. Fuel quality processing study, volume 1

    Science.gov (United States)

    Ohara, J. B.; Bela, A.; Jentz, N. E.; Syverson, H. T.; Klumpe, H. W.; Kessler, R. E.; Kotzot, H. T.; Loran, B. L.

    1981-01-01

    A fuel quality processing study to provide a data base for an intelligent tradeoff between advanced turbine technology and liquid fuel quality, and also, to guide the development of specifications of future synthetic fuels anticipated for use in the time period 1985 to 2000 is given. Four technical performance tests are discussed: on-site pretreating, existing refineries to upgrade fuels, new refineries to upgrade fuels, and data evaluation. The base case refinery is a modern Midwest refinery processing 200,000 BPD of a 60/40 domestic/import petroleum crude mix. The synthetic crudes used for upgrading to marketable products and turbine fuel are shale oil and coal liquids. Of these syncrudes, 50,000 BPD are processed in the existing petroleum refinery, requiring additional process units and reducing petroleum feed, and in a new refinery designed for processing each syncrude to produce gasoline, distillate fuels, resid fuels, and turbine fuel, JPGs and coke. An extensive collection of synfuel properties and upgrading data was prepared for the application of a linear program model to investigate the most economical production slate meeting petroleum product specifications and turbine fuels of various quality grades. Technical and economic projections were developed for 36 scenarios, based on 4 different crude feeds to either modified existing or new refineries operated in 2 different modes to produce 7 differing grades of turbine fuels. A required product selling price of turbine fuel for each processing route was calculated. Procedures and projected economics were developed for on-site treatment of turbine fuel to meet limitations of impurities and emission of pollutants.

  18. Process and apparatus for burning solid fuel

    NARCIS (Netherlands)

    Lin, W.; Van den Bleek, C.M.

    1995-01-01

    Abstract of NL 9301828 (A) Described is a process for burning solid fuel, in which nitrogen in the form of NH3 is released from said fuel, for example by gasification, said NH3 being excluded from the combustion process but being admixed, together with CO likewise released, to the gases released

  19. Fluidized-Solid-Fuel Injection Process

    Science.gov (United States)

    Taylor, William

    1992-01-01

    Report proposes development of rocket engines burning small grains of solid fuel entrained in gas streams. Main technical discussion in report divided into three parts: established fluidization technology; variety of rockets and rocket engines used by nations around the world; and rocket-engine equation. Discusses significance of specific impulse and ratio between initial and final masses of rocket. Concludes by stating three important reasons to proceed with new development: proposed engines safer; fluidized-solid-fuel injection process increases variety of solid-fuel formulations used; and development of fluidized-solid-fuel injection process provides base of engineering knowledge.

  20. An updated interpretation of the Hanö Bay Basin, Baltic Sea, based on recently re-processed vintage 2D seismic data

    Science.gov (United States)

    Bell, Nicholas; Sopher, Daniel; Juhlin, Christopher

    2014-05-01

    The Hanö Bay Basin is a relatively small, tectonically controlled, Mesozoic basin in the SW Baltic Sea, Northern Europe. In this study a new seismic interpretation has been made of the basin based on re-processed vintage 2D marine seismic data. A large dataset acquired between 1970 and 1984 by Oljeprospektering AB (OPAB) containing seismic lines across the Hanö Bay Basin has recently been made available by the Swedish Geological Survey (SGU). Seismic interpretation studies within the Hanö Bay Basin were last conducted in the mid-1990's. Since this time, computer power and seismic processing methods have advanced. Re-processing of a grid of lines across the Hanö Bay Basin has allowed updated interpretations to be made which more accurately reflect the geological history of the area. Multi channel seismic data from four surveys within the OPAB dataset: NA79, D72, W70 and EA73, along with two wells H1 and H4, were used in this study. An updated interpretation of the pre-Cambrian basement, which exhibits a distinctive, sharply undulating morphology, was undertaken. The basement horizon across parts of the Hanö Bay appears to be very rugose, containing a number of distinctive troughs and peaks that are over 50m in amplitude. Within these basement troughs a set of distinct packages of sediment is observed. These packages are discontinuous and are most prevalent in a small circular area in the central section of the study area. The age of these sediment packages is uncertain, being either early Mesozoic or the erosional remnants of older Paleozoic sediments. Interpretations of the re-processed seismic data indicate, in some areas, that basin fill has occurred in a significantly different way to previous interpretations during the Mesozoic. The model proposed in this study takes into account normal movement on the Christiansø Fault prior to Cretaceous inversion.

  1. The Photometric Calibration of the Dark Energy Survey (DES): Results from the Summer 2013 Re-processing of the DES Science Verification Data

    Science.gov (United States)

    Tucker, Douglas L.; Allam, S. S.; Annis, J. T.; Armstrong, R.; Bauer, A.; Bernstein, G.; Burke, D.; Fix, M.; Foust, W.; Gruendl, R. A.; Head, H.; Kuehn, K.; Kuhlmann, S.; Li, T.; Lin, H.; Rykoff, E. S.; Smith, J.; Wester, W.; Wyatt, S.; Yanny, B.; Energy Survey, Dark

    2014-01-01

    The Dark Energy Survey (DES) -- a five-year 5000 sq deg grizY survey of the Southern sky to probe the parameters of dark energy -- recently began operations using the new 3 sq deg DECam imager on the Blanco 4m telescope at the Cerro Tololo Interamerican Observatory. In order to achieve its science goals, the DES has tight requirements on both its relative and absolute photometric calibrations. The 5-year requirements are (1) an internal (relative) photometric calibration of 2% rms (2) an absolute color calibration of 0.5%, and (3) an absolute flux calibration of 0.5% (in i-band relative to BD+17 4708). In preparation for DES operations, the instrument+telescope underwent a period of Science Verification between November 2012 and February 2013. These Science Verification (SV) data were quickly processed to determine whether the image data were being produced with sufficient quality and efficiency to meet DES science goals. These data were also useful for initial science, and they were re-processed and re-calibrated during Summer 2013. The photometric goals for Summer 2013 re-processing of the DES SV were intentionally more relaxed than the requirements for the final 5-year survey: (1) an all-sky internal (relative) calibration goal of 3%, (2) an absolute color goal of 3%, and (3) an absolute flux goal of 3%. Here, we describe the results from the photometric calibration of the Summer 2013 re-processing of the DES SV data, the lessons learned, and plans for the future.

  2. Fuel Conditioning Facility Electrorefiner Process Model

    Energy Technology Data Exchange (ETDEWEB)

    DeeEarl Vaden

    2005-10-01

    The Fuel Conditioning Facility at the Idaho National Laboratory processes spent nuclear fuel from the Experimental Breeder Reactor II using electro-metallurgical treatment. To process fuel without waiting for periodic sample analyses to assess process conditions, an electrorefiner process model predicts the composition of the electrorefiner inventory and effluent streams. For the chemical equilibrium portion of the model, the two common methods for solving chemical equilibrium problems, stoichiometric and non stoichiometric, were investigated. In conclusion, the stoichiometric method produced equilibrium compositions close to the measured results whereas the non stoichiometric method did not.

  3. Mathematical modeling of biomass fuels formation process.

    Science.gov (United States)

    Gaska, Krzysztof; Wandrasz, Andrzej J

    2008-01-01

    The increasing demand for thermal and electric energy in many branches of industry and municipal management accounts for a drastic diminishing of natural resources (fossil fuels). Meanwhile, in numerous technical processes, a huge mass of wastes is produced. A segregated and converted combustible fraction of the wastes, with relatively high calorific value, may be used as a component of formed fuels. The utilization of the formed fuel components from segregated groups of waste in associated processes of co-combustion with conventional fuels causes significant savings resulting from partial replacement of fossil fuels, and reduction of environmental pollution resulting directly from the limitation of waste migration to the environment (soil, atmospheric air, surface and underground water). The realization of technological processes with the utilization of formed fuel in associated thermal systems should be qualified by technical criteria, which means that elementary processes as well as factors of sustainable development, from a global viewpoint, must not be disturbed. The utilization of post-process waste should be preceded by detailed technical, ecological and economic analyses. In order to optimize the mixing process of fuel components, a mathematical model of the forming process was created. The model is defined as a group of data structures which uniquely identify a real process and conversion of this data in algorithms based on a problem of linear programming. The paper also presents the optimization of parameters in the process of forming fuels using a modified simplex algorithm with a polynomial worktime. This model is a datum-point in the numerical modeling of real processes, allowing a precise determination of the optimal elementary composition of formed fuels components, with assumed constraints and decision variables of the task.

  4. Fuel corrosion processes under waste disposal conditions

    Energy Technology Data Exchange (ETDEWEB)

    Shoesmith, D.W. [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada)

    1999-09-01

    Under the oxidizing conditions likely to be encountered in the Yucca Mountain Repository, fuel dissolution is a corrosion process involving the coupling of the anodic dissolution of the fuel with the cathodic reduction of oxidants available within the repository. The oxidants potentially available to drive fuel corrosion are environmental oxygen, supplied by the transport through the permeable rock of the mountain and molecular and radical species produced by the radiolysis of available aerated water. The mechanism of these coupled anodic and cathodic reactions is reviewed in detail. While gaps in understanding remain, many kinetic features of these reactions have been studied in considerable detail, and a reasonably justified mechanism for fuel corrosion is available. The corrosion rate is determined primarily by environmental factors rather than the properties of the fuel. Thus, with the exception of increase in rate due to an increase in surface area, pre-oxidation of the fuel has little effect on the corrosion rate.

  5. Gloves Reprocessing: Does It Really Save Money?

    Science.gov (United States)

    Arora, Pankaj; Kumari, Santosh; Sodhi, Jitender; Talati, Shweta; Gupta, Anil Kumar

    2015-12-01

    Gloves are reprocessed and reused in health-care facilities in resource-limited settings to reduce the cost of availability of gloves. The study was done with the aim to compute the cost of reprocessing of gloves so that an economically rationale decision can be taken. A retrospective record-based cross-sectional study was undertaken in a central sterile supply department where different steps during reprocessing of gloves were identified and the cost involved in reprocessing per pair of gloves was calculated. The cost of material and manpower was calculated to arrive at the cost of reprocessing per pair of gloves. The cost of a reprocessed pair of surgical gloves was calculated to be Indian Rupee (INR) 14.33 which was greater than the cost of a new pair of disposable surgical gloves (INR 9.90) as the cost of sterilization of one pair of gloves itself came out to  be INR 10.97. The current study showed that the purchase of sterile disposable single-use gloves is cheaper than the process of recycling. Reprocessing of gloves is not economical on tangible terms even in resource-limited settings, and from the perspective of better infection control as well as health-care worker safety, it further justifies the use of disposable gloves.

  6. Carbon oxides free fuel processing for fuel cell applications

    Science.gov (United States)

    Choudhary, Tushar V.

    Fuel processing represents a very important aspect of fuel cell technology. The widespread utilization of fuel cells will only be possible if CO x-free hydrogen producing technologies are developed. Towards this objective, step-wise reforming of hydrocarbons and catalytic decomposition of ammonia were investigated for hydrogen production. Also, novel Au-based catalysts were synthesized for preferentially eliminating CO in the presence of excess hydrogen. The step-wise reforming of hydrocarbons was investigated for production of CO-free hydrogen for proton exchange membrane fuel cells. Proof of concept pulse reactor experiments employing Ni-based catalysts clearly showed the feasibility of the cyclic step-wise reforming process for clean hydrogen production. Under optimum conditions the CO content in the hydrogen was found to be less than 20 ppm by this process (a large amount of CO is obtained as a by-product from conventional methods of hydrogen production). The step-wise reforming process thus greatly simplifies fuel reforming, as expensive and circuitous post-reforming hydrogen purification processes are eliminated. The process was profoundly influenced by the operating temperature, space velocity and nature of the catalyst support. Catalytic ammonia decomposition was investigated for COx-free hydrogen production for alkaline fuel cells. These studies revealed that Ru, Ir and Ni-based catalysts were active for the process with Ru being the most active and Ni the least. The catalyst supports played a decisive role in determining the ammonia decomposition activity. Partial pressure dependence studies of the reaction rate on model Ir (100) catalysts yielded a positive order (0.9 +/- 0.l) with respect to ammonia and negative order (-0.7 +/- 0.l) with respect to hydrogen. The negative order with respect to hydrogen was attributed to the enhancement in the reverse of the ammonia decomposition reaction in the presence of surface hydrogen atoms. Novel nano-Au catalysts

  7. Report on design and technical standard planning of vibration controlling structure on the buildings, in the Tokai Reprocessing Facility, Power Reactor and Nuclear Fuel Development Corporation

    Energy Technology Data Exchange (ETDEWEB)

    Uryu, Mitsuru; Terada, Shuji; Shinohara, Takaharu; Yamazaki, Toshihiko; Nakayama, Kazuhiko [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works; Kondo, Toshinari; Hosoya, Hisashi

    1997-10-01

    The Tokai reprocessing facility buildings are constituted by a lower foundation, vibration controlling layers, and upper structure. At the vibration controlling layer, a laminated rubber aiming support of the building load and extension of the eigenfrequency and a damper aiming absorption of earthquake energy are provided. Of course, the facility buildings are directly supported at the arenaceous shale (Taga Layer) of the Miocene in the Neogene confirmed to the stablest ground, as well the buildings with high vibration resistant importance in Japan. This report shows that when the vibration controlling structure is adopted for the reprocessing facility buildings where such high vibration resistance is required, reduction of input acceleration for equipments and pipings can be achieved and the earthquake resistant safety can also be maintained with sufficient tolerance and reliability. (G.K.)

  8. Report on design and technical standard planning of vibration controlling structure on the buildings, in the Tokai Reprocessing Facility, Power Reactor and Nuclear Fuel Development Corporation

    Energy Technology Data Exchange (ETDEWEB)

    Uryu, Mitsuru; Terada, Shuji; Shinohara, Takaharu; Yamazaki, Toshihiko; Nakayama, Kazuhiko [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works; Kondo, Toshinari; Hosoya, Hisashi

    1997-10-01

    The Tokai reprocessing facility buildings are constituted by a lower foundation, vibration controlling layers, and upper structure. At the vibration controlling layer, a laminated rubber aiming support of the building load and extension of the eigenfrequency and a damper aiming absorption of earthquake energy are provided. Of course, the facility buildings are directly supported at the arenaceous shale (Taga Layer) of the Miocene in the Neogene confirmed to the stablest ground, as well the buildings with high vibration resistant importance in Japan. This report shows that when the vibration controlling structure is adopted for the reprocessing facility buildings where such high vibration resistance is required, reduction of input acceleration for equipments and pipings can be achieved and the earthquake resistant safety can also be maintained with sufficient tolerance and reliability. (G.K.)

  9. Some Thinking of Nuclear Fuel Reprocessing/Recycling in China%关于我国核燃料后处理/再循环的一些思考

    Institute of Scientific and Technical Information of China (English)

    顾忠茂; 柴之芳

    2011-01-01

    Based on the uranium-plutonium fuel cycle,the once-through fuel cycle,closed fuel cycle for thermal reactors and closed fuel cycle for fast reactors are analyzed and compared from the view point of sustainable development of nuclear fission energy.It is pointed out that both the once-through fuel cycle and the closed fuel cycle of thermal reactor could not meet the strategic needs of sustainable development of nuclear energy and fast reactor with multi-recycling of nuclear fuel is the best option to develop the nuclear fission energy in a sustainable way.The present status and the major trend of reprocessing/recycling technologies in the world are introduced and the gap between China and the major nuclear energy countries is evaluated.Keeping the domestic situation in mind,we try to explore the general considerations of the development of nuclear fuel reprocessing/recycling technologies in China and the technical options to be taken.The key techniques to be solved and the associated supporting measures are also proposed.%本文从核裂变能可持续发展的角度,对基于铀-钚循环的核燃料循环体系中的"一次通过"循环、热堆闭式循环和快堆闭式循环的特点进行分析和比较,指出"一次通过"循环和热堆闭式循环均不能满足核能可持续发展的战略需要,快堆及核燃料多次循环才是我国核裂变能可持续发展的根本出路;介绍了国际上核燃料后处理/再循环的技术现状和主流发展趋势,指出我国在后处理/再循环技术方面与国际先进水平之间的差距;结合我国国情探讨我国核燃料后处理/再循环技术发展的总体构想和拟采取的技术路线,提出需要突破的关键技术问题和相应的配套措施。

  10. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  11. SOFC system with integrated catalytic fuel processing

    Energy Technology Data Exchange (ETDEWEB)

    Finnerty, C.; Tompsett, G.A.; Kendall, K.; Ormerod, R.M. [Birchall Centre for Inorganic Chemistry and Materials Science, Keele Univ. (United Kingdom)

    2000-03-01

    In recent years, there has been much interest in the development of solid oxide fuel cell technology operating directly on hydrocarbon fuels. The development of a catalytic fuel processing system, which is integrated with the solid oxide fuel cell (SOFC) power source is outlined here. The catalytic device utilises a novel three-way catalytic system consisting of an in situ pre-reformer catalyst, the fuel cell anode catalyst and a platinum-based combustion catalyst. The three individual catalytic stages have been tested in a model catalytic microreactor. Both temperature-programmed and isothermal reaction techniques have been applied. Results from these experiments were used to design the demonstration SOFC unit. The apparatus used for catalytic characterisation can also perform in situ electrochemical measurements as described in previous papers [C.M. Finnerty, R.H. Cunningham, K. Kendall, R.M. Ormerod, Chem. Commun. (1998) 915-916; C.M. Finnerty, N.J. Coe, R.H. Cunningham, R.M. Ormerod, Catal. Today 46 (1998) 137-145]. This enabled the performance of the SOFC to be determined at a range of temperatures and reaction conditions, with current output of 290 mA cm{sup -2} at 0.5 V, being recorded. Methane and butane have been evaluated as fuels. Thus, optimisation of the in situ partial oxidation pre-reforming catalyst was essential, with catalysts producing high H{sub 2}/CO ratios at reaction temperatures between 873 K and 1173 K being chosen. These included Ru and Ni/Mo-based catalysts. Hydrocarbon fuels were directly injected into the catalytic SOFC system. Microreactor measurements revealed the reaction mechanisms as the fuel was transported through the three-catalyst device. The demonstration system showed that the fuel processing could be successfully integrated with the SOFC stack. (orig.)

  12. SOFC system with integrated catalytic fuel processing

    Science.gov (United States)

    Finnerty, Caine; Tompsett, Geoff. A.; Kendall, Kevin; Ormerod, R. Mark

    In recent years, there has been much interest in the development of solid oxide fuel cell technology operating directly on hydrocarbon fuels. The development of a catalytic fuel processing system, which is integrated with the solid oxide fuel cell (SOFC) power source is outlined here. The catalytic device utilises a novel three-way catalytic system consisting of an in situ pre-reformer catalyst, the fuel cell anode catalyst and a platinum-based combustion catalyst. The three individual catalytic stages have been tested in a model catalytic microreactor. Both temperature-programmed and isothermal reaction techniques have been applied. Results from these experiments were used to design the demonstration SOFC unit. The apparatus used for catalytic characterisation can also perform in situ electrochemical measurements as described in previous papers [C.M. Finnerty, R.H. Cunningham, K. Kendall, R.M. Ormerod, Chem. Commun. (1998) 915-916; C.M. Finnerty, N.J. Coe, R.H. Cunningham, R.M. Ormerod, Catal. Today 46 (1998) 137-145]. This enabled the performance of the SOFC to be determined at a range of temperatures and reaction conditions, with current output of 290 mA cm -2 at 0.5 V, being recorded. Methane and butane have been evaluated as fuels. Thus, optimisation of the in situ partial oxidation pre-reforming catalyst was essential, with catalysts producing high H 2/CO ratios at reaction temperatures between 873 K and 1173 K being chosen. These included Ru and Ni/Mo-based catalysts. Hydrocarbon fuels were directly injected into the catalytic SOFC system. Microreactor measurements revealed the reaction mechanisms as the fuel was transported through the three-catalyst device. The demonstration system showed that the fuel processing could be successfully integrated with the SOFC stack.

  13. The reprocessing of irradiated fuels by halides and their compounds; Le traitement des combustibles irradies par les halogenes et leurs composes

    Energy Technology Data Exchange (ETDEWEB)

    Bourgeois, M.; Faugeras, P. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    A brief description is given of the experiments leading to the choice of the process volatilization of fluorides by gas phase attack. The chemical process is described for certain current types of clad Fuels: the aluminium or the zirconium cladding is first volatilized as chloride by attack with gaseous hydrogen chloride. The uranium is then transformed into volatile hexafluoride by attack with fluorine. These reactions are carried out consecutively in the same reactor in the presence of a fluidized bed of alumina which facilitates heat exchange. The experiments have been carried out in quantities from 100 gms to several kilograms of fuel, first without activity, and then with tracers. A description is given of the laboratory research which was carried out simultaneously on the separation of uranium and plutonium fluorides. Finally, an apparatus is described which is intended to test the process on irradiated fuel at an activity level of several thousands of curies of fission products. (authors) [French] On rappelle brievement les experimentations qui nous ont permis de decider du procede adopte volatilisation des fluorures par attaque en phase gazeuse. On decrit le processus chimique pour certains types courants de combustibles Gaines: dans un premier stade, l'aluminium ou le zirconium est volatilise sous forme de chlorure par action de l'acide chlorhydrique. Ensuite, l'uranium est transforme en hexafluorure volatil par action du fluor. Ces operations se font successivement dans un meme reacteur, en presence d'un lit fluidise d'alumine qui a pour but de faciliter les echanges thermiques. L'experimentation a ete conduite sur des quantites allant de 100 g a plusieurs kg de combustibles, en inactif, puis avec des traceurs. On decrit les etudes de laboratoire menees parallelement sur la separation des fluorures d'uranium et de plutonium. Enfin, on decrit une installation en construction destinee a experimenter le procede sur

  14. Alternative Fuel for Portland Cement Processing

    Energy Technology Data Exchange (ETDEWEB)

    Schindler, Anton K; Duke, Steve R; Burch, Thomas E; Davis, Edward W; Zee, Ralph H; Bransby, David I; Hopkins, Carla; Thompson, Rutherford L; Duan, Jingran; ; Venkatasubramanian, Vignesh; Stephen, Giles

    2012-06-30

    The production of cement involves a combination of numerous raw materials, strictly monitored system processes, and temperatures on the order of 1500 °C. Immense quantities of fuel are required for the production of cement. Traditionally, energy from fossil fuels was solely relied upon for the production of cement. The overarching project objective is to evaluate the use of alternative fuels to lessen the dependence on non-renewable resources to produce portland cement. The key objective of using alternative fuels is to continue to produce high-quality cement while decreasing the use of non-renewable fuels and minimizing the impact on the environment. Burn characteristics and thermodynamic parameters were evaluated with a laboratory burn simulator under conditions that mimic those in the preheater where the fuels are brought into a cement plant. A drop-tube furnace and visualization method were developed that show potential for evaluating time- and space-resolved temperature distributions for fuel solid particles and liquid droplets undergoing combustion in various combustion atmospheres. Downdraft gasification has been explored as a means to extract chemical energy from poultry litter while limiting the throughput of potentially deleterious components with regards to use in firing a cement kiln. Results have shown that the clinkering is temperature independent, at least within the controllable temperature range. Limestone also had only a slight effect on the fusion when used to coat the pellets. However, limestone addition did display some promise in regards to chlorine capture, as ash analyses showed chlorine concentrations of more than four times greater in the limestone infused ash as compared to raw poultry litter. A reliable and convenient sampling procedure was developed to estimate the combustion quality of broiler litter that is the best compromise between convenience and reliability by means of statistical analysis. Multi-day trial burns were conducted

  15. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  16. 1. round table - Spent fuels composition. Back-end of the fuel cycle and reprocessing, plutonium and other nuclear materials management. 2. round table - Separation-transmutation. 3. round table - Scenarios for a long term inventory of nuclear materials and wastes; 1. table ronde - La composition des combustibles uses. L'aval du combustible et le retraitement, la gestion du plutonium et des autres matieres nucleaires. 2. table ronde - Separation-transmutation. 3. table ronde - Scenarii pour un inventaire des matieres et des dechets nucleaires a LT

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    The law from December 30, 1991, precisely defines 3 axes of researches for the management of high level and long-lived radioactive wastes: separation/transmutation, surface storage and underground disposal. A global evaluation report about these researches is to be supplied in 2006 by the French government to the Parliament. A first synthesis of the knowledge gained after 14 years of research has led the national commission of the public debate (CNDP) to organize a national debate about the general options of management of high-level and long-lived radioactive wastes before the 2006 date line. The debate comprises 4 public hearings (September 2005: Bar-le-Duc, Saint-Dizier, Pont-du-Gard, Cherbourg), 12 round-tables (October and November 2005: Paris, Joinville, Caen, Nancy, Marseille), a synthesis meeting (December 2005, Dunkerque) and a closing meeting (January 2006, Lyon). This document is the synthesis of the round table debates which took place at Paris on the reprocessing of spent fuels. Three aspects are discussed: the risks linked with the recovery of valorizable materials, the economical viability of the separation/transmutation option, and the future of the nuclear option in the French energy policy. Six presentations (transparencies) are attached with these proceedings which treat of: the reprocessing/recycling to the test, perspectives of future wastes, present day wastes/valorizable materials and future scenarios, critical analysis scenarios, why reprocessing spent fuels?, processing of spent fuels and recycling, separation and transmutation of long-lived radioactive wastes, thorium-uranium cycle. (J.S.)

  17. Efficacy of a Solution-Based Approach for Making Sodalite Waste Forms for an Oxide Reduction Salt Utilized in the Reprocessing of Used Uranium Oxide Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Pierce, David A.; Frank, Steven M.; Matyas, Josef; Burns, Carolyn A.

    2015-04-01

    This paper describes various approaches for making sodalite with a LiCl-Li2O oxide reduction salt used to recover uranium from used oxide fuel. The approaches include sol-gel and solution-based synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na2O-B2O3-SiO2 glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na+ and Cl- to form halite in solution and Li2O and SiO2 to form lithium silicates (e.g., Li2SiO3 or Li2Si2O5) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt.

  18. Efficacy of a solution-based approach for making sodalite waste forms for an oxide reduction salt utilized in the reprocessing of used uranium oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J., E-mail: brian.riley@pnnl.gov [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Pierce, David A. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Frank, Steven M. [Idaho National Laboratory, Idaho Falls, ID 83402 (United States); Matyáš, Josef; Burns, Carolyne A. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States)

    2015-04-15

    This paper describes the various approaches evaluated for making solution-derived sodalite with a LiCl–Li{sub 2}O oxide reduction salt selected to dissolve used uranium oxide fuel so the uranium can be recovered and recycled. The approaches include modified sol–gel and solution-based synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na{sub 2}O–B{sub 2}O{sub 3}–SiO{sub 2} glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na{sup +} and Cl{sup −} to form halite in solution and Li{sub 2}O and SiO{sub 2} to form lithium silicates (e.g., Li{sub 2}SiO{sub 3} or Li{sub 2}Si{sub 2}O{sub 5}) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions in the crystalline product (∼92 mass%) and low porosities using a solution-based approach and this LiCl–Li{sub 2}O salt but that the incorporation of Li into the sodalite is low.

  19. Lab-scale demonstration of the UREX+1a process using spent nuclear fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, C.; Vandegrift, G. F.; Regalbuto, M. C.; Bakel, A.; Bowers, D.; Gelia, A. V.; Hebden, A. S.; Maggos, L. E.; Stepinski, D.; Tsai, Y.; Laidler, J. J.; Chemical Engineering

    2007-01-01

    The Global Nuclear Energy Partnership (GNEP) is developing technologies to greatly expand repository capacity, improve proliferation resistance, and recover valuable energy that would otherwise be discarded; thus assuring a stable energy supply for the future. An important element of this initiative is the separation of key radionuclides followed by either superior waste-disposal forms and/or transmutation of long-lived isotopes. To that end, the GNEP is developing advanced fuel reprocessing systems that separate key radionuclides from spent fuel. One of these systems is the UREX+1a process. The UREX+1a process is a series of four solvent-extraction flowsheets that perform the following operations: (1) recovery of U and Tc (UREX), (2) recovery of Cs and Sr (CCD-PEG), and (3) recovery of TRU and rare earth elements (TRUEX), and (4) separation of TRU elements from the rare earths (TALSPEAK). This paper discusses the results of the demonstration of the UREX, TRUEX, and TALSPEAK processes using spent nuclear fuel, as well as future development needs and plans.

  20. The behaviour of ¹²⁹I released from nuclear fuel reprocessing factories in the North Atlantic Ocean and transport to the Arctic assessed from numerical modelling.

    Science.gov (United States)

    Villa, M; López-Gutiérrez, J M; Suh, Kyung-Suk; Min, Byung-Il; Periáñez, R

    2015-01-15

    A quantitative evaluation of the fate of (129)I, released from the European reprocessing plants of Sellafield (UK) and La Hague (France), has been made by means of a Lagrangian dispersion model. Transport of radionuclides to the Arctic Ocean has been determined. Thus, 5.1 and 16.6 TBq of (129)I have been introduced in the Arctic from Sellafield and La Hague respectively from 1966 to 2012. These figures represent, respectively, 48% and 55% of the cumulative discharge to that time. Inventories in the North Atlantic, including shelf seas, are 4.4 and 13.8 TBq coming from Sellafield and La Hague respectively. These figures are significantly different from previous estimations based on field data. The distribution of these inventories among several shelf seas and regions has been evaluated as well. Mean ages of tracers have been finally obtained, making use of the age-averaging hypothesis. It has been found that mean ages for Sellafield releases are about 3.5 year larger than for La Hague releases.

  1. Transformative monitoring approaches for reprocessing.

    Energy Technology Data Exchange (ETDEWEB)

    Cipiti, Benjamin B.

    2011-09-01

    The future of reprocessing in the United States is strongly driven by plant economics. With increasing safeguards, security, and safety requirements, future plant monitoring systems must be able to demonstrate more efficient operations while improving the current state of the art. The goal of this work was to design and examine the incorporation of advanced plant monitoring technologies into safeguards systems with attention to the burden on the operator. The technologies examined include micro-fluidic sampling for more rapid analytical measurements and spectroscopy-based techniques for on-line process monitoring. The Separations and Safeguards Performance Model was used to design the layout and test the effect of adding these technologies to reprocessing. The results here show that both technologies fill key gaps in existing materials accountability that provide detection of diversion events that may not be detected in a timely manner in existing plants. The plant architecture and results under diversion scenarios are described. As a tangent to this work, both the AMUSE and SEPHIS solvent extraction codes were examined for integration in the model to improve the reality of diversion scenarios. The AMUSE integration was found to be the most successful and provided useful results. The SEPHIS integration is still a work in progress and may provide an alternative option.

  2. Advanced laser processing in fuel cells production

    Energy Technology Data Exchange (ETDEWEB)

    Stollhof, J.; Havrilla, D.; Schaupp, R. [TRUMPF Inc., Plymouth, MI (United States); Loeffler, K. [TRUMPF Laser und Systemtechnik TLD, Ditzingen (Germany)

    2009-07-01

    This paper discussed TRUMPF methods of joining bipolar plates to create fuel cell stacks and manufacture thin foils using diode pumped solid state lasers (DPSSLs). Beam delivery systems and processing optics were discussed. CW disk lasers were used to allow spot diameters smaller than 30 {mu}m and combined with a Galvo technology-based scanning optics systems to enable welding speeds greater than 1 m/s. A TruFiber 300 diffraction limited fiber laser was used for CW laser cutting. Short and ultra-short laser pulses were used to drill thousands of holes per second without a measurable heat-affected zone. The attributes and specifications of the 3 major TRUMPF lasers developed to manufacture fuel cells were also provided.

  3. Spent Fuel Source Term Calculation of Daya Bay Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    XU; Zhi-long; WAN; Hai-xia; LI; Long; WU; Xiao-chun; SHAO; Jing; LIU; Li-li; ZHANG; Jing

    2013-01-01

    The spent fuel of nuclear power plant should be transported to reprocessing plant for reprocessing after reserving for a period of time.Before that,safety analysis and environmental impact assessment should be carried on to the transportation process,which need radioactive source term calculation and analysis.The task of Daya Bay Nuclear Power Plant spent fuel source term calculation includes estimation of

  4. Evaluation and development plan of NRTA measurement methods for the Rokkasho Reprocessing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Li, T.K.; Hakkila, E.A.; Flosterbuer, S.F. [and others

    1995-08-01

    Near-real-time accounting (NRTA) has been proposed as a safeguards method at the Rokkasho Reprocessing Plant (RRP), a large-scale commercial boiling water and pressurized water reactors spent-fuel reprocessing facility. NRTA for RRP requires material balance closures every month. To develop a more effective and practical NRTA system for RRP, we have evaluated NRTA measurement techniques and systems that might be implemented in both the main process and the co-denitration process areas at RRP to analyze the concentrations of plutonium in solutions and mixed oxide powder. Based on the comparative evaluation, including performance, reliability, design criteria, operation methods, maintenance requirements, and estimated costs for each possible measurement method, recommendations for development were formulated. This paper discusses the evaluations and reports on the recommendation of the NRTA development plan for potential implementation at RRP.

  5. VENUS: cold prototype installation of the head-end of the reprocessing of HTR fuel elements. Activity report, 1 July 1976--31 December 1976

    Energy Technology Data Exchange (ETDEWEB)

    Boehnert, R.; Walter, C.

    1977-02-15

    The purpose of the VENUS Project is advance planning for the construction of a cold prototype system to incinerate HTR fuel element graphite. The Venus Project is organized into four phases between advance planning and experimental operation, corresponding to the maturity of the work. It is in the advance planning phase. Status of individual studies is given. (LK)

  6. Application of a room temperature ionic liquid for nuclear spent fuel reprocessing: speciation of trivalent europium and solvatation effects; Application d'un liquide ionique basse temperature pour les procedes de separation: speciation de l'europium trivalent et effets solvatation

    Energy Technology Data Exchange (ETDEWEB)

    Moutiers, G.; Mekki, S. [CEA Saclay, Dept. de Physico-Chimie, Service de Chimie Physique, 91 - Gif sur Yvette (France); Billard, I. [IN2P3/CNRS, 69 - Villeurbanne (France)

    2007-07-01

    One of the solutions proposed for the optimization of the long term storage and conditioning of spent nuclear fuel is to separate actinide and lanthanide both from each other and from other less radioactive metallic species. The industrial proposed processes, based on liquid liquid extraction steps, involve solvents with non negligible vapour pressure and may generate contaminated liquid wastes that will have to be reprocessed. During the last decade, some room-temperature ionic liquids have been studied and integrated into industrial processes. The interest on this class of solvent came out from their 'green' properties (non volatile, non flammable, recyclable, etc...), but also from the variability of their physico-chemical properties (stability, hydrophobicity, viscosity) as a function of the RTIL chemical composition. Indeed, it has been shown that classical chemical industrial processes could be transferred into those media, even more improved, while a certain number of difficulties arising from using traditional solvent can be avoided. In this respect, it could be promising to investigate the ability to use room temperature ionic liquid into the spent nuclear fuel reprocessing field. The aim of this this study is to test the ability of the specific ionic liquid bumimTf{sub 2}N to allow trivalent europium extraction. The choice of this metal is based on the chemical analogy with trivalent minor actinides Curium and Americium which are contributing the greatest part of the long-lived high level radioactive wastes. Handling these elements needs to be very cautious for the safety and radioprotection aspect. Moreover, europium is a very sensitive luminescent probe to its environment even at the microscopic scale. The report is structured with four parts. In a first chapter, we present the main physico-chemical properties of an imidazolium-based ionic liquid family, and then we choose the ionic liquid bumimTf{sub 2}N for the whole thesis and start with

  7. FY09 PROGRESS: MULTI-ISOTOPE PROCESS (MIP) MONITOR

    Energy Technology Data Exchange (ETDEWEB)

    Schwantes, Jon M.; Orton, Christopher R.; Fraga, Carlos G.; Christensen, Richard; Laspe, Amy R.; Ward, Rebecca M.

    2009-10-18

    Model and experimental estimates of the Multi-Isotope Process Monitor performance for determining burnup after dissolution and acid concentration during solvent extraction steps during reprocessing of spent nuclear fuel are presented.

  8. Complex plasmochemical processing of solid fuel

    Directory of Open Access Journals (Sweden)

    Vladimir Messerle

    2012-12-01

    Full Text Available Technology of complex plasmaochemical processing of solid fuel by Ecibastuz bituminous and Turgay brown coals is presented. Thermodynamic and experimental study of the technology was fulfilled. Use of this technology allows producing of synthesis gas from organic mass of coal and valuable components (technical silicon, ferrosilicon, aluminum and silicon carbide and microelements of rare metals: uranium, molybdenum, vanadium etc. from mineral mass of coal. Produced a high-calorific synthesis gas can be used for methanol synthesis, as high-grade reducing gas instead of coke, as well as energy gas in thermal power plants.

  9. MONOLITHIC FUEL FABRICATION PROCESS DEVELOPMENT AT THE IDAHO NATIONAL LABORATORY_

    Energy Technology Data Exchange (ETDEWEB)

    G. A. Moore; F. J. Rice; N. E. Woolstenhulme; J-F. Jue; B. H. Park; S. E. Steffler; N. P. Hallinan; M. D. Chapple; M. C. Marshall; B. L. Mackowiak; C. R. Clark; B. H. Rabin

    2009-11-01

    Full-size/prototypic U10Mo monolithic fuel-foils and aluminum clad fuel plates are being developed at the Idaho National Laboratory’s (INL) Materials and Fuels Complex (MFC). These efforts are focused on realizing Low Enriched Uranium (LEU) high density monolithic fuel plates for use in High Performance Research and Test Reactors. The U10Mo fuel foils under development afford a fuel meat density of ~16 gU/cc and thus have the potential to facilitate LEU conversions without any significant reactor-performance penalty. An overview is provided of the ongoing monolithic UMo fuel development effort, including application of a zirconium barrier layer on fuel foils, fabrication scale-up efforts, and development of complex/graded fuel foils. Fuel plate clad bonding processes to be discussed include: Hot Isostatic Pressing (HIP) and Friction Bonding (FB).

  10. Conceptual designs of NDA instruments for the NRTA system at the Rokkasho Reprocessing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Li, T.K.; Klosterbuer, S.F.; Menlove, H.O. [Los Alamos National Lab., NM (United States). Safeguards Science and Technology Group] [and others

    1996-09-01

    The authors are studying conceptual designs of selected nondestructive assay (NDA) instruments for the near-real-time accounting system at the rokkasho Reprocessing Plant (RRP) of Japan Nuclear Fuel Limited (JNFL). The JNFL RRP is a large-scale commercial reprocessing facility for spent fuel from boiling-water and pressurized-water reactors. The facility comprises two major components: the main process area to separate and produce purified plutonium nitrate and uranyl nitrate from irradiated reactor spent fuels, and the co-denitration process area to combine and convert the plutonium nitrate and uranyl nitrate into mixed oxide (MOX). The selected NDA instruments for conceptual design studies are the MOX-product canister counter, holdup measurement systems for calcination and reduction furnaces and for blenders in the co-denitration process, the isotope dilution gamma-ray spectrometer for the spent fuel dissolver solution, and unattended verification systems. For more effective and practical safeguards and material control and accounting at RRP, the authors are also studying the conceptual design for the UO{sub 3} large-barrel counter. This paper discusses the state-of-the-art NDA conceptual design and research and development activities for the above instruments.

  11. Analysis of nuclear proliferation resistance reprocessing and recycling technologies

    Energy Technology Data Exchange (ETDEWEB)

    Patricia Paviet-Hartmann; Gary Cerefice; Marcela Stacey; Steven Bakhtiar

    2011-05-01

    The PUREX process has been progressively and continuously improved during the past three decades, and these improvements account for successful commercialization of reprocessing in a few countries. The renewed interest in nuclear energy and the international growth of nuclear electricity generation do not equate – and should not be equated -with increasing proliferation risks. Indeed, the nuclear renaissance presents a unique opportunity to enhance the culture of non-proliferation. With the recent revival of interest in nuclear technology, technical methods for prevention of nuclear proliferation are being revisited. Robust strategies to develop new advanced separation technologies are emerging worldwide for sustainability and advancement of nuclear energy with enhanced proliferation resistance. On the other hand, at this moment, there are no proliferation resistance advanced technologies. . Until now proliferation resistance as it applies to reprocessing has been focused on not separating a pure stream of weapons-usable plutonium. France, as an example, has proposed a variant of the PUREX process, the COEX TM process, which does not result on a pure plutonium product stream. A further step is to implement a process based on group extraction of actinides and fission products associated with a homogeneous recycling strategy (UNEX process in the US, GANEX process in France). Such scheme will most likely not be deployable on an industrial scale before 2030 or so because it requires intensive R&D and robust flowsheets. Finally, future generation recycling schemes will handle the used nuclear fuel in fast neutron reactors. This means that the plutonium throughput of the recycling process may increase. The need is obvious for advanced aqueous recycling technologies that are intrinsically more proliferation resistant than the commercial PUREX process. In this paper, we review the actual PUREX process along with the advanced recycling technologies that will enhance

  12. Geant4 Model Validation of Compton Suppressed System for Process monitoring of Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bender, Sarah; Unlu, Kenan; Orton, Christopher R.; Schwantes, Jon M.

    2013-05-01

    Nuclear material accountancy is of continuous concern for the regulatory, safeguards, and verification communities. In particular, spent nuclear fuel reprocessing facilities pose one of the most difficult accountancy challenges: monitoring highly radioactive, fluid sample streams in near real-time. The Multi-Isotope Process monitor will allow for near-real-time indication of process alterations using passive gamma-ray detection coupled with multivariate analysis techniques to guard against potential material diversion or to enhance domestic process monitoring. The Compton continuum from the dominant 661.7 keV 137Cs fission product peak obscures lower energy lines which could be used for spectral and multivariate analysis. Compton suppression may be able to mitigate the challenges posed by the high continuum caused by scattering. A Monte Carlo simulation using the Geant4 toolkit is being developed to predict the expected suppressed spectrum from spent fuel samples to estimate the reduction in the Compton continuum. Despite the lack of timing information between decay events in the particle management of Geant4, encouraging results were recorded utilizing only the information within individual decays without accounting for accidental coincidences. The model has been validated with single and cascade decay emitters in two steps: as an unsuppressed system and with suppression activated. Results of the Geant4 model validation will be presented.

  13. Practice comparisons between accelerated resolution therapy, eye movement desensitization and reprocessing and cognitive processing therapy with case examples.

    Science.gov (United States)

    Hernandez, Diego F; Waits, Wendi; Calvio, Lisseth; Byrne, Mary

    2016-12-01

    Recent outcomes for Cognitive Processing Therapy (CPT) and Prolonged Exposure (PE) therapy indicate that as many as 60-72% of patients retain their PTSD diagnosis after treatment with CPT or PE. One emerging therapy with the potential to augment existing trauma focused therapies is Accelerated Resolution Therapy (ART). ART is currently being used along with evidence based approaches at Fort Belvoir Community Hospital and by report has been both positive for clients as well as less taxing on professionals trained in ART. The following is an in-practice theoretical comparison of CPT, EMDR and ART with case examples from Fort Belvoir Community Hospital. While all three approaches share common elements and interventions, ART distinguishes itself through emphasis on the rescripting of traumatic events and the brevity of the intervention. While these case reports are not part of a formal study, they suggest that ART has the potential to augment and enhance the current delivery methods of mental health care in military environments.

  14. Thermal Lens Spectroscopy as a 'new' analytical tool for actinide determination in nuclear reprocessing processes

    Energy Technology Data Exchange (ETDEWEB)

    Canto, Fabrice; Couston, Laurent; Magnaldo, Alastair [CEA-Valrho DEN/DRCP/SCPS/LCAM BP17171 30207 Bagnols/Ceze cedex (France); Broquin, Jean-Emmanuel [IMEP/ENSERG 23 rue des Martyrs BP257 38016 Grenoble (France); Signoret, Philippe [UM2/IES UMR 5214. Place Eugene Bataillon 34095 Montpellier cedex5 (France)

    2008-07-01

    Thermal Lens Spectroscopy (TLS) consists of measuring the effects induced by the relaxation of molecules excited by photons. Twenty years ago, the Cea already worked on TLS. Technologic reasons impeded. But, needs in sensitive analytical methods coupled with very low sample volumes (for example, traces of Np in the COEX{sup TM} process) and also the reduction of the nuclear wastes encourage us to revisit this method thanks to the improvement of optoelectronic technologies. We can also imagine coupling TLS with micro-fluidic technologies, decreasing significantly the experiments cost. Generally two laser beams are used for TLS: one for the selective excitation by molecular absorption (inducing the thermal lens) and one for probing the thermal lens. They can be coupled with different geometries, collinear or perpendicular, depending on the application and on the laser mode. Also, many possibilities of measurement have been studied to detect the thermal lens signal: interferometry, direct intensities variations, deflection etc... In this paper, one geometrical configuration and two measurements have been theoretically evaluated. For a single photodiode detection (z-scan) the limit of detection is calculated to be near 5*10{sup -6} mol*L{sup -1} for Np(IV) in dodecane. (authors)

  15. Cost reductions of fuel cells for transport applications: fuel processing options

    Science.gov (United States)

    Teagan, W. P.; Bentley, J.; Barnett, B.

    The highly favorable efficiency/environmental characteristics of fuel cell technologies have now been verified by virtue of recent and ongoing field experience. The key issue regarding the timing and extent of fuel cell commercialization is the ability to reduce costs to acceptable levels in both stationary and transport applications. It is increasingly recognized that the fuel processing subsystem can have a major impact on overall system costs, particularly as ongoing R&D efforts result in reduction of the basic cost structure of stacks which currently dominate system costs. The fuel processing subsystem for polymer electrolyte membrane fuel cell (PEMFC) technology, which is the focus of transport applications, includes the reformer, shift reactors, and means for CO reduction. In addition to low cost, transport applications require a fuel processor that is compact and can start rapidly. This paper describes the impact of factors such as fuel choice, operating temperature, material selection, catalyst requirements, and controls on the cost of fuel processing systems. There are fuel processor technology paths which manufacturing cost analyses indicate are consistent with fuel processor subsystem costs of under 150/kW in stationary applications and 30/kW in transport applications. As such, the costs of mature fuel processing subsystem technologies should be consistent with their use in commercially viable fuel cell systems in both application categories.

  16. Economics and resources analysis of the potential use of reprocessing options by the current Spanish nuclear reactor park

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez-Velarde, F.; Merino Rodriguez, I.; Gonzalez-Romero, E.

    2014-07-01

    Reprocessing of irradiated nuclear fuel serves multiple purposes, from Pu separation and recovery for MOX fuel fabrication to reduction of high level waste volume, and is nowadays being implemented in several countries like France, Japan, Russia or United Kingdom. This work is aimed at exploring the possibility (in resources and economic terms) of implementing reprocessing for MOX fabrication in Spain. (Author)

  17. Scientific research on the back-end of the fuel cycle for the 21. century; Les recherches scientifiques sur l'aval du cycle pour le 21. siecle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    The aim of the Atalante-2000 conference is to present the major research axis concerning the nuclear fuel cycle back-end. The different topics are: - Present options concerning fuel cycle back-end; - Reprocessing of spent fuel; - Advanced separation for transmutation; - Processing and packaging of radioactive wastes; - Design and fabrication of targets for transmutation; and - Conversion of military plutonium into MOX fuels.

  18. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-26

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  19. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, task 17208: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-26

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  20. Selection of Fuel by Using Analytical Hierarchy Process

    Directory of Open Access Journals (Sweden)

    Asilata M. Damle,

    2015-04-01

    Full Text Available Selection of fuel is a very important and critical decision one has to make. Various criteria are to be considered while selecting a fuel. Some of important criteria are Fuel Economy, Availability of fuel, Pollution from vehicle, Maintenance of the vehicle. Selection of best fuel is a complex situation. It needs a multi-criteria analysis. Earlier, the solution to the problem were found by applying classical numerical methods which took into account only technical and economic merits of the various alternatives. By applying multi-criteria tools, it is possible to obtain more realistic results. This paper gives a systematic analysis for selection of fuel by using Analytical Hierarchy Process (AHP. This is a multi-criteria decision making process. By using AHP we can select the fuel by comparing various factors in a mathematical model. This is a scientific method to find out the best fuel by making pairwise comparisons.

  1. Motion State of Fuel within Shell in Projection Acceleration Process

    Directory of Open Access Journals (Sweden)

    Qi Zhang

    2003-07-01

    Full Text Available The fuel-air explosive (FAE warheads are charged with the liquid-solid mixture fuel. The fuel is different 'om conventional solid explosives in physical and mechanical properties. The mass centre of the charged fuel changes during projecting the projectile. In this study, a method to calculate the mass centre change of the charged fuel is suggested and the  influence of this change on the projectile motion state in the projection process is discussed. The results show that in projection, the fuel mass centre varies with the projection acceleration and the deformation characteristics of the mixture fuel. The higher is the acceleration, the larger is the displacement of the mass centre. This displacement also increases with the compressibility of the fuel. It constitutes an influence on the state of motion for the whole projectile in the projection process, whose calculation approach is also proposed. The result provides a theoretical basis for the design of the FAE weapons.

  2. Development of Advanced Spent Fuel Management Process

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chung Seok; Choi, I. K.; Kwon, S. G. (and others)

    2007-06-15

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm{sup 2} and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields.

  3. Monitoring 85Kr at China Reprocessing and Radiochemistry Laboratory

    Institute of Scientific and Technical Information of China (English)

    LV; Xue-sheng; LIU; Guo-rong; WANG; Chen; XU; Chang-kun; ZHOU; Hao

    2015-01-01

    85Kr is a long life fission product and a noble gas and one of the main target radionuclides in monitoring environment for the reprocessing plant.During the course of the spent fuel elements are dissolved,85Kr is released and part of these 85Kr will be let out through the stack after

  4. MEDIUM PRESSURE HYDROUPGRADING PROCESS (MHUG) AND PRODUCTION OF CLEAN FUELS

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The medium pressure hydroupgrading process (MHUG) unit with an 800 kt/a processing capacity of Jinzhou Petrochemical Company is used to hydroupgrade the mixture of FCC LCO fuel and straight-run diesel fuel in the presence of RN/RT series catalysts for improvement of the quality of the diesel fuel. Meanwhile, catalytic reforming feedstock is also obtained. The sulfur, nitrogen and aromatics contained in the hydroupgraded diesel fuel products can be minimized and the cetane number can be heightened. The produced clean fuels can meet the requirements of environmental protection.

  5. Fabrication characteristics of dry process fuel with a variation of fuel burn-ups

    Energy Technology Data Exchange (ETDEWEB)

    Park, Geun Il; Kim, W. K.; Lee, J. W. [and others

    2004-11-01

    Fabrication characteristics of the dry processed fuel with a variation of fuel burn-ups in a range of 27,300 to 65,000 MWD/tU were experimentally evaluated. Density comparison of powders which were fabricated from oxidation, OREOX and milling processes at same process conditions was performed with a function of fuel burn-ups respectively. The influence of fuel burn-ups on sintering characteristics of dry processed fuel was studied by comparing the density change of sintered pellet as well as green pellet. Weight gain by fuel oxidation to U{sub 3}O{sub 8} showed semi-linear dependence with increasing fuel burn-ups. OREOX powder density increased up to 3.7 g/cm{sup 3} at high burn-up fuel, and the density of milled powder with fuel burn-ups represented almost similar value of 3.2{+-}0.2 g/cm{sup 3}. Also, the green pellet density compacted by 120 MPa decreased smoothly with increasing fuel burn-ups, and the density change of sintered pellet showed the similar trend as green pellet. The sintered density of pellet in a range of 27,000 to 40,000 MWD/tU was found to be more 95% of Theoretical Density(T.D.), but the sintered pellet density fabricated from high burn-up fuel showed a range of 92 % to 93% of T.D.

  6. HTGR fuel recycle development program. Quarterly progress report for the period ending August 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    The work reported includes the development of unit processes and equipment for reprocessing of High-Temperature Gas-Cooled Reactor (HTGR) fuel, the design and development of an integrated pilot line to demonstrate the head end of HTGR reprocessing using unirradiated fuel materials, and design work in support of Hot Engineering Tests (HET). Work is also described on tradeoff studies concerning the required design of facilities and equipment for the large-scale recycle of HTGR fuels in order to guide the development activities for HTGR fuel recycle.

  7. Fuel Cell Stations Automate Processes, Catalyst Testing

    Science.gov (United States)

    2010-01-01

    Glenn Research Center looks for ways to improve fuel cells, which are an important source of power for space missions, as well as the equipment used to test fuel cells. With Small Business Innovation Research (SBIR) awards from Glenn, Lynntech Inc., of College Station, Texas, addressed a major limitation of fuel cell testing equipment. Five years later, the company obtained a patent and provided the equipment to the commercial world. Now offered through TesSol Inc., of Battle Ground, Washington, the technology is used for fuel cell work, catalyst testing, sensor testing, gas blending, and other applications. It can be found at universities, national laboratories, and businesses around the world.

  8. Aluminum-alloy processing of Th- and U-based fuels. Quarterly technical progress report, April-June 1978

    Energy Technology Data Exchange (ETDEWEB)

    Lukens, H.R.; Bryan, D.E.; MacKenzie, J.K.; Preskitt, C.A.; Rock, D.H.; Selph, W.E.; Snellen, G.; Snowden, D.P.; Teitel, R.J.

    1978-06-15

    The objective of the work is to develop economically feasible, proliferation-proof flowsheets for the reprocessing of Th- and U-based fuels with aluminum as a key solvent element. Included within this objective are such considerations as compatible head-end and finishing steps, materials recycling, fission-product disposal, and hot-processing. An important goal of the program is that there not be a complete separation of fissile isotopes from nonfissile heavy metals at any point in the process. Demonstration of a suitable process is the final objective of the current project. During the present quarter several new alternative flowsheets have been formulated that reflect both the contributions of data obtained in an ongoing literature survey and concepts that seem worth investigating with respect to project goals. Also, experimental work is now under way toward the gathering of essential data that is not in the literature. 7 figures.

  9. Materials and processes for solar fuel production

    CERN Document Server

    Viswanathan, Balasubramanian; Lee, Jae Sung

    2014-01-01

    This book features different approaches to non-biochemical pathways for solar fuel production. This one-of-a-kind book addresses photovoltaics, photocatalytic water splitting for clean hydrogen production and CO2 conversion to hydrocarbon fuel through in-depth comprehensive contributions from a select blend of established and experienced authors from across the world. The commercial application of solar based systems, with particular emphasis on non-PV based devices have been discussed. This book intends to serve as a primary resource for a multidisciplinary audience including chemists, engineers and scientists providing a one-stop location for all aspects related to solar fuel production. The material is divided into three sections: Solar assisted water splitting to produce hydrogen; Solar assisted CO2 utilization to produce green fuels and Solar assisted electricity generation. The content strikes a balance between theory, material synthesis and application with the central theme being solar fuels.

  10. Guide to the selection, training, and licensing or certification of reprocessing plant operators. Volume I

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-06-01

    The Code of Federal Regulations, Title 10, Part 55, establishes procedures and criteria for the licensing of operators, including senior operators, in ''Production and Utilization Facilities'', which includes plants for reprocessing irradiated fuel. A training guide is presented which will facilitate the licensing of operators for nuclear reprocessing plants by offering generalized descriptions of the basic principles (theory) and the unit operations (mechanics) employed in reprocessing spent fuels. In the present volume, details about the portions of a training program that are of major interest to management are presented. (JSR)

  11. Modeling the Thermal Rocket Fuel Preparation Processes in the Launch Complex Fueling System

    Directory of Open Access Journals (Sweden)

    A. V. Zolin

    2015-01-01

    Full Text Available It is necessary to carry out fuel temperature preparation for space launch vehicles using hydrocarbon propellant components. A required temperature is reached with cooling or heating hydrocarbon fuel in ground facilities fuel storages. Fuel temperature preparing processes are among the most energy-intensive and lengthy processes that require the optimal technologies and regimes of cooling (heating fuel, which can be defined using the simulation of heat exchange processes for preparing the rocket fuel.The issues of research of different technologies and simulation of cooling processes of rocket fuel with liquid nitrogen are given in [1-10]. Diagrams of temperature preparation of hydrocarbon fuel, mathematical models and characteristics of cooling fuel with its direct contact with liquid nitrogen dispersed are considered, using the numerical solution of a system of heat transfer equations, in publications [3,9].Analytical models, allowing to determine the necessary flow rate and the mass of liquid nitrogen and the cooling (heating time fuel in specific conditions and requirements, are preferred for determining design and operational characteristics of the hydrocarbon fuel cooling system.A mathematical model of the temperature preparation processes is developed. Considered characteristics of these processes are based on the analytical solutions of the equations of heat transfer and allow to define operating parameters of temperature preparation of hydrocarbon fuel in the design and operation of the filling system of launch vehicles.The paper considers a technological system to fill the launch vehicles providing the temperature preparation of hydrocarbon gases at the launch site. In this system cooling the fuel in the storage tank before filling the launch vehicle is provided by hydrocarbon fuel bubbling with liquid nitrogen. Hydrocarbon fuel is heated with a pumping station, which provides fuel circulation through the heat exchanger-heater, with

  12. Efficient method for variable reprocessing of paraffinic and highly paraffinic feedstocks

    Energy Technology Data Exchange (ETDEWEB)

    Odintsov, O.K.; Pereverzev, A.N.; Brovarskaya, S.S.

    1981-01-01

    Different procedures for reprocessing are necessary to obtain the maximum yields of desirable products and minimize the refining costs for feedstocks having significant concentrations of high molecular weight paraffins. A detailed processing scheme is given for each of two reprocessed oils: a Belorussian type containing 25-26 weight % n-paraffins in the kerosene-diesel fractions and a Romashkin crude from the ''Druzhba'' pipeline containing 13-15 weight % n-paraffins in the same fraction. The deparaffining process for the feedstocks must be stable and variable in response to the concentration of paraffins in the crude; an acceptable reactive fuel, complying with thermal stability requirements, was obtained in an 8% by volume yield and had an initial crystallization temperature of -55/sup 0/C and an aromatic content of 19 weight %; an acceptable fuel oil, complying with appropriate technical standards, was obtained by mixing the diesel fraction and fuel oil (in a 2:3 ratio) after the atmospheric distillation of the oil mixture from the ''Druzhba'' pipeline.

  13. Renewable hydrogen production for fossil fuel processing

    Energy Technology Data Exchange (ETDEWEB)

    Greenbaum, E.; Lee, J.W.; Tevault, C.V. [and others

    1995-06-01

    In the fundamental biological process of photosynthesis, atmospheric carbon dioxide is reduced to carbohydrate using water as the source of electrons with simultaneous evolution of molecular oxygen: H{sub 2}O + CO{sub 2} + light {yields} O{sub 2} + (CH{sub 2}O). It is well established that two light reactions, Photosystems I and II (PSI and PSII) working in series, are required to perform oxygenic photosynthesis. Experimental data supporting the two-light reaction model are based on the quantum requirement for complete photosynthesis, spectroscopy, and direct biochemical analysis. Some algae also have the capability to evolve molecular hydrogen in a reaction energized by the light reactions of photosynthesis. This process, now known as biophotolysis, can use water as the electron donor and lead to simultaneous evolution of molecular hydrogen and oxygen. In green algae, hydrogen evolution requires prior incubation under anaerobic conditions. Atmospheric oxygen inhibits hydrogen evolution and also represses the synthesis of hydrogenase enzyme. CO{sub 2} fixation competes with proton reduction for electrons relased from the photosystems. Interest in biophotolysis arises from both the questions that it raises concerning photosynthesis and its potential practical application as a process for converting solar energy to a non-carbon-based fuel. Prior data supported the requirement for both Photosystem I and Photosystem II in spanning the energy gap necessary for biophotolysis of water to oxygen and hydrogen. In this paper we report the at PSII alone is capable of driving sustained simultaneous photoevolution of molecular hydrogen and oxygen in an anaerobically adapted PSI-deficient strain of Chlamydomonas reinhardtii, mutant B4, and that CO{sub 2} competes as an electron acceptor.

  14. Criticality Safety Experimental Investigation of Heterogeneous Fuel

    Institute of Scientific and Technical Information of China (English)

    WANG; Fan; ZHOU; Qi; XIA; Zhao-dong; ZHU; Qing-fu

    2015-01-01

    The spent fuel dissolver is the most important component in the reprocessing plant of the spent fuel dissolver reprocessing steps.The tonnage throughput,criticality safety and economical efficiency of the reprocess or mostly depend on the tonnage throughput,treatment rate and criticality safety of the dissolver.Because of the

  15. DUPIC nuclear fuel manufacturing and process technology development

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Park, J. J.; Lee, J. W. [and others

    2000-05-01

    In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated.

  16. DUPIC nuclear fuel manufacturing and process technology development

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Yang, M. S.; Kim, S. S. [and others

    2002-05-01

    In this study, the DUPIC fuel fabrication technology for DUPIC pellet and element satisfying the standard specification was verified through (1) the improvement of fabrication technology and equipment, (2) remote operation of fuel manufacturing and inspection equipment installed at DFDF and (3) the study on the material properties of DUPIC fuel. The blending process was newly developed for making DUPIC powder composition homogeneous, and mixing process was added to the DUPIC process flow for fabricating crack-free pellets. A series of fabrication experiments were carried out in terms of various process conditions. Based on these experimental results, the optimal process flow and conditions for DUPIC fuel fabrication were established. 6 DUPIC elements and 6 mini-elements for irradiation test in HANARO was successfully fabricated using 7.4 kg of spent PWR fuel in 2000. The process qualification tests has been performed using 10 kg of spent PWR fuel since May 2001. The optimal DUPIC fuel fabrication process meeting AECL's quality requirements has been established and qualified. Quality assurance system for DUPIC fuel fabrication was also established in cooperation of AECL.

  17. Fuel ethanol production: process design trends and integration opportunities.

    Science.gov (United States)

    Cardona, Carlos A; Sánchez, Oscar J

    2007-09-01

    Current fuel ethanol research and development deals with process engineering trends for improving biotechnological production of ethanol. In this work, the key role that process design plays during the development of cost-effective technologies is recognized through the analysis of major trends in process synthesis, modeling, simulation and optimization related to ethanol production. Main directions in techno-economical evaluation of fuel ethanol processes are described as well as some prospecting configurations. The most promising alternatives for compensating ethanol production costs by the generation of valuable co-products are analyzed. Opportunities for integration of fuel ethanol production processes and their implications are underlined. Main ways of process intensification through reaction-reaction, reaction-separation and separation-separation processes are analyzed in the case of bioethanol production. Some examples of energy integration during ethanol production are also highlighted. Finally, some concluding considerations on current and future research tendencies in fuel ethanol production regarding process design and integration are presented.

  18. Reprocessing and reuse of urological armamentarium: How correct are we!

    Directory of Open Access Journals (Sweden)

    Krutik Vipulbhai Raval

    2017-01-01

    Full Text Available Healthcare is expensive for a large proportion of the population in spite of high per capita income and good health insurance penetration. In an effort to reduce cost of the procedure, reprocessing of devices was started in the late 1970s. Reprocessing practice includes various measures such as proper cleaning, disinfection, and sterilization procedures. As reprocessing is aimed at reducing cost, there is a potential risk of compromising patient safety due to cross contamination after inadequate sterilization. There is also risk of performance alteration of urological reprocessed devices during sterilization/disinfection processing. Therefore, there is a need for formulating proper guidelines to decide methods of reprocessing for various urological equipment. There is also need to discuss the problematic areas that urologists face and to find their solutions. A PubMed search was made in September 2016, using key words “reprocessing of medical devices,” “Single Use Devices,” “methods of reprocessing of devices in clinical practice,” “use of formalin chamber,” “urological disposable sterilization,” etc., After excluding duplicates, all English articles were reviewed by title and abstract. Full texts of selected articles were obtained, and these articles were cross-referenced to find any other related articles. All the articles were reviewed. A product can be reused if it can be economically reprocessed with validated protocols with preservation of its function. There is no reason to discard it after one use. This practice is useful for controlling economics of a urological case and to reduce the financial burden. Current Food and Drug Administration guidelines are stringent. The contamination described to test the sterilization process in the suggested guidelines actually does never exist in clinical practice. Therefore, new guidelines considering the clinical practice scenario are desirable.

  19. Suomi NPP VIIRS Reflective Solar Bands Operational Calibration Reprocessing

    Directory of Open Access Journals (Sweden)

    Slawomir Blonski

    2015-12-01

    Full Text Available Radiometric calibration coefficients for the VIIRS (Visible Infrared Imaging Radiometer Suite reflective solar bands have been reprocessed from the beginning of the Suomi NPP (National Polar-orbiting Partnership mission until present. An automated calibration procedure, implemented in the NOAA (National Oceanic and Atmospheric Administration JPSS (Joint Polar Satellite System operational data production system, was applied to reprocess onboard solar calibration data and solar diffuser degradation measurements. The latest processing parameters from the operational system were used to include corrected solar vectors, optimized directional dependence of attenuation screens transmittance and solar diffuser reflectance, updated prelaunch calibration coefficients without an offset term, and optimized Robust Holt-Winters filter parameters. The parameters were consistently used to generate a complete set of the radiometric calibration coefficients for the entire duration of the Suomi NPP mission. The reprocessing has demonstrated that the automated calibration procedure can be successfully applied to all solar measurements acquired from the beginning of the mission until the full deployment of the automated procedure in the operational processing system. The reprocessed calibration coefficients can be further used to reprocess VIIRS SDR (Sensor Data Record and other data products. The reprocessing has also demonstrated how the automated calibration procedure can be used during activation of the VIIRS instruments on the future JPSS satellites.

  20. Reprocessing technology development for irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, H.; Sakamoto, N. [Oarai Research Establishment, Ibaraki-ken (Japan); Tatenuma, K. [KAKEN Co., Ibaraki-ken (Japan)] [and others

    1995-09-01

    At present, beryllium is under consideration as a main candidate material for neutron multiplier and plasma facing material in a fusion reactor. Therefore, it is necessary to develop the beryllium reprocessing technology for effective resource use. And, we have proposed reprocessing technology development on irradiated beryllium used in a fusion reactor. The preliminary reprocessing tests were performed using un-irradiated and irradiated beryllium. At first, we performed beryllium separation tests using un-irradiated beryllium specimens. Un-irradiated beryllium with beryllium oxide which is a main impurity and some other impurities were heat-treated under chlorine gas flow diluted with Ar gas. As the results high purity beryllium chloride was obtained in high yield. And it appeared that beryllium oxide and some other impurities were removed as the unreactive matter, and the other chloride impurities were separated by the difference of sublimation temperature on beryllium chloride. Next, we performed some kinds of beryllium purification tests from beryllium chloride. And, metallic beryllium could be recovered from beryllium chloride by the reduction with dry process. In addition, as the results of separation and purification tests using irradiated beryllium specimens, it appeared that separation efficiency of Co-60 from beryllium was above 96%. It is considered that about 4% Co-60 was carried from irradiated beryllium specimen in the form of cobalt chloride. And removal efficiency of tritium from irradiated beryllium was above 95%.

  1. DUPIC fuel fabrication in shielded facilities in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.W.; Kim, W.K.; Kim, S.S.; Yang, M.S.; Park, H.S. [Korea Atomic Energy Research Institute, Yusong-ku, Taejon City (Korea, Republic of)

    2001-07-01

    The DUPIC(Direct use of spent PWR fuel in CANDU reactors) fuel cycle is to directly refabricate the CANDU fuel from spent PWR fuel materials by thermal and mechanical processes without wet reprocessing process. The concept was proposed and termed DUPIC in joint research program between the Korea Atomic Energy Research Institute (KAERI), Atomic Energy of Canada Limited (AECL) and the US Departments of State in 1992. The DUPIC fuel cycle has many advantages over direct disposal or wet reprocessing and MOX fuel cycle in terms of proliferation resistance, reduction of spent fuel accumulation and uranium resource utilization, etc. Since the material in the DUPIC fuel fabrication process is highly radioactive due to no separation of uranium, plutonium and fission products, which is an intrinsic characteristic of the DUPIC process, all fabrication and characterization processes should be performed remotely in highly shielded hot cell facilities. KAERI has developed the remote fuel fabrication equipment and has successfully completed the installation of them in the shielded facilities, called DFDF (DUPIC Fuel Development Facility), at KAERI in early 2000. Based on the fuel fabrication technologies, including powder treatment, pelletizing and laser welding, KAERI has successfully fabricated DUPIC fuel pellets and elements with various design specifications to evaluate the performance of DUPIC fuel through irradiation tests at the HANARO research reactor. This paper describes KAERI's progress in DUPIC fuel fabrication. (author)

  2. Effects of Fuel Quantity on Soot Formation Process for Biomass-Based Renewable Diesel Fuel Combustion

    KAUST Repository

    Jing, Wei

    2016-12-01

    Soot formation process was investigated for biomass-based renewable diesel fuel, such as biomass to liquid (BTL), and conventional diesel combustion under varied fuel quantities injected into a constant volume combustion chamber. Soot measurement was implemented by two-color pyrometry under quiescent type diesel engine conditions (1000 K and 21% O2 concentration). Different fuel quantities, which correspond to different injection widths from 0.5 ms to 2 ms under constant injection pressure (1000 bar), were used to simulate different loads in engines. For a given fuel, soot temperature and KL factor show a different trend at initial stage for different fuel quantities, where a higher soot temperature can be found in a small fuel quantity case but a higher KL factor is observed in a large fuel quantity case generally. Another difference occurs at the end of combustion due to the termination of fuel injection. Additionally, BTL flame has a lower soot temperature, especially under a larger fuel quantity (2 ms injection width). Meanwhile, average soot level is lower for BTL flame, especially under a lower fuel quantity (0.5 ms injection width). BTL shows an overall low sooting behavior with low soot temperature compared to diesel, however, trade-off between soot level and soot temperature needs to be carefully selected when different loads are used.

  3. Systems and processes for conversion of ethylene feedstocks to hydrocarbon fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lilga, Michael A.; Hallen, Richard T.; Albrecht, Karl O.; Cooper, Alan R.; Frye, John G.; Ramasamy, Karthikeyan Kallupalayam

    2017-05-30

    Systems, processes, and catalysts are disclosed for obtaining fuel and fuel blends containing selected ratios of open-chain and closed-chain fuel-range hydrocarbons suitable for production of alternate fuels including gasolines, jet fuels, and diesel fuels. Fuel-range hydrocarbons may be derived from ethylene-containing feedstocks and ethanol-containing feedstocks.

  4. Modeling of large-scale oxy-fuel combustion processes

    DEFF Research Database (Denmark)

    Yin, Chungen

    2012-01-01

    , among which radiative heat transfer under oxy-fuel conditions is one of the fundamental issues. This paper demonstrates the nongray-gas effects in modeling of large-scale oxy-fuel combustion processes. Oxy-fuel combustion of natural gas in a 609MW utility boiler is numerically studied, in which....... The simulation results show that the gray and non-gray calculations of the same oxy-fuel WSGGM make distinctly different predictions in the wall radiative heat transfer, incident radiative flux, radiative source, gas temperature and species profiles. In relative to the non-gray implementation, the gray...

  5. Calibration of burnup monitor installed in Rokkasho Reprocessing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Naito, Hirofumi; Hirota, Masanari [Japan Nuclear Fuel Co. Ltd., Rokkasho, Aomori (Japan); Natsume, Koichiro [Isogo Engineering Center, Toshiba Corporation, Yokohama, Kanagawa (Japan); Kumanomido, Hironori [Nuclear Engineering Laboratory, Toshiba Corporation, Kawasaki, Kanagawa (Japan)

    2000-06-01

    Rokkasho Reprocessing Plant uses burnup credit for criticality control at the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. A burnup monitor measures nondestructively burnup value of a spent fuel assembly and guarantees the credit for burnup. For practical reasons, a standard radiation source is not used in calibration of the burnup monitor, but the burnup values of many spent fuel assemblies are measured based on operator-declared burnup values. This paper describes the concept of burnup credit, the burnup monitor, and the calibration method. It is concluded, from the results of calibration tests, that the calibration method is valid. (author)

  6. Integrated coke, asphalt and jet fuel production process and apparatus

    Science.gov (United States)

    Shang, Jer Y.

    1991-01-01

    A process and apparatus for the production of coke, asphalt and jet fuel m a feed of fossil fuels containing volatile carbon compounds therein is disclosed. The process includes the steps of pyrolyzing the feed in an entrained bed pyrolyzing means, separating the volatile pyrolysis products from the solid pyrolysis products removing at least one coke from the solid pyrolysis products, fractionating the volatile pyrolysis products to produce an overhead stream and a bottom stream which is useful as asphalt for road pavement, condensing the overhead stream to produce a condensed liquid fraction and a noncondensable, gaseous fraction, and removing water from the condensed liquid fraction to produce a jet fuel-containing product. The disclosed apparatus is useful for practicing the foregoing process. the process provides a useful method of mass producing and jet fuels from materials such as coal, oil shale and tar sands.

  7. Electrochemical fluorination for processing of used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2016-07-05

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  8. Galvanic cell for processing of used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2017-02-07

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  9. Rethinking nuclear fuel recycling.

    Science.gov (United States)

    von Hippel, Frank N

    2008-05-01

    Spent nuclear fuel contains plutonium which can be extracted and used in new fuel. To reduce the amount of long-lived radioactive waste, the U.S. Department of Energy has proposed reprocessing spent fuel in this way and then "burning" the plutonium in special reactors. But reprocesssing is very expensive. Also, spent fuel emits lethal radiation, whereas separated plutonium can be handled easily. So reprocessing invites the possibility that terrorists might steal plutonium and construct an atom bormb. The authors argue against reprocessing and for storing the waste in casks until an underground repository is ready.

  10. Distillate fuel-oil processing for phosphoric acid fuel-cell power plants

    Energy Technology Data Exchange (ETDEWEB)

    Ushiba, K. K.

    1980-02-01

    The current efforts to develop distillate oil-steam reforming processes are reviewed, and the applicability of these processes for integration with the fuel cell are discussed. The development efforts can be grouped into the following processing approaches: high-temperature steam reforming (HTSR); autothermal reforming (ATR); autothermal gasification (AG); and ultra desulfurization followed by steam reforming. Sulfur in the feed is a key problem in the process development. A majority of the developers consider sulfur as an unavoidable contaminant of distillate fuel and are aiming to cope with it by making the process sulfur-tolerant. In the HTSR development, the calcium aluminate catalyst developed by Toyo Engineering represents the state of the art. United Technology (UTC), Engelhard, and Jet Propulsion Laboratory (JPL) are also involved in the HTSR research. The ATR of distillate fuel is investigated by UTC and JPL. The autothermal gasification (AG) of distillate fuel is being investigated by Engelhard and Siemens AG. As in the ATR, the fuel is catalytically gasified utilizing the heat generated by in situ partial combustion of feed, however, the goal of the AG is to accomplish the initial breakdown of the feed into light gases and not to achieve complete conversion to CO and H/sub 2/. For the fuel-cell integration, a secondary reforming of the light gases from the AG step is required. Engelhard is currently testing a system in which the effluent from the AG section enters the steam-reforming section, all housed in a single vessel. (WHK)

  11. Separation of Plutonium from Irradiated Fuels and Targets

    Energy Technology Data Exchange (ETDEWEB)

    Gray, Leonard W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Holliday, Kiel S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Murray, Alice [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Thompson, Major [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Thorp, Donald T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Yarbro, Stephen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Venetz, Theodore J. [Hanford Site, Benton County, WA (United States)

    2015-09-30

    Spent nuclear fuel from power production reactors contains moderate amounts of transuranium (TRU) actinides and fission products in addition to the still slightly enriched uranium. Originally, nuclear technology was developed to chemically separate and recover fissionable plutonium from irradiated nuclear fuel for military purposes. Military plutonium separations had essentially ceased by the mid-1990s. Reprocessing, however, can serve multiple purposes, and the relative importance has changed over time. In the 1960’s the vision of the introduction of plutonium-fueled fast-neutron breeder reactors drove the civilian separation of plutonium. More recently, reprocessing has been regarded as a means to facilitate the disposal of high-level nuclear waste, and thus requires development of radically different technical approaches. In the last decade or so, the principal reason for reprocessing has shifted to spent power reactor fuel being reprocessed (1) so that unused uranium and plutonium being recycled reduce the volume, gaining some 25% to 30% more energy from the original uranium in the process and thus contributing to energy security and (2) to reduce the volume and radioactivity of the waste by recovering all long-lived actinides and fission products followed by recycling them in fast reactors where they are transmuted to short-lived fission products; this reduces the volume to about 20%, reduces the long-term radioactivity level in the high-level waste, and complicates the possibility of the plutonium being diverted from civil use – thereby increasing the proliferation resistance of the fuel cycle. In general, reprocessing schemes can be divided into two large categories: aqueous/hydrometallurgical systems, and pyrochemical/pyrometallurgical systems. Worldwide processing schemes are dominated by the aqueous (hydrometallurgical) systems. This document provides a historical review of both categories of reprocessing.

  12. Alternatives for nuclear fuel disposal

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Badillo A, V.; Palacios H, J.; Celis del Angel, L., E-mail: ramon.ramirez@inin.gob.m [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2010-10-15

    The spent fuel is one of the most important issues in the nuclear industry, currently spent fuel management is been cause of great amount of research, investments in the construction of repositories or constructing the necessary facilities to reprocess the fuel, and later to recycle the plutonium recovered in thermal reactors. What is the best solution? or, What is the best technology for a specific solution? Many countries have deferred the decision on selecting an option, while other works actively constructing repositories and others implementing the reprocessing facilities to recycle the plutonium obtained from nuclear spent fuel. In Mexico the nuclear power is limited to two reactors BWR type and medium size. So the nuclear spent fuel discharged has been accommodated at reactor's spent fuel pools. Originally these pools have enough capacity to accommodate spent fuel for the 40 years of designed plant operation. However, currently is under process an extended power up rate to 20% of their original power and also there are plans to extend operational life for 20 more years. Under these conditions there will not be enough room for spent fuel in the pools. So this work describes some different alternatives that have been studied in Mexico to define which will be the best alternative to follow. (Author)

  13. Development of Voloxidation Process for Treatment of LWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. J.; Jung, I. H.; Shin, J. M. (and others)

    2007-08-15

    The objective of the project is to develop a process which provides a means to recover fuel from the cladding, and to simplify downstream processes by recovering volatile fission products. This work focuses on the process development in three areas ; the measurement and assessment of the release behavior for the volatile and semi-volatile fission products from the voloxidation process, the assessment of techniques to trap and recover gaseous fission products, and the development of process cycles to optimize fuel cladding separation and fuel particle size. High temperature adsorption method of KAERI was adopted in the co-design of OTS for hot experiment in INL. KAERI supplied 6 sets of filter for hot experiment. Three hot experiment in INL hot cell from the 25th of November for two weeks with attaching 4 KAERI staffs had been carried out. The results were promising. For example, trapping efficiency of Cs was 95% and that of I was 99%, etc.

  14. Analysis of the ATR fuel element swaging process

    Energy Technology Data Exchange (ETDEWEB)

    Richins, W.D.; Miller, G.K.

    1995-12-01

    This report documents a detailed evaluation of the swaging process used to connect fuel plates to side plates in Advanced Test Reactor (ATR) fuel elements. The swaging is a mechanical process that begins with fitting a fuel plate into grooves in the side plates. Once a fuel plate is positioned, a lip on each of two side plate grooves is pressed into the fuel plate using swaging wheels to form the joints. Each connection must have a specified strength (measured in terms, of a pullout force capacity) to assure that these joints do not fail during reactor operation. The purpose of this study is to analyze the swaging process and associated procedural controls, and to provide recommendations to assure that the manufacturing process produces swaged connections that meet the minimum strength requirement. The current fuel element manufacturer, Babcock and Wilcox (B&W) of Lynchburg, Virginia, follows established procedures that include quality inspections and process controls in swaging these connections. The procedures have been approved by Lockheed Martin Idaho Technologies and are designed to assure repeatability of the process and structural integrity of each joint. Prior to July 1994, ATR fuel elements were placed in the Hydraulic Test Facility (HTF) at the Idaho National Engineering Laboratory (AGNAIL), Test Reactor Area (TRA) for application of Boehmite (an aluminum oxide) film and for checking structural integrity before placement of the elements into the ATR. The results presented in this report demonstrate that the pullout strength of the swaged connections is assured by the current manufacturing process (with several recommended enhancements) without the need for- testing each element in the HTF.

  15. Multivariate analysis of gamma spectra to characterize used nuclear fuel

    Science.gov (United States)

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    2017-04-01

    The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. This approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.

  16. Multivariate analysis of gamma spectra to characterize used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    2017-04-01

    Abstract—The Multi-Isotope Process (MIP) Monitor provides an efficient approach to monitoring the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of reprocessing streams in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type. Locally weighted PLS models were fitted on-the-fly to estimate continuous fuel characteristics. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE. This automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters and material diversions.

  17. National nuclear energy policy and Community law. Germany`s international commitments due to the EURATOM treaty and membership in the EC and their possible effects on a national policy for abandonment of spent fuel reprocessing and a phase-out of nuclear power; Nationale Kernenergiepolitik und Gemeinschaftsrecht. Die Bindungen des Euratom- und des EG-Vertrages fuer einen Verzicht auf die Wiederaufarbeitung und einen Ausstieg aus der wirtschaftlichen Nutzung der Kernenergie

    Energy Technology Data Exchange (ETDEWEB)

    Wahl, R.; Hermes, G.

    1995-08-01

    The spent fuel management concept of direct ultimate disposal is compatible in its fundamental features with the law of the European Community. This applies to a national law prohibiting spent fuel reprocessing and the handling of plutonium in Germany, imposing restrictions on exporting spent fuel assemblies and importing plutonium and reprocessing remnants, and on power plant operators to employ reprocessing services abroad. Also, a nuclear power phase-out decided by the German government would in principle not mean a breach of the EURATOM treaty. (orig./HP) [Deutsch] Als Alternative zur geltenden Rechtslage werden im Deutschen Rundestag Aenderungen vorgeschlagen, die das Entsorgungskonzept der direkten Endlagerung - d.h. ein Verbot der Wiederaufarbeitung von Brennelementen aus deutschen Kernkraftwerken auch im europaeischen Ausland - vorschreiben und einen ``Ausstieg`` aus der Kernenergienutzung anordnen. Vor dem Hintergrund einer Analyse der tatsaechlichen Situation untersucht die Studie, ob sich die Bundesrepublik Deutschland in Widerspruch zu ihren Verpflichtungen als Mitglied der Europaeischen Union, insbesondere der Europaeischen Atomgemeinschaft, setzen wuerde, wenn die Gesetzgebungsorgane diesen Aenderungsvorschlaegen folgen wuerden. Die Grundsatzfrage danach, ob es mit dem Recht der Euratom-Gemeinschaft vereinbar ist, wenn ein Mitgliedstaat sich gegen die wirtschaftliche Nutzung der Kernenergie entscheidet, wird bejaht. (orig./HP)

  18. Design of demonstration facility for advanced spent fuel conditioning process

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, W. M.; Koo, J. H.; Jeo, I. J.; Kook, D. H.; Lee, E. P.; Baek, S. R.; Lee, K. I.; You, K. S.; Park, S. W. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    The Advanced spent fuel Conditioning Process(ACP) was proposed and developed for effective management of the PWR spent fuel. The detail plan was established for demonstration and verification of the ACP, and an existing hot cell will be modified as {alpha}-{gamma} type hot cell. In this study, the process mechanical flow was analysed for the optimum arrangement to ensure effective process operation in hot cell, and the detail design of hot cell including the auxiliary facility and safety analysis was performed to secure conservative safety of hot cell system. And then, this results will be utilized for hot cell refurbishment and license.

  19. Microwave Processing of Simulated Advanced Nuclear Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    D.E. Clark; D.C. Folz

    2010-08-29

    Throughout the three-year project funded by the Department of Energy (DOE) and lead by Virginia Tech (VT), project tasks were modified by consensus to fit the changing needs of the DOE with respect to developing new inert matrix fuel processing techniques. The focus throughout the project was on the use of microwave energy to sinter fully stabilized zirconia pellets using microwave energy and to evaluate the effectiveness of techniques that were developed. Additionally, the research team was to propose fundamental concepts as to processing radioactive fuels based on the effectiveness of the microwave process in sintering the simulated matrix material.

  20. Nuclear fuels: Development, processing and disposal

    Energy Technology Data Exchange (ETDEWEB)

    Allday, C.

    1982-08-01

    The successful development of the world's energy resources has enabled industries in the more advanced countries to provide the economic basis on which improved living standards are based. As the less well-developed countries seek to improve their standards of living the pressure on existing energy resources will increase. In this context it is essential not to allow the current industrial recession in the developed countries, with its associated apparent abundancy of coal, oil and gas, to mask the longer-term energy situation. It is not here proposed to discuss the role of nuclear power in the energy scene except to say that, with the continuing need to develop energy resources, nuclear as a proven safe and economic system - will have a vital role to fulfil in meeting the world's future energy demands. This paper is concerned with the development of nuclear fuel and the industry which has grown around it during the last 30 years. It shall concentrate on its development in this country and describe the history and activities of BNFL.

  1. Closed Fuel Cycle Waste Treatment Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, J. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Collins, E. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Crum, J. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, S. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Garn, T. G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gombert, D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maio, V. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Matyas, J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Nenoff, T. M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Riley, B. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sevigny, G. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Thallapally, P. K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, J. H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-02-01

    with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

  2. Novel Redox Processes for Carbonaceous Fuel Conversion

    Science.gov (United States)

    He, Feng

    The current study investigates oxygen carrier development, process intensification, and oxygen carrier attrition behaviors for a number of novel, redox-based energy conversion schemes. (Abstract shortened by ProQuest.).

  3. Fully integrated safeguards and security for reprocessing plant monitoring.

    Energy Technology Data Exchange (ETDEWEB)

    Duran, Felicia Angelica; Ward, Rebecca; Cipiti, Benjamin B.; Middleton, Bobby D.

    2011-10-01

    Nuclear fuel reprocessing plants contain a wealth of plant monitoring data including material measurements, process monitoring, administrative procedures, and physical protection elements. Future facilities are moving in the direction of highly-integrated plant monitoring systems that make efficient use of the plant data to improve monitoring and reduce costs. The Separations and Safeguards Performance Model (SSPM) is an analysis tool that is used for modeling advanced monitoring systems and to determine system response under diversion scenarios. This report both describes the architecture for such a future monitoring system and present results under various diversion scenarios. Improvements made in the past year include the development of statistical tests for detecting material loss, the integration of material balance alarms to improve physical protection, and the integration of administrative procedures. The SSPM has been used to demonstrate how advanced instrumentation (as developed in the Material Protection, Accounting, and Control Technologies campaign) can benefit the overall safeguards system as well as how all instrumentation is tied into the physical protection system. This concept has the potential to greatly improve the probability of detection for both abrupt and protracted diversion of nuclear material.

  4. Digital mock-up for the spent fuel disassembly processes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Kim, S. H.; Song, T. G.; Kim, Y. H.; Hong, D. H.; Yoon, J. S

    2000-12-01

    In this study, the graphical design system is developed and the digital mock-up is implemented for designing the spent fuel handling and disassembly processes. The system consists of a 3D graphical modeling system, a devices assembling system, and a motion simulation system. This system is used throughout the design stages from the conceptual design to the motion analysis. By using this system, all the process involved in the spent fuel handling and disassembly processes are analyzed and optimized. Also, this system is used in developing the on-line graphic simulator which synchronously simulates the motion of the equipment in a real time basis by connecting the device controllers with the graphic server through the TCP/IP network. This simulator can be effectively used for detecting the malfunctions of the process equipment which is remotely operated. Thus, the simulator enhances the reliability and safety of the spent fuel handling process by providing the remote monitoring function of the process. The graphical design system and the digital mock-up system can be effectively used for designing the process equipment, as well as the optimized process and maintenance process. And the on-line graphic simulator can be an alternative of the conventional process monitoring system which is a hardware based system.

  5. Proceedings of the workshop on hydrocarbon processing mixing and scale-up problems. [Fuels processing for fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Gabor, J. D. [ed.

    1978-01-01

    A workshop was convened by the Division of Fossil Fuel Utilization of the US Department of Energy in cooperation with the Particulate and Multiphase Process Program of the National Science Foundation to identify needs for fundamental engineering support for the design of chemical reactors for processing heavy hydrocarbon liquids. The problems associated with dispersing liquid hydrocarbons in a reacting gas and mixing within the gas phase are of primary concern. The transactions of the workshop begin with an introduction to the immediate goals of the Department of Energy. Fuel cell systems and current research and development are reviewed. Modeling of combustion and the problems of soot formation and deposits in hydrocarbon fuels are next considered. The fluid mechanics of turbulent mixing and its effect on chemical reactions are then presented. Current experimental work and process development provide an update on the present state-of-the-art.

  6. Advanced Fuels and Combustion Processes for Propulsion

    Science.gov (United States)

    2010-09-01

    production from biomass steam reforming – Conduct a feasibility analysis of the proposed integrated process Energia Technologies - D. Nguyen & K. Parimi...strength foam material development by Ultramet – Combustion experiments performed U. Of Alabama – End-user input provided by Solar Turbines Major

  7. Nuclear fuel materials processing in reactive gas plasma

    Energy Technology Data Exchange (ETDEWEB)

    Min, Jin Young; Yang, Myung Seung; Seo, Yong Dae; Kim Yong Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-07-01

    DUPIC fuel cycle development project in KAERI of Korea was initiated in 1991 and has advanced in relevant technologies for last 10 years. The project includes five different topics such as nuclear fuel manufacturing, compatibility evaluation, performance evaluation, manufacturing facility management, and safeguards. The contents and results of DUPIC R and D up to now are as follow: - the basic foundation was established for the critically required pelletizing technology and powder treatment technology for DUPIC. - development of DUPIC process line and deployment of 20 each process equipment and examination instruments in DFDF. - powder and pellet characterization study was done at PIEF based on the simfuel study results, and 30 DUPIC pellets were successfully produced. - the manufactured pellets were used for sample fuel rods irradiated in July,2000 in HANARO research reactor in KAERI and have been under post irradiation examination. (Hong, J. S.)

  8. DUPIC nuclear fuel manufacturing and process technology development at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yim, Sung Paal; Lee, Jung Won; Kim, Jong Ho; Kim, Soo Sung; Kim, Woong Ki; Yang, Myung Seung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-07-01

    DUPIC fuel cycle development project in KAERI of Korea was initiated in 1991 and has advanced in relevant technologies for last 10 years. The project includes five different topics such as nuclear fuel manufacturing, compatibility evaluation, performance evaluation, manufacturing facility management, and safeguards. The contents and results of DUPIC R and D up to now are as follow: - the basic foundation was established for the critically required pelletizing technology and powder treatment technology for DUPIC. - development of DUPIC process line and deployment of 20 each process equipment and examination instruments in DFDF. - powder and pellet characterization study was done at PIEF based on the simfuel study results, and 30 DUPIC pellets were successfully produced. - the manufactured pellets were used for sample fuel rods irradiated in July,2000 in HANARO research reactor in KAERI and has been under post irradiation examination. (Hong, J. S.)

  9. Sustainable Production of Asphalt using Biomass as Primary Process Fuel

    DEFF Research Database (Denmark)

    Bühler, Fabian; Nguyen, Tuong-Van; Elmegaard, Brian

    2016-01-01

    is the heating and drying of aggregate,where natural gas, fuel oil or LPG is burned in a direct-fired rotary dryer. Replacing this energy source with amore sustainable one presents several technical and economic challenges, as high temperatures, short startuptimes and seasonal production variations are required......The production of construction materials is very energy intensive and requires large quantities of fossil fuels.Asphalt is the major road paving material in Europe and is being produced primarily in stationary batch mixasphalt factories. The production process requiring the most energy....... This paper analyses different pathways for the useof biomass feedstock as a primary process fuel. The analysed cases consider the gasification of straw andwood chips and the direct combustion of wood pellets. The additional use of syngas from the gasifier for theproduction of heat or combined heat and power...

  10. Removal efficiency of silver impregnated filter materials and performance of iodie filters in the off-gas of the Karlsruhe reprocessing plant WAK

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, F.J.; Herrmann, B.; Hoeflich, V. [Wiederaufarbeitungsanlage Karlsruhe (Germany)] [and others

    1997-08-01

    An almost quantitative retention of iodine is required in reprocessing plants. For the iodine removal in the off-gas streams of a reprocessing plant various sorption materials had been tested under realistic conditions in the Karlsruhe reprocessing plant WAK in cooperation with the Karlsruhe research center FZK. The laboratory results achieved with different iodine sorption materials justified long time performance tests in the WAK Plant. Technical iodine filters and sorption materials for measurements of iodine had been tested from 1972 through 1992. This paper gives an overview over the most important results, Extended laboratory, pilot plant, hot cell and plant experiences have been performed concerning the behavior and the distribution of iodine-129 in chemical processing plants. In a conventional reprocessing plant for power reactor fuel, the bulk of iodine-129 and iodine-127 is evolved into the dissolver off-gas. The remainder is dispersed over many aqueous, organic and gaseous process and waste streams of the plant. Iodine filters with silver nitrate impregnated silica were installed in the dissolver off-gas of the Karlsruhe reprocessing plant WAK in 1975 and in two vessel vent systems in 1988. The aim of the Karlsruhe iodine research program was an almost quantitative evolution of the iodine during the dissolution process to remove as much iodine with the solid bed filters as possible. After shut down of the WAK plant in December 1990 the removal efficiency of the iodine filters at low iodine concentrations had been investigated during the following years. 12 refs., 2 figs., 2 tabs.

  11. Effect of fuel size and process temperature on fuel gas quality from CFB gasification of biomass

    Energy Technology Data Exchange (ETDEWEB)

    Van der Drift, A.; Van Doorn, J. [ECN Biomass, Petten (Netherlands)

    2000-07-01

    A bench-scale circulating fluidized bed (CFB) gasifier with a capacity of max. 500 kWh{sub th} has been used to study the effect of fuel size and process temperature. A higher process temperature (range tested: 750 to 910C) results in more air needed to maintain the desired temperature, a lower heating value of the product gas, a higher carbon conversion and a net increase of cold gas efficiency of the gasifier. A higher process temperature also results in less heavy tars. However, light tars (measured using the solid phase adsorbent (SPA) method) do show an odd behaviour. Some individual components within the group of light tars even increase in concentration when process temperature is raised. The main reason probably is that heavy tars decompose to these relatively stable light tar components. The particle size of the fuel does influence some product gas parameters considerably. The presence of small particles seems to increase the (heavy) tar concentration and decrease the conversion of fuel-nitrogen to ammonia. Small particles can also be responsible for large temperature gradients along the axis of the riser of a CFB-gasifier. This effect can be avoided by either mixing the fuel with larger particles or feed the small particles at the bottom of the reactor. 5 refs.

  12. Commercial Nuclear Reprocessing in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Sherrill, Charles Leland [Brigham Young Univ., Provo, UT (United States); Balatsky, Galya Ivanovna [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-09

    The short presentation outline: Reprocessing Overview; Events leading up to Carter’s Policy; Results of the decision; Policy since Nuclear Nonproliferation Act. Conclusions reached: Reprocessing ban has become an easy and visible fix to the public concern about proliferation, but has not completely stopped proliferation; and, Reprocessing needs to become detached from political considerations, so technical research can continue, regardless of the policy decisions we decide to take.

  13. High temperature polymer fuel cells and their Interplay with fuel processing systems

    DEFF Research Database (Denmark)

    Jensen, Jens Oluf; Qingfeng, Li; He, R.

    2003-01-01

    This paper reports recent results from our group on polymer electrolyte membrane fuel cells (PEMFC) based on the temperature resistant polymer polybenzimidazole (PBI), which allow working temperatures up to 200°C. The membrane has a water drag number near zero and need no water management at all....... The high working temperature allows for utilization of the excess heat for fuel processing. Moreover, it provides an excellent CO tolerance of several percent, and the system needs no purification of hydrogen from a reformer. Continuous service for over 6 months at 150°C has been demonstrated....

  14. Pyrolysis process for producing fuel gas

    Science.gov (United States)

    Serio, Michael A. (Inventor); Kroo, Erik (Inventor); Wojtowicz, Marek A. (Inventor); Suuberg, Eric M. (Inventor)

    2007-01-01

    Solid waste resource recovery in space is effected by pyrolysis processing, to produce light gases as the main products (CH.sub.4, H.sub.2, CO.sub.2, CO, H.sub.2O, NH.sub.3) and a reactive carbon-rich char as the main byproduct. Significant amounts of liquid products are formed under less severe pyrolysis conditions, and are cracked almost completely to gases as the temperature is raised. A primary pyrolysis model for the composite mixture is based on an existing model for whole biomass materials, and an artificial neural network models the changes in gas composition with the severity of pyrolysis conditions.

  15. Study of physico-chemical release of uranium and plutonium oxides during the combustion of polycarbonate and of ruthenium during the combustion of solvents used in the reprocessing of nuclear fuel; Etude de la mise en suspension physico-chimique des oxydes de plutonium et d'uranium lors de la combustion de polycarbonate et de ruthenium lors de la combustion des solvants de retraitement du combustible irradie

    Energy Technology Data Exchange (ETDEWEB)

    Bouilloux, L

    1998-07-01

    The level of consequences concerning a fire in a nuclear facility is in part estimated by the quantities and the physico-chemical forms of radioactive compounds that may be emitted out of the facility. It is therefore necessary to study the contaminant release from the fire. Because of the multiplicity of the scenarios, two research subjects were retained. The first one concerns the study of the uranium or plutonium oxides chemical release during the combustion of the polycarbonate glove box sides. The second one is about the physico chemical characterisation of the ruthenium release during the combustion of an organic solvent mixture (tributyl phosphate-dodecane) used for the nuclear fuel reprocessing. Concerning the two research subjects, the chemical release, i.e. means the generation of contaminant compounds gaseous in the fire, was modelled using thermodynamical simulations. Experiments were done in order to determine the ruthenium release factor during solvent combustion. A cone calorimeter was used for small scale experiments. These results were then validated by large scale tests under conditions close to the industrial process. Thermodynamical simulations, for the two scenarios studied. Furthermore, the experiments on solvent combustion allowed the determination of a suitable ruthenium release factor. Finally, the mechanism responsible of the ruthenium release has been found. (author)

  16. Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media

    Science.gov (United States)

    Leturcq, G.; Grandjean, A.; Rigaud, D.; Perouty, P.; Charlot, M.

    2005-12-01

    Spent nuclear fuel reprocessing to recover energy-producing elements such as uranium or plutonium can be performed by a pyrochemical process. In such method, the actinides and fission products are extracted by electrodeposition in a molten chloride medium. These processes generate chlorinated alkali salt flows contaminated by fission products, mainly Cs, Ba, Sr and rare earth elements constituting high-level waste. Two possible alternatives are investigated for managing this wasteform; a protocol is described for dechlorinating the fission products to allow vitrification, and mineral phases capable of immobilizing chlorides are listed to allow specification of a dedicated ceramic matrix suitable for containment of these chlorinated waste streams. The results of tests to synthesize chlorosilicate phases are also discussed.

  17. Defining the Costs of Reusable Flexible Ureteroscope Reprocessing Using Time-Driven Activity-Based Costing.

    Science.gov (United States)

    Isaacson, Dylan; Ahmad, Tessnim; Metzler, Ian; Tzou, David T; Taguchi, Kazumi; Usawachintachit, Manint; Zetumer, Samuel; Sherer, Benjamin; Stoller, Marshall; Chi, Thomas

    2017-09-20

    Careful decontamination and sterilization of reusable flexible ureteroscopes used in ureterorenoscopy cases prevent the spread of infectious pathogens to patients and technicians. However, inefficient reprocessing and unavailability of ureteroscopes sent out for repair can contribute to expensive operating room (OR) delays. Time-driven activity-based costing (TDABC) was applied to describe the time and costs involved in reprocessing. Direct observation and timing were performed for all steps in reprocessing of reusable flexible ureteroscopes following operative procedures. Estimated times needed for each step by which damaged ureteroscopes identified during reprocessing are sent for repair were characterized through interviews with purchasing analyst staff. Process maps were created for reprocessing and repair detailing individual step times and their variances. Cost data for labor and disposables used were applied to calculate per minute and average step costs. Ten ureteroscopes were followed through reprocessing. Process mapping for ureteroscope reprocessing averaged 229.0 ± 74.4 minutes, whereas sending a ureteroscope for repair required an estimated 143 minutes per repair. Most steps demonstrated low variance between timed observations. Ureteroscope drying was the longest and highest variance step at 126.5 ± 55.7 minutes and was highly dependent on manual air flushing through the ureteroscope working channel and ureteroscope positioning in the drying cabinet. Total costs for reprocessing totaled $96.13 per episode, including the cost of labor and disposable items. Utilizing TDABC delineates the full spectrum of costs associated with ureteroscope reprocessing and identifies areas for process improvement to drive value-based care. At our institution, ureteroscope drying was one clearly identified target area. Implementing training in ureteroscope drying technique could save up to 2 hours per reprocessing event, potentially preventing expensive OR delays.

  18. Process development and fabrication for sphere-pac fuel rods. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted.

  19. The use of electro spray mass spectrometry for the determination of dissolved species application for determination of zirconium complexes in reprocessing spent fuel matrice; Electrospray/spectrometrie de masse, technique d'avenir pour l'etude des complexes. Applications aux systemes U(6)/Zr(4) dans les conditions simulees du procede Purex

    Energy Technology Data Exchange (ETDEWEB)

    Lamouroux, Ch.; Moulin, Ch. [CEA Saclay, Dept. des Procedes d' Enrichissement (DCC/DPE/SPCP), 91 - Gif-sur-Yvette (France); Blanc, P. [CEA Valrho, (DCC/DRRV/SEMP), 30 - Marcoule (France)

    2000-07-01

    Liquid-liquid extraction of Zirconium one of the most important fission products, was investigated by Electro-spray Mass Spectrometry (ESI/MS) in simulated nuclear reprocessing spent fuel process conditions. Zr{sup IV} can precipitate at the organic / water interface after its extraction by dibutyl-phosphoric acid (HDBP), the most common degradation product of tributylphosphate (TBP) under radiolysis. However, precipitation of ZrI{sup IV} is restricted and particularly dependent on the ratio 'r': (HDBP){sub tot}/(Zr{sup IV}]{sub tot}. The type and characterization of the precipitate is reported in different papers as Zr(NO{sub 3}){sub 2}(DBP){sub 2}. complex. However, some uncertainties exist about the composition and structures of the dissolved Zr species extracted. Techniques already used for such purposes are NMR (Nuclear Magnetic Resonance) and vibrational spectroscopy, but identification of the extracted metal complex structures is debatable. To obtain more definitive information, the use of ESI/MS could be a complementary tool for characterizing the extracted metal complexes. ESI allows ionization/desorption of non-volatile analytes into gas phase detected by mass spectrometry with high sensitivity, making it a complementary tool for examining the speciation of dissolved metal species. Extractions were carried out for the system (ZrI{sup IV} in HNO{sub 3} 3M)/(TBP/C{sub 12}H{sub 26} 30/70 spiked with HDBP) by varying the ratio r. ZrI{sup IV} extraction was confirmed by ICP-AES (Inductively Coupled plasma-Atomic Emission Spectroscopy) measurements on the aqueous phase, and dissolved metal complexes were identified by ESI/MS experiments on the organic phase. Different complexes could be detected with ESI used in positive and negative ion mode as a function of the extraction conditions such as the ratio r. Good agreement is observed between the variation in mass spectra and Zr behavior described for solutions. For a ratio 0

  20. Hydrocarbon fuel processing of micro solid oxide fuel cell systems[Dissertation 17455

    Energy Technology Data Exchange (ETDEWEB)

    Stutz, M. J.

    2007-07-01

    The scope of this thesis is the numerical and experimental investigation of the fuel processing of a micro solid oxide fuel cell (SOFC) running on hydrocarbon fuel. The goal is to enhance the overall system efficiency by optimization of the reforming process in the steady state and the improvement of the start-up process. Micro SOFC are a potential alternative to the currently used batteries in portable devices. Liquid butane in a cartridge could be the energy source. This dissertation is focused on the fuel processing of the system, namely the reforming and post-combusting processes. The reformer converts the hydrocarbon fuel to a hydrogen rich gas that can be utilized by the SOFC. The post-combustor depletes the toxic and/or explosive gases before leaving the exhaust. Chapter One presents a short introduction to the field of hydrocarbon fuel processing in micro solid oxide fuel cell systems, the next three chapters deal with computational modeling of the transport phenomena inside a micro-reformer, which leads to a better understanding of the chemistry and the physics therein, hence progress in the design and operation parameters. The experimental part (i.e. Chapter Five) of this thesis focuses on the feasibility of a novel hybrid start-up method of a fuel cell system that employs existing components as an additional heat source. In Chapter Two the effect of wall heat conduction on the syngas (hydrogen and carbon monoxide) production of a micro-reformer, representing micro-fabricated channels or monoliths, is investigated. Methane is used as a model hydrocarbon fuel since its heterogeneous reaction path on rhodium is known and validated. The simulations demonstrate that the axial wall conduction strongly influences the performance of the micro-reformer and should not be neglected without a careful a priori investigation of its impact. Methane conversion and hydrogen yield are strongly dependent of the wall inner surface temperature, which is influenced by the

  1. Hydrocarbon fuel processing of micro solid oxide fuel cell systems[Dissertation 17455

    Energy Technology Data Exchange (ETDEWEB)

    Stutz, M. J.

    2007-07-01

    The scope of this thesis is the numerical and experimental investigation of the fuel processing of a micro solid oxide fuel cell (SOFC) running on hydrocarbon fuel. The goal is to enhance the overall system efficiency by optimization of the reforming process in the steady state and the improvement of the start-up process. Micro SOFC are a potential alternative to the currently used batteries in portable devices. Liquid butane in a cartridge could be the energy source. This dissertation is focused on the fuel processing of the system, namely the reforming and post-combusting processes. The reformer converts the hydrocarbon fuel to a hydrogen rich gas that can be utilized by the SOFC. The post-combustor depletes the toxic and/or explosive gases before leaving the exhaust. Chapter One presents a short introduction to the field of hydrocarbon fuel processing in micro solid oxide fuel cell systems, the next three chapters deal with computational modeling of the transport phenomena inside a micro-reformer, which leads to a better understanding of the chemistry and the physics therein, hence progress in the design and operation parameters. The experimental part (i.e. Chapter Five) of this thesis focuses on the feasibility of a novel hybrid start-up method of a fuel cell system that employs existing components as an additional heat source. In Chapter Two the effect of wall heat conduction on the syngas (hydrogen and carbon monoxide) production of a micro-reformer, representing micro-fabricated channels or monoliths, is investigated. Methane is used as a model hydrocarbon fuel since its heterogeneous reaction path on rhodium is known and validated. The simulations demonstrate that the axial wall conduction strongly influences the performance of the micro-reformer and should not be neglected without a careful a priori investigation of its impact. Methane conversion and hydrogen yield are strongly dependent of the wall inner surface temperature, which is influenced by the

  2. Recovery of minor actinides from irradiated superfact fuels

    Energy Technology Data Exchange (ETDEWEB)

    Apoltolidis, C.; Glatz, J.P.; Molinet, R.; Nicholl, A.; Pagliosa, G.; Romer, K.; Bokelund, H.; Koch, L. [European Commission, JRC, Institute fuer Transuranium Elements, Karlsruhe (Germany)

    1995-12-31

    It could be demonstrated that the reprocessing of fast reactor oxide fuels containing up to 45 % MA (Np and Am), irradiated in the PHENIX reactor in the frame of a transmutation study, is possible. The fuels were dissolved under PUREX type conditions in order to determine their behaviour in the head-end step of the reprocessing process. For one of the fuels containing 20 % Am and 20 % Np before irradiation, an almost complete partitioning of actinides from the dissolver solution could be achieved. Chromatographic extraction was used for the separation of the main bulk elements U, Pu and Np, whereas centrifugal extractors were used to separate the minor actinides from the remaining high level liquid wastes (HLLW). For the relevant radio-toxic isotopes a high recovery rate from the irradiation targets was reached. Those elements are thus available for new fuel fabrication. (authors) 12 refs.

  3. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    Directory of Open Access Journals (Sweden)

    Peel Ross

    2016-01-01

    Full Text Available This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mixed oxide (MOX fuel of plutonium in depleted uranium, within the enhanced CANDU-6 (EC-6 reactor. This work proposes an alternative heterogeneous fuel concept based on the same reactor and CANFLEX fuel bundle, with eight large-diameter fuel elements loaded with natural thorium oxide and 35 small-diameter fuel elements loaded with a MOX of plutonium and reprocessed uranium stocks from UK MAGNOX and AGR reactors. Indicative neutronic calculations suggest that such a fuel would be neutronically feasible. A similar MOX may alternatively be fabricated from reprocessed <5% enriched light water reactor fuel, such as the fuel of the AREVA EPR reactor, to consume newly produced plutonium from reprocessing, similar to the DUPIC (direct use of PWR fuel in CANDU process.

  4. Modular, High-Volume Fuel Cell Leak-Test Suite and Process

    Energy Technology Data Exchange (ETDEWEB)

    Ru Chen; Ian Kaye

    2012-03-12

    Fuel cell stacks are typically hand-assembled and tested. As a result the manufacturing process is labor-intensive and time-consuming. The fluid leakage in fuel cell stacks may reduce fuel cell performance, damage fuel cell stack, or even cause fire and become a safety hazard. Leak check is a critical step in the fuel cell stack manufacturing. The fuel cell industry is in need of fuel cell leak-test processes and equipment that is automatic, robust, and high throughput. The equipment should reduce fuel cell manufacturing cost.

  5. X-ray reprocessing in binaries

    Science.gov (United States)

    Paul, Biswajit

    2016-07-01

    We will discuss several aspects of X-ray reprocessing into X-rays or longer wavelength radiation in different kinds of binary systems. In high mass X-ray binaries, reprocessing of hard X-rays into emission lines or lower temperature black body emission is a useful tool to investigate the reprocessing media like the stellar wind, clumpy structures in the wind, accretion disk or accretion stream. In low mass X-ray binaries, reprocessing from the surface of the companion star, the accretion disk, warps and other structures in the accretion disk produce signatures in longer wavelength radiation. X-ray sources with temporal structures like the X-ray pulsars and thermonuclear burst sources are key in such studies. We will discuss results from several new investigations of X-ray reprocessing phenomena in X-ray binaries.

  6. The nuclear fuel cycle; Le cycle du combustible nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-05-01

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

  7. The Thermal Stability of Tertiary Pyridine Resin for the Application to Multi-functional Reprocessing Process - Adv.-ORIENT Cycle Development

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Yoshihiko; Okada, K.; Akiyoshi, M.; Matsunaga, T. [AIST Tsukuba Central 5, Higashi 1-1-1, Tsukuba, Ibaraki 305-8565 (Japan); Suzuki, T. [Tokyo Tech (Japan); Koyama, S.; Ozawa, M. [Japan Atomic Energy Agency - JAEA (Japan)

    2009-06-15

    As part of 'Adv.-ORIENT' (Advanced Optimization by Recycling Instructive Elements) cycle technologies which aim to develop new fuel cycle based on FBR (Fast Breeder Reactor), the fundamental thermochemical properties of tertiary pyridine resin (TPR) and its mixtures with methanol/HCl and HNO{sub 3} were investigated and heating tests on gram scale with TPR/methanol/HNO{sub 3} were carried out in order to evaluate the thermal stability of TPR and to determine the conditions necessary to avoid runaway reactions. It was found that TPR with HCl was thermally stable. Evident thermal decomposition peaks were identified with TPR in the presence of concentrated HNO{sub 3}. No specific effect was observed for methanol involving. However, it was considered that the rapidly exothermic reaction can be controlled by heating temperature. (authors)

  8. Understanding the transport processes in polymer electrolyte membrane fuel cells

    Science.gov (United States)

    Cheah, May Jean

    Polymer electrolyte membrane (PEM) fuel cells are energy conversion devices suitable for automotive, stationary and portable applications. An engineering challenge that is hindering the widespread use of PEM fuel cells is the water management issue, where either a lack of water (resulting in membrane dehydration) or an excess accumulation of liquid water (resulting in fuel cell flooding) critically reduces the PEM fuel cell performance. The water management issue is addressed by this dissertation through the study of three transport processes occurring in PEM fuel cells. Water transport within the membrane is a combination of water diffusion down the water activity gradient and the dragging of water molecules by protons when there is a proton current, in a phenomenon termed electro-osmotic drag, EOD. The impact of water diffusion and EOD on the water flux across the membrane is reduced due to water transport resistance at the vapor/membrane interface. The redistribution of water inside the membrane by EOD causes an overall increase in the membrane resistance that regulates the current and thus EOD, thereby preventing membrane dehydration. Liquid water transport in the PEM fuel cell flow channel was examined at different gas flow regimes. At low gas Reynolds numbers, drops transitioned into slugs that are subsequently pushed out of the flow channel by the gas flow. The slug volume is dependent on the geometric shape, the surface wettability and the orientation (with respect to gravity) of the flow channel. The differential pressure required for slug motion primarily depends on the interfacial forces acting along the contact lines at the front and the back of the slug. At high gas Reynolds number, water is removed as a film or as drops depending on the flow channel surface wettability. The shape of growing drops at low and high Reynolds number can be described by a simple interfacial energy minimization model. Under flooding conditions, the fuel cell local current

  9. Technical and regulatory review of the Rover nuclear fuel process for use on Fort St. Vrain fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hertzler, T. [Science Applications International Corp., Idaho Falls, ID (United States)

    1993-02-01

    This report describes the results of an analysis for processing and final disposal of Fort St. Vrain (FSV) irradiated fuel in Rover-type equipment or technologies. This analysis includes an evaluation of the current Rover equipment status and the applicability of this technology in processing FSV fuel. The analyses are based on the physical characteristics of the FSV fuel and processing capabilities of the Rover equipment. Alternate FSV fuel disposal options are also considered including fuel-rod removal from the block, disposal of the empty block, or disposal of the entire fuel-containing block. The results of these analyses document that the current Rover hardware is not operable for any purpose, and any effort to restart this hardware will require extensive modifications and re-evaluation. However, various aspects of the Rover technology, such as the successful fluid-bed burner design, can be applied with modification to FSV fuel processing. The current regulatory climate and technical knowledge are not adequately defined to allow a complete analysis and conclusion with respect to the disposal of intact fuel blocks with or without the fuel rods removed. The primary unknowns include the various aspects of fuel-rod removal from the block, concentration of radionuclides remaining in the graphite block after rod removal, and acceptability of carbon in the form of graphite in a high level waste repository.

  10. Towards Extrusion of Ionomers to Process Fuel Cell Membranes

    Directory of Open Access Journals (Sweden)

    Jean-Yves Sanchez

    2011-07-01

    Full Text Available While Proton Exchange Membrane Fuel Cell (PEMFC membranes are currently prepared by film casting, this paper demonstrates the feasibility of extrusion, a solvent-free alternative process. Thanks to water-soluble process-aid plasticizers, duly selected, it was possible to extrude acidic and alkaline polysulfone ionomers. Additionally, the feasibility to extrude composites was demonstrated. The impact of the plasticizers on the melt viscosity was investigated. Following the extrusion, the plasticizers were fully removed in water. The extrusion was found to impact neither on the ionomer chains, nor on the performances of the membrane. This environmentally friendly process was successfully validated for a variety of high performance ionomers.

  11. Technology of the light water reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Wymer, R. G.

    1979-01-01

    This essay presents elements of the processes used in the fuel cycle steps and gives an indication of the types of equipment used. The amounts of radioactivity released in normal operation of the processes are indicated and related to radiation doses. Types and costs of equipment or processes required to lower these radioactivity releases are in some cases suggested. Mining and milling, conversion of uranium concentrate to UF/sub 6/, uranium isotope separation, LWR fuel fabrication, fuel reprocessing, transportation, and waste management are covered in this essay. 40 figures, 34 tables. (DLC)

  12. The cost of spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Palacios H, J. C.; Badillo, V.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    Spent fuel is one of the most important issues in the nuclear industry, currently spent fuel management is been cause of great amount of research, investments, constructing repositories or constructing the necessary facilities to reprocess the fuel, and later to recycle the plutonium recovered in thermal reactors. What is the best solution?, or What is the best technology for an specific solution? Many countries have deferred the decision on selecting an option, while others works actively constructing repositories and others implementing the reprocessing facilities to recycle the plutonium obtained from nuclear spent fuel. In Mexico the nuclear power is limited to two reactors BWR type and medium size. So the nuclear spent fuel discharged has been accommodated at reactor's spent fuel pools. Originally these pools have enough capacity to accommodate spent fuel for the 40 years of designed plant operation. However currently, the plants are under a process for extended power up-rate to 20% of original power and also there are plans to extended operational life for 20 more years. Under these conditions there will not be enough room for spent fuel in the pools. (Author)

  13. Nonproliferation and safeguards aspects of the DUPIC fuel cycle concept

    Energy Technology Data Exchange (ETDEWEB)

    Persiani, P. K. [Argonne National Lab., IL (United States)

    1997-07-01

    The purpose of the study is to comment on the proliferation characteristic profiles of some of the proposed fuel cycle alternatives to help ensure that nonproliferation concerns are introduced into the early stages of a fuel cycle concept development program, and to perhaps aid in the more effective implementation of the international nonproliferation regime initiative and safeguards systems. Alternative recycle concepts proposed by several countries involve the recycle of spent fuel without the separation of plutonium from uranium and fission products. The concepts are alternatives to either the direct long-term storage deposition of or the purex reprocessing of the spent fuels. The alternate fuel cycle concepts reviewed include: the dry-recycle processes such as the direct use of reconfigured PWR spent fuel assemblies into CANDU reactors(DUPIC); low-decontamination, single-cycle co-extraction of fast reactor fuels in a wet-purex type of reprocessing; and on a limited scale the thorium-uranium fuel cycle. The nonproliferation advantages usually associated with the above non-separation processes are: the highly radioactive spent fuel presents a barrier to the physical diversion of the nuclear material; avoid the need to dissolve and chemically separate the plutonium from the uranium and fission products; and that the spent fuel isotopic quality of the plutonium vector is further degraded. Although the radiation levels and the need for reprocessing may be perceived as barriers to the terrorist or the subnational level of safeguards, the international level of nonproliferation concerns is addressed primarily by material accountancy and verification activities. On the international level of nonproliferation concerns, the non-separation fuel cycle concepts involved have to be evaluated on the bases of the impact the processes may have on nuclear materials accountancy. (author).

  14. Use of Hansen Solubility Parameters in Fuel Treatment Processes

    Science.gov (United States)

    2014-03-17

    Charts 3. DATES COVERED (From - To) Jan 2014- Mar 2014 4. TITLE AND SUBTITLE 5a. CONTRACT NUMBER In-House Use of Hansen Solubility Parameters in...distribution is unlimited. AFRL Public Affairs Clearance # USE OF HANSEN SOLUBILITY PARAMETERS IN FUEL TREATMENT PROCESSES 17 March 2014 Andrew J...Treatment Needs – Hansen Solubility Parameters • Dyes – Experimental HSP Determination – Extrapolation to Other Dyes • Predictions for Extraction Fluids

  15. Combination and long term stability of the IGS Reprocessing campaign

    Science.gov (United States)

    Booker, David; Clarke, Peter J.; Lavallée, David A.

    2010-05-01

    During the relatively short life of the Global Positioning System (GPS) there have been several changes to the analysis procedure, leading to inhomogeneous coordinate time series. Although they have reduced systematic errors in more recent solutions, these changes have modified the apparent periodic signals observed and led to spurious discontinuities. The International GNSS Service (IGS) reprocessing campaign uses the latest operational models and techniques to reprocess the back catalogue of GPS data to produce remove inconsistencies caused by the various model changes, thus producing a homogeneous time series of station coordinates and Earth Rotation Parameters (ERPs). Weekly coordinate and ERP solutions from up to 11 reprocessing analysis centres (ACs) have been aligned to the ITRF and combined using the TANYA software in a rigorous weighted least-squares solution. Analysis of the time series of station coordinates and Helmert transformation parameters between the combined solution and the ITRF shows a at least a 50 percent improvement in the stability of the reprocessed weekly solutions compared with earlier operational products. There is a gradual decrease in the weighted root mean square coordinate difference, both between the combined weekly solutions and the ITRF and between the individual AC solutions and their weekly combination, which reaches a minimum around the end of 2005 with a slight increase thereafter. We observe clear differences in the periodicity of Helmert transformation parameters between the individual AC solutions and the combined solution, which presumably result from variations in AC processing strategy. There is a clear annual or near annual periodic variation in the scale difference between the combined solution and the ITRF05 and some less clear variation between the translation parameters, which needs further analysis as to its cause. Keywords: GPS, ITRF, IGS reprocessing campaign, periodic errors

  16. Physical Simulation of Burning Process of Alternative Engine Fuels

    Directory of Open Access Journals (Sweden)

    M. S. Assad

    2008-01-01

    Full Text Available Visualization of burning process in the closed vessel has been fulfilled with the help of method high-speed photography through a transparent glass. This method as an efficient means for investigation of fast processes permits to obtain a visual, convenient visual perception insight about the development of the burning process and understand peculiarities of the development of flame in the closed vessels.The paper contains a description of an experimental stand and methodology for execution of an experiment on visualization of the flame development and measurement of main parameters of the burning process in a closed vessel that is in the simulating combustion chamber.According to the obtained photos an analysis of form, structure and dynamics of flame front development has been carried out; some peculiarities and differences of flames of various fuel-air mixtures have been established and the paper proves an occurrence of the secondary glow during burning in the closed vessel.Body of data obtained with the help of the visualization of burning process makes it possible to determine main parameters of the burning process. In particular, relation of the pressure developed in the chamber with the mass of burnt-out mixture has been investigated and dependence has been obtained that shows the law of fuel burning-out in the graphic form.

  17. Characterization of used nuclear fuel with multivariate analysis for process monitoring

    Science.gov (United States)

    Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.; Schwantes, Jon M.

    2014-01-01

    This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict used nuclear fuel burnup. Nuclide activities for prototypic used fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.

  18. 77 FR 38789 - Notice of Availability of Draft Waste Incidental to Reprocessing Evaluation for the Concentrator...

    Science.gov (United States)

    2012-06-29

    ... vitrifying waste from reprocessing of spent nuclear fuel and certain treatment material at the West Valley... canisters where the mixture hardened into a solid glass waste form. DOE operated the vitrification system... of Chapter IV of DOE Manual 435.1-1, provided the waste will be incorporated in a solid physical...

  19. The Separation Method of Neptunium in Reprocessed Uranium Product by TEVA-UTEVA Column Extraction Chromatography

    Institute of Scientific and Technical Information of China (English)

    JIN; Hua; SU; Yu-lan; YING; Zhe-cong; ZHAO; Sheng-yang

    2012-01-01

    <正>237Np, as a highly toxic nuclide, is limited strictly in the final uranium product of spent nuclear fuel reprocessing plant. Due to the low concentration level of 237Np, which is lower than 2.5 μg/g U, its accurate measurement is one of the most difficult analytical works in

  20. Strategic research of advanced fuel cycle technologies in JNC

    Energy Technology Data Exchange (ETDEWEB)

    Kawata, T.; Fukushima, M.; Nomura, S. [Japan Nuclear Cycle Development Institute, Tokai Works (Japan)

    2000-07-01

    Key technologies for the future nuclear fuel cycle have been proposed and are being reviewed in JNC as a part of the Feasibility Study for an Advanced Fuel Cycle, which is to achieve a more flexible energy choice to satisfy a sustainable energy security and global environmental protection. The candidate reprocessing technologies are: 1) aqueous simplified PUREX process, 2) oxide or metallic electrowinning, and 3) fluoride volatilization for oxide, metal, or nitride fuels. The fuel fabrication methods being investigated are: 1) simplified pellet process, 2) sphere/vibro-packed process for MOX/MN fuel, and 3) casting for metal fuel. These candidate technologies are currently being compared based on past experiences, technical issues to be solved, industrial applicability for future plants, feasible options for MA/LLFP separation, and nonproliferation aspects. Alter two years of the present reviewing process, selected key technologies will be developed over the next five years to evaluate industrial applicability of reprocessing and fuel manufacturing processes for the advanced fuel cycle. (authors)

  1. The Fourth (A)ATSR Data Reprocessing

    Science.gov (United States)

    Goryl, Philippe; Cocevar, Pauline; Done, Fay; Aatsr Quality Working Group

    2016-08-01

    This paper aims to inform users of the upcoming Fourth Reprocessing of ATSR-1, ATSR-2 and AATSR data. The main objective of the Fourth Reprocessing is to generate (A)ATSR Level 1B data products in a similar format to SLSTR products from Sentinel-3. In this way, users can easily access the 20-year dataset from the ERS and ENVISAT (A)ATSR missions and carry the analysis forward into the Sentinel era. In addition to the product format change, the dataset will build on the improvements implemented in the Third Reprocessing, and will contain further improvements and enhancements, as described below.

  2. Oxidative dissolution of spent nuclear fuel in aqueous alkaline solutions - An alternative to the Purex process?

    Energy Technology Data Exchange (ETDEWEB)

    Runde, Wolfgang; Peper, Shane; Brodnax, Lia; Crooks, William; Zehnder, Ralph; Jarvinen, Gordon

    2004-07-01

    As an alternative to acidic reprocessing of spent nuclear, oxidative dissolution of UO{sub 2} into aqueous alkaline solutions and subsequent separation of fission products is considered. The efficacy of such a method is limited by the kinetics of the UO{sub 2} dissolution and the capacity of alkaline solutions for dissolved U(VI) species. We performed a series of dissolution studies on UO{sub 2} and U{sub 3}O{sub 8} in aqueous alkaline solutions applying various oxidants. Among the oxidative agents commonly used to transform low-valence actinides into their higher oxidation states, H{sub 2}O{sub 2} has proven to be the most effective in basic media. Consequently, we investigated the dissolution of UO{sub 2} and U{sub 3}O{sub 8} in NaOH-H{sub 2}O{sub 2} and Na{sub 2}CO{sub 3}-H{sub 2}O{sub 2} solutions and determined the dissolution kinetics as a function of peroxide and hydroxide (carbonate) concentrations. Methods to remove fission products, e.g., Cs, Sr, Ba and Zr, from alkaline solutions will be evaluated based upon their decontamination factors. We will discuss the feasibility of using chemically oxidizing alkaline solutions as an alternative spent nuclear fuel reprocessing method based on results from experimental quantitative investigations. (authors)

  3. Plate-Based Fuel Processing System Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Carlos Faz; Helen Liu; Jacques Nicole; David Yee

    2005-12-22

    On-board reforming of liquid fuels into hydrogen is an enabling technology that could accelerate consumer usage of fuel cell powered vehicles. The technology would leverage the convenience of the existing gasoline fueling infrastructure while taking advantage of the fuel cell efficiency and low emissions. Commercial acceptance of on-board reforming faces several obstacles that include: (1) startup time, (2) transient response, and (3) system complexity (size, weight and cost). These obstacles are being addressed in a variety of projects through development, integration and optimization of existing fuel processing system designs. In this project, CESI investigated steam reforming (SR), water-gas-shift (WGS) and preferential oxidation (PrOx) catalysts while developing plate reactor designs and hardware where the catalytic function is integrated into a primary surface heat exchanger. The plate reactor approach has several advantages. The separation of the reforming and combustion streams permits the reforming reaction to be conducted at a higher pressure than the combustion reaction, thereby avoiding costly gas compression for combustion. The separation of the two streams also prevents the dilution of the reformate stream by the combustion air. The advantages of the plate reactor are not limited to steam reforming applications. In a WGS or PrOx reaction, the non-catalytic side of the plate would act as a heat exchanger to remove the heat generated by the exothermic WGS or PrOx reactions. This would maintain the catalyst under nearly isothermal conditions whereby the catalyst would operate at its optimal temperature. Furthermore, the plate design approach results in a low pressure drop, rapid transient capable and attrition-resistant reactor. These qualities are valued in any application, be it on-board or stationary fuel processing, since they reduce parasitic losses, increase over-all system efficiency and help perpetuate catalyst durability. In this program, CESI

  4. Plate-Based Fuel Processing System Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Carlos Faz; Helen Liu; Jacques Nicole; David Yee

    2005-12-22

    On-board reforming of liquid fuels into hydrogen is an enabling technology that could accelerate consumer usage of fuel cell powered vehicles. The technology would leverage the convenience of the existing gasoline fueling infrastructure while taking advantage of the fuel cell efficiency and low emissions. Commercial acceptance of on-board reforming faces several obstacles that include: (1) startup time, (2) transient response, and (3) system complexity (size, weight and cost). These obstacles are being addressed in a variety of projects through development, integration and optimization of existing fuel processing system designs. In this project, CESI investigated steam reforming (SR), water-gas-shift (WGS) and preferential oxidation (PrOx) catalysts while developing plate reactor designs and hardware where the catalytic function is integrated into a primary surface heat exchanger. The plate reactor approach has several advantages. The separation of the reforming and combustion streams permits the reforming reaction to be conducted at a higher pressure than the combustion reaction, thereby avoiding costly gas compression for combustion. The separation of the two streams also prevents the dilution of the reformate stream by the combustion air. The advantages of the plate reactor are not limited to steam reforming applications. In a WGS or PrOx reaction, the non-catalytic side of the plate would act as a heat exchanger to remove the heat generated by the exothermic WGS or PrOx reactions. This would maintain the catalyst under nearly isothermal conditions whereby the catalyst would operate at its optimal temperature. Furthermore, the plate design approach results in a low pressure drop, rapid transient capable and attrition-resistant reactor. These qualities are valued in any application, be it on-board or stationary fuel processing, since they reduce parasitic losses, increase over-all system efficiency and help perpetuate catalyst durability. In this program, CESI

  5. Conversion of microalgae to jet fuel: process design and simulation.

    Science.gov (United States)

    Wang, Hui-Yuan; Bluck, David; Van Wie, Bernard J

    2014-09-01

    Microalgae's aquatic, non-edible, highly genetically modifiable nature and fast growth rate are considered ideal for biomass conversion to liquid fuels providing promise for future shortages in fossil fuels and for reducing greenhouse gas and pollutant emissions from combustion. We demonstrate adaptability of PRO/II software by simulating a microalgae photo-bio-reactor and thermolysis with fixed conversion isothermal reactors adding a heat exchanger for thermolysis. We model a cooling tower and gas floatation with zero-duty flash drums adding solids removal for floatation. Properties data are from PRO/II's thermodynamic data manager. Hydrotreating is analyzed within PRO/II's case study option, made subject to Jet B fuel constraints, and we determine an optimal 6.8% bioleum bypass ratio, 230°C hydrotreater temperature, and 20:1 bottoms to overhead distillation ratio. Process economic feasibility occurs if cheap CO2, H2O and nutrient resources are available, along with solar energy and energy from byproduct combustion, and hydrotreater H2 from product reforming.

  6. Femtosecond laser processing of fuel injectors - a materials processing evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Stuart, B C; Wynne, A

    2000-12-16

    Lawrence Livermore National Laboratory (LLNL) has developed a new laser-based machining technology that utilizes ultrashort-pulse (0.1-1.0 picosecond) lasers to cut materials with negligible generation of heat or shock. The ultrashort pulse laser, developed for the Department of Energy (Defense Programs) has numerous applications in operations requiring high precision machining. Due to the extremely short duration of the laser pulse, material removal occurs by a different physical mechanism than in conventional machining. As a result, any material (e.g., hardened steel, ceramics, diamond, silicon, etc.) can be machined with minimal heat-affected zone or damage to the remaining material. As a result of the threshold nature of the process, shaped holes, cuts, and textures can be achieved with simple beam shaping. Conventional laser tools used for cutting or high-precision machining (e.g., sculpting, drilling) use long laser pulses (10{sup -8} to over 1 sec) to remove material by heating it to the melting or boiling point (Figure 1.1a). This often results in significant damage to the remaining material and produces considerable slag (Figure 1.2a). With ultrashort laser pulses, material is removed by ionizing the material (Figure 1.1b). The ionized plasma expands away from the surface too quickly for significant energy transfer to the remaining material. This distinct mechanism produces extremely precise and clean-edged holes without melting or degrading the remaining material (Figures 1.2 and 1.3). Since only a very small amount of material ({approx} <0.5 microns) is removed per laser pulse, extremely precise machining can be achieved. High machining speed is achieved by operating the lasers at repetition rates up to 10,000 pulses per second. As a diagnostic, the character of the short-pulse laser produced plasma enables determination of the material being machined between pulses. This feature allows the machining of multilayer materials, metal on metal or metal on

  7. Selective CO Methanation Catalysts for Fuel Processing Applications

    Energy Technology Data Exchange (ETDEWEB)

    Dagle, Robert A.; Wang, Yong; Xia, Guanguang G.; Strohm, James J.; Holladay, Jamie D.; Palo, Daniel R.

    2007-07-15

    Abstract Selective CO methanation as a strategy for CO removal in micro fuel processing applications was investigated over Ru-based catalysts. Ru loading, pretreatment and reduction conditions, and choice of support were shown to affect catalyst activity, selectivity, and stability. Even operating at a gas-hourly-space-velocity as high as 13,500 hr-1, a 3%Ru/Al2O3 catalyst was able to lower CO in a reformate to less than 100 ppm over a wide temperature range from 240oC to 285 oC, while keeping hydrogen consumption below 10%.

  8. Phase equilibria of continuous fossil fuel process oils

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, M.J.; Weil, S.A. (Institute of Gas Technology, Chicago, IL (US))

    1988-04-01

    Fossil fuel process oils consist of such a large number of components that their only proper description is in terms of continuous distribution functions of a suitable characteristic variable. A methodology is presented to describe the oils in terms of a generalized distribution function. The characteristic variable is determined from measurements of the equilibrium ratios of two test oils, at ambient pressure. Application of the proposed methodology to a sequence of operations shows that, unlike the pseudocomponents technique, the level of accuracy can be maintained.

  9. Phase equilibria of continuous fossil fuel process oils

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, M.J.; Weil, S.A.

    1987-01-01

    Fossil fuel process oils consist of such a large number of components that their only proper description is in terms of continuous distribution functions of a suitable characteristic variable. A methodology is presented here to describe the oils in terms of a generalized distribution function. The characteristic variable is determined from measurements of the equilibrium ratios of two test oils, at ambient pressure. Application of the proposed methodology to a sequence of operations shows that, unlike the pseudocomponents technique, the level of accuracy can be maintained. 22 refs., 10 figs., 4 tabs.

  10. Phase equilibria of continuous fossil fuel process oils

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, M.J.

    1987-01-01

    Fossil fuel processes oils consist of such a large number of components that their only proper description is in terms of continuous distribution functions of a suitable characteristic variable. A methodology is presented here to describe the oils in terms of a generalized distribution function. The characteristic variable is determined from measurements of the equilibrium ratios of two test oils, at ambient pressure. Application of the proposed methodology to a sequence of operations shows that, unlike the psuedocomponents technique, the level of accuracy can be maintained.

  11. The Himalaya-Bengal Fan source to sink system - new insights by correlation of re-processed seismic data and IODP Expedition 354 results

    Science.gov (United States)

    Bergmann, Fenna; Schwenk, Tilmann; Spiess, Volkard; France-Lanord, Christian

    2016-04-01

    connect the sites of the drilling transect by means of seismo-stratigraphic analysis a large seismo-acoustic dataset gathered during cruises SO93 (1994), SO125/126 (1997) and SO188 (2006), all carried out in cooperation between the University of Bremen and the BGR, Hannover, is available. The dataset contains multichannel seismic data acquired with differ-ent seismic sources (GI-Gun/Watergun) to achieve differing subbottom penetration/resolution ratios. Although most of the pre-site survey data were already processed, major improve-ment could be gained by thoroughly (re) processing using new processing techniques and software developments. First processing results show significantly enhanced S/N ratio, reso-lution and reflector coherency. Full processing of the Watergun data was conducted for the first time. This high vertical resolution data has so far never been investigated and comple-ments the database, especially for a more detailed study of the upper few hundred meters of Bengal Fan deposits. First examinations of the watergun data in combination with drilling results proved them to be beneficial for the crucial borehole - seismic correlation and the investigations of the internal levee architecture, especially for the latest active channel-levee system.

  12. Thermochemical Process Development Unit: Researching Fuels from Biomass, Bioenergy Technologies (Fact Sheet)

    Energy Technology Data Exchange (ETDEWEB)

    2009-01-01

    The Thermochemical Process Development Unit (TCPDU) at the National Renewable Energy Laboratory (NREL) is a unique facility dedicated to researching thermochemical processes to produce fuels from biomass.

  13. Microbial fuel cell treatment of ethanol fermentation process water

    Science.gov (United States)

    Borole, Abhijeet P [Knoxville, TN

    2012-06-05

    The present invention relates to a method for removing inhibitor compounds from a cellulosic biomass-to-ethanol process which includes a pretreatment step of raw cellulosic biomass material and the production of fermentation process water after production and removal of ethanol from a fermentation step, the method comprising contacting said fermentation process water with an anode of a microbial fuel cell, said anode containing microbes thereon which oxidatively degrade one or more of said inhibitor compounds while producing electrical energy or hydrogen from said oxidative degradation, and wherein said anode is in electrical communication with a cathode, and a porous material (such as a porous or cation-permeable membrane) separates said anode and cathode.

  14. Process fuel equivalent (PFE)-a new parameter for process and energy audit

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharyyu, S.

    1983-04-01

    Process fuel equivalent (PFE) is an energy audit parameter which takes care of all the factors involved in the manufacturing process. In this parameter, all energy items--direct or indirect including those of purchased reagents, fluxes and other raw materials--are converted into a single variable fuel equivalent, and indicated the same in appropriate energy units. PFE value is an inverse order of efficiency--a low value corresponds to high efficiency. In this article, the PFE method has been used for production of copper from mine to wire bar.

  15. Maximizing the liquid fuel yield in a biorefining process.

    Science.gov (United States)

    Zhang, Bo; von Keitz, Marc; Valentas, Kenneth

    2008-12-01

    Biorefining strives to recover the maximum value from each fraction, at minimum energy cost. In order to seek an unbiased and thorough assessment of the alleged opportunity offered by biomass fuels, the direct conversion of various lignocellulosic biomass was studied: aspen pulp wood (Populus tremuloides), aspen wood pretreated with dilute acid, aspen lignin, aspen logging residues, corn stalk, corn spathe, corn cob, corn stover, corn stover pellet, corn stover pretreated with dilute acid, and lignin extracted from corn stover. Besides the heating rate, the yield of liquid products was found to be dependent on the final liquefaction temperature and the length of liquefaction time. The major compounds of the liquid products from various origins were identified by GC-MS. The lignin was found to be a good candidate for the liquefaction process, and biomass fractionation was necessary to maximize the yield of the liquid bio-fuel. The results suggest a biorefinery process accompanying pretreatment, fermentation to ethanol, liquefaction to bio-crude oil, and other thermo-conversion technologies, such as gasification. Other biorefinery options, including supercritical water gasification and the effectual utilization of the bio-crude oil, are also addressed.

  16. Automatic inspection for remotely manufactured fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Reifman, J.; Vitela, J.E. [Argonne National Lab., IL (United States); Gibbs, K.S.; Benedict, R.W. [Argonne National Lab., Idaho Falls, ID (United States)

    1995-06-01

    Two classification techniques, standard control charts and artificial neural networks, are studied as a means for automating the visual inspection of the welding of end plugs onto the top of remotely manufactured reprocessed nuclear fuel element jackets. Classificatory data are obtained through measurements performed on pre- and post-weld images captured with a remote camera and processed by an off-the-shelf vision system. The two classification methods are applied in the classification of 167 dummy stainless steel (HT9) fuel jackets yielding comparable results.

  17. Process engineering of ceramic composite coatings for fuel cell systems

    Energy Technology Data Exchange (ETDEWEB)

    Li, G.; Kim, H.; Chen, M.; Yang, Q.; Troczynski, T. [British Columbia Univ., Vancouver, BC (Canada). Dept. of Metals and Materials Engineering

    2003-07-01

    Researchers at UBCeram at the Department of Metals and Materials Engineering at the University of British Columbia have developed a technology to chemically bond composite sol-gel (CB-CSG) coating onto metallic surfaces of complex or concave shapes. The process has been optimized for electrically resistive coatings and corrosion-resistant coatings. The CSG is sprayed onto metallic surfaces and is heat-treated at 300 degrees C to partially dehydrate the hydroxides. The CSG film is then chemically bonded through reaction of active alumina with metal phosphates, such as aluminium phosphate. A new chromate-free process is being developed to address the issue of coatings porosity. The electrodeposition technique involves polymer particles mixed with suspended fine alumina particles which are co-deposited by electrophoretic means or by electrocoagulation. The composite e-coatings have excellent mechanical properties and are being considered as a protective coating for various components of fuel cell systems. 9 refs., 7 figs.

  18. Integrated microchemical systems for fuel processing in micro fuel cell applications

    Science.gov (United States)

    Pattekar, Ashish V.

    Rapid advances in microelectronics technology over the last decade have led to the search for novel applications of miniaturization to all aspects of engineering. Microreaction engineering, which involves the development of miniature reactors on microchips for novel applications, has been a key area of interest in this quest for miniaturization. The idea of a fully integrated microplant with embedded control electronics, sensors and actuators on a single silicon chip has been gaining increasing acceptance as significant progress is being made in this area. The aim of this project has been to demonstrate a working microreaction system for hydrogen delivery to miniature proton exchange membrane (PEM) fuel cells through the catalytic steam reforming of methanol. The complete reformer - fuel cell unit is proposed as an alternative to conventional portable sources of electricity such as batteries due to its ability to provide an uninterrupted supply of electricity as long as a supply of methanol and water can be provided. This technology also offers significantly higher energy storage densities, which translates into less frequent 'recharging' through the refilling of methanol fuel. Various aspects of the design of a miniature methanol reformer on a silicon substrate are discussed with a focus on the theoretical understanding of microreactor operation and optimum utilization of the semiconductor-processing techniques used for fabricating the devices. Three prototype microreactor designs have been successfully fabricated and tested. Issues related to microchannel capping, on-chip heating and temperature sensing, introduction and trapping of catalyst particles in microchannels, microfluidic interfacing, pressure drop reduction, and thermal insulation have been addressed. Details regarding modeling and simulation of the designs to provide an insight into the working of the microreactor are presented along with a description of the microfabrication steps followed to

  19. Electrochemical processing of spent nuclear fuels: An overview of oxide reduction in pyroprocessing technology

    Directory of Open Access Journals (Sweden)

    Eun-Young Choi

    2015-12-01

    Full Text Available The electrochemical reduction process has been used to reduce spent oxide fuel to a metallic form using pyroprocessing technology for a closed fuel cycle in combination with a metal-fuel fast reactor. In the electrochemical reduction process, oxides fuels are loaded at the cathode basket in molten Li2O–LiCl salt and electrochemically reduced to the metal form. Various approaches based on thermodynamic calculations and experimental studies have been used to understand the electrode reaction and efficiently treat spent fuels. The factors that affect the speed of the electrochemical reduction have been determined to optimize the process and scale-up the electrolysis cell. In addition, demonstrations of the integrated series of processes (electrorefining and salt distillation with the electrochemical reduction have been conducted to realize the oxide fuel cycle. This overview provides insight into the current status of and issues related to the electrochemical processing of spent nuclear fuels.

  20. Development of OTM Syngas Process and Testing of Syngas Derived Ultra-clean Fuels in Diesel Engines and Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    E.T. (Skip) Robinson; James P. Meagher; Prasad Apte; Xingun Gui; Tytus R. Bulicz; Siv Aasland; Charles Besecker; Jack Chen Bart A. van Hassel; Olga Polevaya; Rafey Khan; Piyush Pilaniwalla

    2002-12-31

    This topical report summarizes work accomplished for the Program from November 1, 2001 to December 31, 2002 in the following task areas: Task 1: Materials Development; Task 2: Composite Development; Task 4: Reactor Design and Process Optimization; Task 8: Fuels and Engine Testing; 8.1 International Diesel Engine Program; 8.2 Nuvera Fuel Cell Program; and Task 10: Program Management. Major progress has been made towards developing high temperature, high performance, robust, oxygen transport elements. In addition, a novel reactor design has been proposed that co-produces hydrogen, lowers cost and improves system operability. Fuel and engine testing is progressing well, but was delayed somewhat due to the hiatus in program funding in 2002. The Nuvera fuel cell portion of the program was completed on schedule and delivered promising results regarding low emission fuels for transportation fuel cells. The evaluation of ultra-clean diesel fuels continues in single cylinder (SCTE) and multiple cylinder (MCTE) test rigs at International Truck and Engine. FT diesel and a BP oxygenate showed significant emissions reductions in comparison to baseline petroleum diesel fuels. Overall through the end of 2002 the program remains under budget, but behind schedule in some areas.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hohimer, J.P.

    1978-09-01

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

  2. Preliminary safety analysis for offgas treatment system of DUPIC fuel manufacturing process at DFDF

    Energy Technology Data Exchange (ETDEWEB)

    Shin, J. M.; Lee, H. H.; Park, J. J.; Yang, M. S

    2000-09-01

    DUPIC fuel fabrication process is a dry processing technology to manufacture CANDU compatible fuel through a direct refabrication process from spent PWR fuel. DUPIC fuel fabrication process consists of the slitting of the spent PWR fuel rods, OREOX processing, homogeneous mixing, pelletizing and sintering. All these processes should be conducted by remote means in a M6 hot cell at IMEF. Since there is a lot of highly radioactive spent fuel(200 kg) to be used in DUPIC fuel fabrication process, safety analysis on DFDF facility is very important to improve the safety of hot cell and to reduce the dose exposure to operator. This report describes the design of IMEF facility, manufacturing equipment and process, offgas treatment system necessary for DUPIC fuel manufacturing process. Also, it provides the flow chart of arising and activity for each nuclide in offgas treatment system and final arising and activity for gaseous waste discharged from offgas treatment equipment into inside of M6 cell during OREOX and sintering processes in DUPIC fuel manufacturing process.

  3. Supercritical Fluids Processing of Biomass to Chemicals and Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Norman K. [Iowa State Univ., Ames, IA (United States)

    2011-09-28

    The main objective of this project is to develop and/or enhance cost-effective methodologies for converting biomass into a wide variety of chemicals, fuels, and products using supercritical fluids. Supercritical fluids will be used both to perform reactions of biomass to chemicals and products as well as to perform extractions/separations of bio-based chemicals from non-homogeneous mixtures. This work supports the Biomass Program’s Thermochemical Platform Goals. Supercritical fluids are a thermochemical approach to processing biomass that, while aligned with the Biomass Program’s interests in gasification and pyrolysis, offer the potential for more precise and controllable reactions. Indeed, the literature with respect to the use of water as a supercritical fluid frequently refers to “supercritical water gasification” or “supercritical water pyrolysis.”

  4. Selective CO methanation catalysts for fuel processing applications

    Energy Technology Data Exchange (ETDEWEB)

    Dagle, Robert A.; Wang, Yong; Xia, Guan-Guang; Strohm, James J.; Holladay, Jamelyn [Pacific Northwest National Laboratory, 902 Battle Boulevard, P.O. Box 999, Richland, WA 99352 (United States); Palo, Daniel R. [Pacific Northwest National Laboratory, 902 Battle Boulevard, P.O. Box 999, Richland, WA 99352 (United States); Microproducts Breakthrough Institute, P.O. Box 2330, Corvallis, OR 97339 (United States)

    2007-07-15

    Selective CO methanation as a strategy for CO removal in fuel processing applications was investigated over Ru-based catalysts. Ru metal loading and crystallite size were shown to affect catalyst activity and selectivity. Even operating at a gas-hourly-space-velocity as high as 13,500 h{sup -1}, a 3% Ru/Al{sub 2}O{sub 3} catalyst with a 34.2 nm crystallite was shown to be capable of reducing CO in a reformate to less than 100 ppm over a wide temperature range from 240 to 280 C, while keeping hydrogen consumption below 10%. We present the effects of metal loading, preparation method, and crystallite size on performance for Ru-based catalysts in the selective methanation of CO in the presence of H{sub 2} and CO{sub 2}. (author)

  5. Environmental assessment for radioisotope heat source fuel processing and fabrication

    Energy Technology Data Exchange (ETDEWEB)

    1991-07-01

    DOE has prepared an Environmental Assessment (EA) for radioisotope heat source fuel processing and fabrication involving existing facilities at the Savannah River Site (SRS) near Aiken, South Carolina and the Los Alamos National Laboratory (LANL) near Los Alamos, New Mexico. The proposed action is needed to provide Radioisotope Thermoelectric Generators (RTG) to support the National Aeronautics and Space Administration's (NASA) CRAF and Cassini Missions. Based on the analysis in the EA, DOE has determined that the proposed action does not constitute a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, an Environmental Impact Statement is not required. 30 refs., 5 figs.

  6. Process for dehydration of oregano using propane gas as fuel

    Directory of Open Access Journals (Sweden)

    Carlos O. Velásquez-Santos

    2014-08-01

    Full Text Available The article describes two important issues, the first is the process to design, implement and validate a mechanical dryer of oregano, using propane gas as fuel, and the second is the cost of the process of dehydrated, taking into account the cost of electric energy consumption by the fan and the cost of propane gas consumption by the heat exchanger. To achieve this, it was necessary review the state of the art and the study of the raw material (oregano, were established as premises of design the necessary technical specifications and the variables involved in the process, using conceptual methods and simulation to ensure that it complies with the ISO standard 7925:1999, which defines the requirements for the marketing of dried oregano and processed. Emphasis was made on the percentage of moisture that is 10%, the moisture of the product was found by the azeotropic distillation method, subsequently was validated the functionality and efficiency, comparing the results from an experimental design, then it was obtained the drying curve of oregano with the prototype of drying and it was checked if it meets ISO 7925:1999 standard and the NTC 4423 standard in order to obtain a final product dehydrated with the percentage of humidity appropriate.

  7. Refuse-derived fuels: Provision and processing; Schnittstelle und Aufbereitungstiefe von Ersatzbrennstoffen fuer die energetische Verwertung

    Energy Technology Data Exchange (ETDEWEB)

    Beckmann, M.; Horeni, M. [Bauhaus Univ. Weimar (Germany). Professur Verfahren und Umwelt; Scholz, R. [Technische Univ. Clausthal (Germany). Inst. fuer Energieverfahrenstechnik und Brennstofftechnik

    2005-07-01

    Refuse-derived fuels are produced by mechanical or mechanical and biological treatment of waste materials with the intention of providing substitute fuels for the chemical and power plant industry. To do this, the properties of refues-derived fuels must be known as well as the process modifications that may be required. Further, energy and pollutant balances (including CO2) must be established in order to find out about the economic efficiency of the substitute fuels. There is still need for research concerning the methodology of characterisation and the classification of substitute fuels. The contribution starts by presenting a classification of substitute fuels, their applications and potentials. This is followed by a description of the characteristics of substitute fuels and the process optimisations required. Using simplified model assumptions, the results are then discussed with regard to the influence of the depth of processing. Finally, development tasks still required are summarized. (orig.)

  8. Calibration of burnup monitor in the Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oheda, K.; Naito, H.; Hirota, M. [Japan Nuclear Fuel Ltd., Aomori (Japan); Natsume, K. [Toshiba Corp., Yokohama, Kawasaki, Kanagawa (Japan); Kumanomido, H. [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    1998-07-01

    The Rokkasho Reprocessing Plant has adopted a credit for burnup in criticality control in the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. The burnup monitor system, prepared for BWR and PWR type fuel assemblies, nondestructively measures the burnup value and determines the residual U-235 enrichment in a spent fuel assembly, and criticality is controlled by the value of residual U-235 enrichment in SFSF and by the value of top 50 cm average burnup in the Dissolution Facility. The burnup monitor consists of three measurement systems; a Boss gamma-ray profile measurement system, a high resolution gamma-ray spectrometry system, and a passive neutron measurement system. The monitor sensitivity is calibrated against operator-declared burnup values through repetitive measurements of 100 spent fuel assemblies: BWR 8 X 8, PWR 14 X 14. and 17 X 17. The outline of the measurement methods, objectives of the calibration, actual calibration method, and an example of calibration performed in a demonstration experiment are presented. (author)

  9. A Two-Dimensional, Finite-Difference Model of the Oxidation of a Uranium Carbide Fuel Pellet

    OpenAIRE

    Shepherd, J; Fairweather, M; Hanson, BC; Heggs, PJ

    2015-01-01

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

  10. Process modeling of fuel cell vehicle power system

    Institute of Scientific and Technical Information of China (English)

    CHEN LiMing; LIN ZhaoJia; MA ZiFeng

    2009-01-01

    Constructed here is a mathematic model of PEM Fuel Cell Vehicle Power System which is composed of fuel supply model, fuel cell stack model and water-heat management model. The model was developed by Matiab/Simulink to evaluate how the major operating variables affect the output performances. Itshows that the constructed model can represent characteristics of the power system closely by comparing modeling results with experimental data, and it can be used in the study and design of fuel cell vehicle power system.

  11. Fault Tree Analysis for Red Oil Explosion in Reprocessing Facility%后处理设施的红油爆炸故障树分析

    Institute of Scientific and Technical Information of China (English)

    王任泽; 王学新; 庄大杰; 曹芳芳

    2013-01-01

    Almost all spent fuel reprocessing facilities have adopted Purex process .T he red oil explosion is a great concern in safety study of spent fuel reprocessing facilities adopting Purex process .The event tree and fault tree analysis was performed for the red oil explosion of a medium level radioactive waste liquid evaporator for the collective decontamination and separation cycle segment in a representative reprocessing facility in this paper . The results show that the occurrence frequency of a red oil explosion is extremely low ,and human errors and common cause failures are major causes to a red oil explosion . Therefore , some relevant measures should be taken to prevent such accidents .%目前世界上几乎所有的乏燃料后处理设施均选用 Purex流程。红油爆炸事故是 Purex流程后处理设施安全研究的焦点问题之一。本文对典型后处理设施共去污分离循环工段的中放废液蒸发器的红油爆炸事故进行了事件树-故障树分析。结果显示,红油爆炸事故的发生频率极低;人因失误和共因失效对整个红油爆炸事故的贡献很大。应采取相应措施,以预防事故的发生。

  12. Fuel Quality/Processing Study. Volume II. Appendix, Task I, literature survey

    Energy Technology Data Exchange (ETDEWEB)

    O' Hara, J B; Bela, A; Jentz, N E; Klumpe, H W; Kessler, R E; Kotzot, H T; Loran, B I

    1981-04-01

    This activity was begun with the assembly of information from Parsons' files and from contacts in the development and commercial fields. A further more extensive literature search was carried out using the Energy Data Base and the American Petroleum Institute Data Base. These are part of the DOE/RECON system. Approximately 6000 references and abstracts were obtained from the EDB search. These were reviewed and the especially pertinent documents, approximately 300, were acquired in the form of paper copy or microfiche. A Fuel Properties form was developed for listing information pertinent to gas turbine liquid fuel properties specifications. Fuel properties data for liquid fuels from selected synfuel processes, deemed to be successful candidates for near future commercial plants were tabulated on the forms. The processes selected consisted of H-Coal, SRC-II and Exxon Donor Solvent (EDS) coal liquefaction processes plus Paraho and Tosco shale oil processes. Fuel properties analyses for crude and distillate syncrude process products are contained in Section 2. Analyses representing synthetic fuels given refinery treatments, mostly bench scale hydrotreating, are contained in Section 3. Section 4 discusses gas turbine fuel specifications based on petroleum source fuels as developed by the major gas turbine manufacturers. Section 5 presents the on-site gas turbine fuel treatments applicable to petroleum base fuels impurities content in order to prevent adverse contaminant effects. Section 7 relates the environmental aspects of gas turbine fuel usage and combustion performance. It appears that the near future stationary industrial gas turbine fuel market will require that some of the synthetic fuels be refined to the point that they resemble petroleum based fuels.

  13. Fuel Quality/Processing Study. Volume II. Appendix, Task I, literature survey

    Energy Technology Data Exchange (ETDEWEB)

    O' Hara, J B; Bela, A; Jentz, N E; Klumpe, H W; Kessler, R E; Kotzot, H T; Loran, B I

    1981-04-01

    This activity was begun with the assembly of information from Parsons' files and from contacts in the development and commercial fields. A further more extensive literature search was carried out using the Energy Data Base and the American Petroleum Institute Data Base. These are part of the DOE/RECON system. Approximately 6000 references and abstracts were obtained from the EDB search. These were reviewed and the especially pertinent documents, approximately 300, were acquired in the form of paper copy or microfiche. A Fuel Properties form was developed for listing information pertinent to gas turbine liquid fuel properties specifications. Fuel properties data for liquid fuels from selected synfuel processes, deemed to be successful candidates for near future commercial plants were tabulated on the forms. The processes selected consisted of H-Coal, SRC-II and Exxon Donor Solvent (EDS) coal liquefaction processes plus Paraho and Tosco shale oil processes. Fuel properties analyses for crude and distillate syncrude process products are contained in Section 2. Analyses representing synthetic fuels given refinery treatments, mostly bench scale hydrotreating, are contained in Section 3. Section 4 discusses gas turbine fuel specifications based on petroleum source fuels as developed by the major gas turbine manufacturers. Section 5 presents the on-site gas turbine fuel treatments applicable to petroleum base fuels impurities content in order to prevent adverse contaminant effects. Section 7 relates the environmental aspects of gas turbine fuel usage and combustion performance. It appears that the near future stationary industrial gas turbine fuel market will require that some of the synthetic fuels be refined to the point that they resemble petroleum based fuels.

  14. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  15. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  16. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  17. Iodine Pathways and Off-Gas Stream Characteristics for Aqueous Reprocessing Plants – A Literature Survey and Assessment

    Energy Technology Data Exchange (ETDEWEB)

    R. T. Jubin; D. M. Strachan; N. R. Soelberg

    2013-09-01

    Used nuclear fuel is currently being reprocessed in only a few countries, notably France, England, Japan, and Russia. The need to control emissions of the gaseous radionuclides to the air during nuclear fuel reprocessing has already been reported for the entire plant. But since the gaseous radionuclides can partition to various different reprocessing off-gas streams, for example, from the head end, dissolver, vessel, cell, and melter, an understanding of each of these streams is critical. These off-gas streams have different flow rates and compositions and could have different gaseous radionuclide control requirements, depending on how the gaseous radionuclides partition. This report reviews the available literature to summarize specific engineering data on the flow rates, forms of the volatile radionuclides in off-gas streams, distributions of these radionuclides in these streams, and temperatures of these streams. This document contains an extensive bibliography of the information contained in the open literature.

  18. System Design Description and Requirements for Modeling the Off-Gas Systems for Fuel Recycling Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Daryl R. Haefner; Jack D. Law; Troy J. Tranter

    2010-08-01

    This document provides descriptions of the off-gases evolved during spent nuclear fuel processing and the systems used to capture the gases of concern. Two reprocessing techniques are discussed, namely aqueous separations and electrochemical (pyrochemical) processing. The unit operations associated with each process are described in enough detail so that computer models to mimic their behavior can be developed. The document also lists the general requirements for the desired computer models.

  19. Fate of virginiamycin through the fuel ethanol production process.

    Science.gov (United States)

    Bischoff, Kenneth M; Zhang, Yanhong; Rich, Joseph O

    2016-05-01

    Antibiotics are frequently used to prevent and treat bacterial contamination of commercial fuel ethanol fermentations, but there is concern that antibiotic residues may persist in the distillers grains coproducts. A study to evaluate the fate of virginiamycin during the ethanol production process was conducted in the pilot plant facilities at the National Corn to Ethanol Research Center, Edwardsville, IL. Three 15,000-liter fermentor runs were performed: one with no antibiotic (F1), one dosed with 2 parts per million (ppm) of a commercial virginiamycin product (F2), and one dosed at 20 ppm of virginiamycin product (F3). Fermentor samples, distillers dried grains with solubles (DDGS), and process intermediates (whole stillage, thin stillage, syrup, and wet cake) were collected from each run and analyzed for virginiamycin M and virginiamycin S using a liquid chromatography-mass spectrometry method. Virginiamycin M was detected in all process intermediates of the F3 run. On a dry-weight basis, virginiamycin M concentrations decreased approximately 97 %, from 41 μg/g in the fermentor to 1.4 μg/g in the DDGS. Using a disc plate bioassay, antibiotic activity was detected in DDGS from both the F2 and F3 runs, with values of 0.69 μg virginiamycin equivalent/g sample and 8.9 μg/g, respectively. No antibiotic activity (<0.6 μg/g) was detected in any of the F1 samples or in the fermentor and process intermediate samples from the F2 run. These results demonstrate that low concentrations of biologically active antibiotic may persist in distillers grains coproducts produced from fermentations treated with virginiamycin.

  20. Fossil-fuel process oils as continuous fluids

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian-Amin, M.J.

    1986-01-01

    The oils produced by fossil fuel conversion processes consist of such a large number of components that their only proper description is as continuous fluids (i.e., continuum of components). A methodology is presented here to describe the vapor liquid equilibrium processes involving continuous oils. It describes the oil in terms of one or more continuous distribution functions (fractional continuous oils) of some measurable quantity (i.e., characteristic variable) that, in the view of the equilibrium ratio relationship, maintain their functional form in equilibrium processes. Parameters of the distributions of the product streams in any equilibrium process (i.e., vapor and liquid) are determined in terms of the parameters of the feed stream and the operating condition (e.g., T,P). In general, the procedure can be applied to both ideal and non-ideal systems, but in view of the experimental results indicating ideality, only those systems were analyzed. An ambient pressure batch distillation system was constructed to collect vapor-liquid equilibrium data of continuous test oils. Two test oils, a shale oil and a coal oil were studied in this work. From measurement of the equilibrium ratios of the test oils it was determined that both oils behave ideally and the equilibrium ratio was independent of the liquid composition. A simple and definable function of the boiling point provided to be a suitable characteristic variable for the proposed methodology to the sequential operation has shown that if the functions are chosen properly, then the error incurred will not propagate at a significant rate and at the same level of accuracy can be maintained.

  1. Tecnored process - high potential in using different kinds of solid fuels

    Directory of Open Access Journals (Sweden)

    José Henrique Noldin Júnior

    2005-12-01

    Full Text Available One important feature of the Brazilian Tecnored ironmaking process is its flexibility to use different types of solid fuels, other than metallurgical coke, as proved in the pilot plant tests by extensively using green petroleum coke, biomasses, high ash cokes, etc. Even if new solid fuels not thus far used are envisaged for a given project, thru the bench scale simulator of the process it is possible to predict the behavior of such solid fuels in the Tecnored furnace and establish the best techno-economical-environmental equation for its use. This paper discusses the key aspects involved in the use of alternative solid fuels in the Tecnored process.

  2. Plutonium production story at the Hanford site: processes and facilities history

    Energy Technology Data Exchange (ETDEWEB)

    Gerber, M.S., Westinghouse Hanford

    1996-06-20

    This document tells the history of the actual plutonium production process at the Hanford Site. It contains five major sections: Fuel Fabrication Processes, Irradiation of Nuclear Fuel, Spent Fuel Handling, Radiochemical Reprocessing of Irradiated Fuel, and Plutonium Finishing Operations. Within each section the story of the earliest operations is told, along with changes over time until the end of operations. Chemical and physical processes are described, along with the facilities where these processes were carried out. This document is a processes and facilities history. It does not deal with the waste products of plutonium production.

  3. Toward mechanistic understanding of nuclear reprocessing chemistries by quantifying lanthanide solvent extraction kinetics via microfluidics with constant interfacial area and rapid mixing.

    Science.gov (United States)

    Nichols, Kevin P; Pompano, Rebecca R; Li, Liang; Gelis, Artem V; Ismagilov, Rustem F

    2011-10-05

    The closing of the nuclear fuel cycle is an unsolved problem of great importance. Separating radionuclides produced in a nuclear reactor is useful both for the storage of nuclear waste and for recycling of nuclear fuel. These separations can be performed by designing appropriate chelation chemistries and liquid-liquid extraction schemes, such as in the TALSPEAK process (Trivalent Actinide-Lanthanide Separation by Phosphorus reagent Extraction from Aqueous Komplexes). However, there are no approved methods for the industrial scale reprocessing of civilian nuclear fuel in the United States. One bottleneck in the design of next-generation solvent extraction-based nuclear fuel reprocessing schemes is a lack of interfacial mass transfer rate constants obtained under well-controlled conditions for lanthanide and actinide ligand complexes; such rate constants are a prerequisite for mechanistic understanding of the extraction chemistries involved and are of great assistance in the design of new chemistries. In addition, rate constants obtained under conditions of known interfacial area have immediate, practical utility in models required for the scaling-up of laboratory-scale demonstrations to industrial-scale solutions. Existing experimental techniques for determining these rate constants suffer from two key drawbacks: either slow mixing or unknown interfacial area. The volume of waste produced by traditional methods is an additional, practical concern in experiments involving radioactive elements, both from disposal cost and experimenter safety standpoints. In this paper, we test a plug-based microfluidic system that uses flowing plugs (droplets) in microfluidic channels to determine absolute interfacial mass transfer rate constants under conditions of both rapid mixing and controlled interfacial area. We utilize this system to determine, for the first time, the rate constants for interfacial transfer of all lanthanides, minus promethium, plus yttrium, under TALSPEAK

  4. Carbon stripping - a critical process step in chemical looping combustion of solid fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kramp, M.; Thon, A.; Hartge, E.U.; Heinrich, S.; Werther, J. [Hamburg University of Technology, Institute of Solids Process Engineering and Particle Technology, Hamburg (Germany)

    2012-03-15

    In chemical looping combustion of solid fuels the well-mixed solids flow from the fuel reactor consisting of char, ash, and oxygen carrier particles cannot be completely separated into its constituents before it enters the air reactor. The slip of carbon will thus lead to char oxidation in the wrong reactor. Process simulation was applied to investigate the carbon stripping process in chemical looping combustion of solid fuels. Depending on the fuel choice, without carbon stripping CO{sub 2} capture rates below 50 % are calculated for 4 min of solids residence time in the fuel reactor. In a process with carbon stripper, however, CO{sub 2} capture rates exceeding 90 % can be achieved for both fuels investigated in this work. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  5. Possible toxic effects from the nuclear reprocessing at Sellafield and Cap de la Hague

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, M.; Coeytaux, X.; Faid, Y.B.; Marignac, Y.; Rouy, E. [Wise, 75 - Paris (France); Thompson, G. [IRSS, Cambridge (United States); Fairlie, I.; Lowry, D.; Sumner, D

    2001-11-15

    The principal aim of this report is to assist the Committee of Petitions of the European Parliament in its consideration of Petition 393/95 brought by Dr. W. Nachtwey. The Petition expresses concerns about radioactive discharges from nuclear reprocessing plants at Sellafield in the UK and La Hague in France, and their possible adverse health effects. Six years after the Petition was introduced, the Petitioner main concerns remain relevant. This report concludes that reprocessing discharges are a valid matter for the Committee consideration. It also concludes that, on balance, the Petitioner's concerns over radioactive discharges from Sellafield and La Hague are justified. The report presents evidence and data on: 1) radioactive discharges from the Sellafield and La Hague sites; 2) resulting nuclide concentrations in environmental media including foodstuffs; 3) radiation doses from nuclide discharges to critical groups near the sites; 4) adverse health effects near the two sites; and 5) resulting collective doses from nuclide discharges. The report also examines a number of current issues in radiobiology concerning health effects from exposure to ionising radiation, in particular genetic and in utero effects. In addition, in accordance with contract specifications, the report examines other major factors that might influence future decision-making on reprocessing. It provides information on the legal framework, the operational history of the plants and the economic case for reprocessing compared with available alternatives for spent nuclear fuel management. The report also makes policy-related recommendations that take into account current knowledge and uncertainties in risk assessment and the availability of alternatives to reprocessing in spent fuel management. (authors)

  6. Chemical Engineering Division Fuel Cycle Programs. Quarterly progress report, October--December 1977

    Energy Technology Data Exchange (ETDEWEB)

    Steindler, M. J.; Ader, M.; Barletta, R. E.

    1978-01-01

    Fuel cycle studies reported for this period include studies of advanced solvent extraction techniques focussed on the development of centrifugal contactors for use in Purex processes. Miniature single-stage and eight-stage centrifugal contactors are being employed in performance studies applicable to larger units. In other work, literature on the dispersion of reagents as a result of explosions is being reviewed to develop systematic data applicable to fuel reprocessing and useful in identifying source terms. In yet other work, scouting studies were performed to obtain criteria for identifying organic solutions suitable for the separation of actinides from fission products. A program has been initiated on pyrochemical and dry processing of nuclear fuel. Literature reviews have been initiated on material development, carbide fuel reprocessing, and thorium-uranium reprocessing in fused salts. A review and evaluation of the encapsulation of high-level waste in a metal matrix is under way. Corrosion and leach rates of simulated waste forms are being measured and a model has been proposed to describe the reaction between solidified high-level waste and metals. In other work, criteria for the handling of fuel assembly hulls are being developed on the basis of past work on the pyrophoricity of zirconium alloys and related criteria from several sources. Experimental work is underway to determine whether nuclear wastes can be safely confined in geologic formations. Information is being obtained on the migration of radionuclides in aqueous solution-rock systems. 17 figures, 27 tables.

  7. Development of nuclear fuel cycle technologies - bases of long-term provision of fuel and environmental safety of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Solonin, M.I.; Polyakov, A.S.; Zakharkin, B.S.; Smelov, V.S.; Nenarokomov, E.A.; Mukhin, I.V. [SSC, RF, A.A. Bochvar ALL-Russia Research Institute of Inorganic Materials, Moscow (Russian Federation)

    2000-07-01

    To-day nuclear power is one of the options, however, to-morrow it may become the main source of the energy, thus, providing for the stable economic development for the long time to come. The availability of the large-scale nuclear power in the foreseeable future is governed by not only the safe operation of nuclear power plants (NPP) but also by the environmentally safe management of spent nuclear fuel, radioactive waste conditioning and long-term storage. More emphasis is to be placed to the closing of the fuel cycle in view of substantial quantities of spent nuclear fuel arisings. The once-through fuel cycle that is cost effective at the moment cannot be considered to be environmentally safe even for the middle term since the substantial build-up of spent nuclear fuel containing thousands of tons Pu will require the resolution of the safe management problem in the nearest future and is absolutely unjustified in terms of moral ethics as a transfer of the responsibility to future generations. The minimization of radioactive waste arisings and its radioactivity is only feasible with the closed fuel cycle put into practice and some actinides and long-lived fission radionuclides burnt out. The key issues in providing the environmentally safe fuel cycle are efficient processes of producing fuel for NPP, radionuclide after-burning included, a long-term spent nuclear fuel storage and reprocessing as well as radioactive waste management. The paper deals with the problems inherent in producing fuel for NPP with a view for the closed fuel cycle. Also discussed are options of the fuel cycle, its effectiveness and environmental safety with improvements in technologies of spent nuclear fuel reprocessing and long-lived radionuclide partitioning. (authors)

  8. Heterogeneous catalytic process for alcohol fuels from syngas

    Energy Technology Data Exchange (ETDEWEB)

    Minahan, D.M.; Nagaki, D.A.

    1995-12-31

    This project is focused on the discovery and evaluation of novel heterogeneous catalyst for the production of oxygenated fuel enhancers from synthesis gas. Catalysts have been studied and optimized for the production of methanol and isobutanol mixtures which may be used for the downstream synthesis of MTBE or related oxygenates. Higher alcohols synthesis (HAS) from syngas was studied; the alcohols that are produced in this process may be used for the downstream synthesis of MTBE or related oxygenates. This work has resulted in the discovery of a catalyst system that is highly selective for isobutanol compared with the prior art. The catalysts operate at high temperature (400{degrees}C), and consist of a spinel oxide support (general formula AB{sub 2}O{sub 4}, where A=M{sup 2+} and B = M{sup 3+}), promoted with various other elements. These catalysts operate by what is believed to be an aldol condensation mechanism, giving a product mix of mainly methanol and isobutanol. In this study, the effect of product feed/recycle (methanol, ethanol. n-propanol, isopropanol, carbon dioxide and water) on the performance of 10-DAN-55 (spinel oxide based catalyst) at 400{degrees}C, 1000 psi, GHSV = 12,000 and syngas (H{sub 2}/CO) ratio = 1:2 (alcohol addition) and 1:1 (carbon dioxide and water addition) was studied. The effect of operation at high temperatures and pressures on the performance of an improved catalyst formulation was also examined.

  9. Control of radio-iodine at the German reprocessing plant WAK during operation and after shutdown

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, F.J.; Herrmann, B.; Kuhn, K.D. [Wiederaufarbeitungsanlage Karlsruhe (Germany)] [and others

    1997-08-01

    During 20 years of operation 207 metric tons of oxide fuel from nuclear power reactors with 19 kg of iodine-129 had been reprocessed in the WAK plant near Karlsruhe. In January 1991 the WAK Plant was shut down. During operation iodine releases of the plant as well as the iodine distribution over the liquid and gaseous process streams had been determined. Most of the iodine is evolved into the dissolver off-gas in volatile form. The remainder is dispersed over many aqueous, organic and especially gaseous process and waste streams. After shut down of the plant in January 1991, iodine measurements in the off-gas streams have been continued up to now. Whereas the iodine-129 concentration in the dissolver off-gas dropped during six months after shutdown by three orders of magnitude, the iodine concentrations in the vessel ventilation system of the PUREX process and the cell vent system decreased only by a factor of 10 during the same period. Iodine-129 releases of the liquid high active waste storage tanks did not decrease distinctly. The removal efficiencies of the silver impregnated iodine filters in the different off-gas streams of the WAK plant depend on the iodine concentration in the off-gas. The reason of the observed dependence of the DF on the iodine-129 concentration might be due to the presence of organic iodine compounds which are difficult to remove. 13 refs., 3 figs.

  10. Japanese perspectives and research on packaging, transport and storage of spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Saegusa, T.; Ito, C.; Yamakawa, H.; Shirai, K. [Central Research Inst. of Electric Power Industry (CRIEPI), Abiko (Japan)

    2004-07-01

    The Japanese policy on spent fuel is reprocessing. Until, reprocessed, spent fuel shall be stored properly. This paper overviews current status of transport and storage of spent fuel with related research in Japan. The research was partly carried out under a contract of Ministry of Economy, Trade and Industry of the Japanese government.

  11. 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.

  12. Properties of Aluminum Deposited by a High-Velocity Oxygen-Fueled Process

    Energy Technology Data Exchange (ETDEWEB)

    Chow, R; Decker, T A; Gansert, R V; Gansert, D; Lee, D

    2001-06-12

    Aluminum coatings deposited by a HVOF process have been demonstrated and relevant coating properties evaluated according to two deposition parameters, the spray distance and the oxygen-to-fuel flow ratio. The coating porosity, surface roughness, and microhardness are measured. The coating properties are fairly insensitive to spray distance, the distance between the nozzle and the workpiece, and fuel ratios, the oxygen-to-fuel flow. Increasing the fuel content does appear to improve the process productivity in terms of surface roughness. Minimization of nozzle loading is discussed.

  13. Rocket Fuel Synthesis by Fisher-Tropsch Process Project

    Data.gov (United States)

    National Aeronautics and Space Administration — While In-Situ Resource Utilization (ISRU) studies for Mars return have emphasized methane fuel, only modest work has been done to develop the methane-powered rocket...

  14. Molten salts processes and generic simulation

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Toru; Minato, Kazuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    Development of dry separation process (pyrochemical process) using molten salts for the application of spent-nuclear fuel reprocessing requires a rather complete fundamental database as well as process simulation technique with wide applicability. The present report concerns recent progress and problems in this field taking behaviors of co-electrodeposition of UO{sub 2} and PuO{sub 2} in molten salts as an example, and using analytical simulation of local equilibrium combined with generic diffusion. (S. Ohno)

  15. Technical and economic feasibility of alternative fuel use in process heaters and small boilers

    Energy Technology Data Exchange (ETDEWEB)

    1980-02-01

    The technical and economic feasibility of using alternate fuels - fuels other than oil and natural gas - in combustors not regulated by the Powerplant and Industrial Fuel Use Act of 1978 (FUA) was evaluated. FUA requires coal or alternate fuel use in most large new boilers and in some existing boilers. Section 747 of FUA authorizes a study of the potential for reduced oil and gas use in combustors not subject to the act: small industrial boilers with capacities less than 100 MMBtu/hr, and process heat applications. Alternative fuel use in combustors not regulated by FUA was examined and the impact of several measures to encourage the substitution of alternative fuels in these combustors was analyzed. The primary processes in which significant fuel savings can be achieved are identified. Since feedstock uses of oil and natural gas are considered raw materials, not fuels, feedstock applications are not examined in this analysis. The combustors evaluated in this study comprise approximately 45% of the fuel demand projected in 1990. These uses would account for more than 3.5 million barrels per day equivalent fuel demand in 1990.

  16. Optimizing photo-Fenton like process for the removal of diesel fuel from the aqueous phase

    OpenAIRE

    Dehghani, Mansooreh; Shahsavani, Esmaeel; Farzadkia, Mahdi; Samaei, Mohammad Reza

    2014-01-01

    Background In recent years, pollution of soil and groundwater caused by fuel leakage from old underground storage tanks, oil extraction process, refineries, fuel distribution terminals, improper disposal and also spills during transferring has been reported. Diesel fuel has created many problems for water resources. The main objectives of this research were focused on assessing the feasibility of using photo-Fenton like method using nano zero-valent iron (nZVI/UV/H2O2) in removing total petro...

  17. International safeguards for reprocessing plants. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Kratzer, M.; Scheinman, L.; Sievering, N.; Wonder, E.; Lipman, D.; Immerman, W.; Elliott, J.M.; Crane, F.

    1981-04-01

    Proliferation risks inherent in reprocessing show the need to employ technically effective safeguards which can detect, with a high degree of assurance and on a timely basis, the diversion of significant quantities of fissionable material. A balance must be struck between what is technically feasible and effective and what is institutionally acceptable. Purpose of this report is to examine the several technical approaches to safeguards in light of their prospective acceptability. This study defines the economic, political and institutional nature of the safeguards problem; surveys generically alternative technical approaches to international safeguards including their effectiveness and relative development; characterizes the institutional implications and uncertainties associated with the acceptance and implementation of each technical alternative; and integrates these assessments into a set of overall judgments on feasible directions for reprocessing plant safeguards systems.

  18. Modeling of the filling and cooling processes of hot fuel mains in Liquid Fuel Rocket Power Plant (LFRPP)

    Science.gov (United States)

    Prisnyakov, V. F.; Pokrishkin, V. V.; Serebryansky, V. N.

    A mathematical model of heat and mass exchange processes during filling and cooling of hot fuel mains of the Liquid Fuel Rocket Power Plant (LFRPP), which allows to define a mass consumption and distribution of two-phase flow parameters by the length of pipeline. Results of calculations are compared with experimental data, taken during filling of the main with a supply of liquid oxygen from the tank into the combustion chamber. Also, the results of modeling of hydrogen main dynamic characteristics of LFRPP in the same conditions are given.

  19. High-level disinfection of gastrointestinal endoscope reprocessing.

    Science.gov (United States)

    Chiu, King-Wah; Lu, Lung-Sheng; Chiou, Shue-Shian

    2015-02-20

    High level disinfection (HLD) of the gastrointestinal (GI) endoscope is not simply a slogan, but rather is a form of experimental monitoring-based medicine. By definition, GI endoscopy is a semicritical medical device. Hence, such medical devices require major quality assurance for disinfection. And because many of these items are temperature sensitive, low-temperature chemical methods, such as liquid chemical germicide, must be used rather than steam sterilization. In summarizing guidelines for infection prevention and control for GI endoscopy, there are three important steps that must be highlighted: manual washing, HLD with automated endoscope reprocessor, and drying. Strict adherence to current guidelines is required because compared to any other medical device, the GI endoscope is associated with more outbreaks linked to inadequate cleaning or disinfecting during HLD. Both experimental evaluation on the surveillance bacterial cultures and in-use clinical results have shown that, the monitoring of the stringent processes to prevent and control infection is an essential component of the broader strategy to ensure the delivery of safe endoscopy services, because endoscope reprocessing is a multistep procedure involving numerous factors that can interfere with its efficacy. Based on our years of experience in the surveillance of culture monitoring of endoscopic reprocessing, we aim in this study to carefully describe what details require attention in the GI endoscopy disinfection and to share our experience so that patients can be provided with high quality and safe medical practices. Quality management encompasses all aspects of pre- and post-procedural care including the efficiency of the endoscopy unit and reprocessing area, as well as the endoscopic procedure itself.

  20. Study on the fuel cycle cost of gas turbine high temperature reactor (GTHTR300). Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Takei, Masanobu; Katanishi, Shoji; Nakata, Tetsuo; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Oda, Takefumi; Izumiya, Toru [Nuclear Fuel Industries, Ltd., Tokyo (Japan)

    2002-11-01

    In the basic design of gas turbine high temperature reactor (GTHTR300), reduction of the fuel cycle cost has a large benefit of improving overall plant economy. Then, fuel cycle cost was evaluated for GTHTR300. First, of fuel fabrication for high-temperature gas cooled reactor, since there was no actual experience with a commercial scale, a preliminary design for a fuel fabrication plant with annual processing of 7.7 ton-U sufficient four GTHTR300 was performed, and fuel fabrication cost was evaluated. Second, fuel cycle cost was evaluated based on the equilibrium cycle of GTHTR300. The factors which were considered in this cost evaluation include uranium price, conversion, enrichment, fabrication, storage of spent fuel, reprocessing, and waste disposal. The fuel cycle cost of GTHTR300 was estimated at about 1.07 yen/kWh. If the back-end cost of reprocessing and waste disposal is included and assumed to be nearly equivalent to LWR, the fuel cycle cost of GTHTR300 was estimated to be about 1.31 yen/kWh. Furthermore, the effects on fuel fabrication cost by such of fuel specification parameters as enrichment, the number of fuel types, and the layer thickness were considered. Even if the enrichment varies from 10 to 20%, the number of fuel types change from 1 to 4, the 1st layer thickness of fuel changes by 30 {mu}m, or the 2nd layer to the 4th layer thickness of fuel changes by 10 {mu}m, the impact on fuel fabrication cost was evaluated to be negligible. (author)

  1. Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels

    Energy Technology Data Exchange (ETDEWEB)

    Olsen, A.R.; Judkins, R.R. (comps.)

    1979-12-01

    The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O/sub 2/ fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required.

  2. Electrical start-up for diesel fuel processing in a fuel-cell-based auxiliary power unit

    Science.gov (United States)

    Samsun, Remzi Can; Krupp, Carsten; Tschauder, Andreas; Peters, Ralf; Stolten, Detlef

    2016-01-01

    As auxiliary power units in trucks and aircraft, fuel cell systems with a diesel and kerosene reforming capacity offer the dual benefit of reduced emissions and fuel consumption. In order to be commercially viable, these systems require a quick start-up time with low energy input. In pursuit of this end, this paper reports an electrical start-up strategy for diesel fuel processing. A transient computational fluid dynamics model is developed to optimize the start-up procedure of the fuel processor in the 28 kWth power class. The temperature trend observed in the experiments is reproducible to a high degree of accuracy using a dual-cell approach in ANSYS Fluent. Starting from a basic strategy, different options are considered for accelerating system start-up. The start-up time is reduced from 22 min in the basic case to 9.5 min, at an energy consumption of 0.4 kW h. Furthermore, an electrical wire is installed in the reformer to test the steam generation during start-up. The experimental results reveal that the generation of steam at 450 °C is possible within seconds after water addition to the reformer. As a result, the fuel processor can be started in autothermal reformer mode using the electrical concept developed in this work.

  3. Nongray-gas Effects in Modeling of Large-scale Oxy-fuel Combustion Processes

    DEFF Research Database (Denmark)

    Yin, Chungen

    2012-01-01

    , among which radiative heat transfer under oxy-fuel conditions is one of the fundamental issues. This paper demonstrates the nongray-gas effects in modeling of large-scale oxy-fuel combustion processes. Oxy-fuel combustion of natural gas in a large-scale utility boiler is numerically investigated...... cases. The simulation results show that the gray and non-gray calculations of the same oxy-fuel WSGGM make distinctly different predictions in the wall radiative heat transfer, incident radiative flux, radiative source, gas temperature and species profiles. In relative to the non-gray implementation...

  4. Seasonal signals in the reprocessed GPS coordinate time series

    Science.gov (United States)

    Kenyeres, A.; van Dam, T.; Figurski, M.; Szafranek, K.

    2008-12-01

    The global (IGS) and regional (EPN) CGPS time series have already been studied in detail by several authors to analyze the periodic signals and noise present in the long term displacement series. The comparisons indicated that the amplitude and phase of the CGPS derived seasonal signals mostly disagree with the surface mass redistribution models. The CGPS results are highly overestimating the seasonal term, only about 40% of the observed annual amplitude can be explained with the joint contribution of the geophysical models (Dong et al. 2002). Additionally the estimated amplitudes or phases are poorly coherent with the models, especially at sites close to coastal areas (van Dam et al, 2007). The conclusion of the studies was that the GPS results are distorted by analysis artifacts (e.g. ocean tide loading, aliasing of unmodeled short periodic tidal signals, antenna PCV models), monument thermal effects and multipath. Additionally, the GPS series available so far are inhomogeneous in terms of processing strategy, applied models and reference frames. The introduction of the absolute phase center variation (PCV) models for the satellite and ground antennae in 2006 and the related reprocessing of the GPS precise orbits made a perfect ground and strong argument for the complete re-analysis of the GPS observations from global to local level of networks. This enormous work is in progress within the IGS and a pilot analysis was already done for the complete EPN observations from 1996 to 2007 by the MUT group (Military University of Warsaw). The quick analysis of the results proved the expectations and the superiority of the reprocessed data. The noise level (weekly coordinate repeatability) was highly reduced making ground for the later analysis on the daily solution level. We also observed the significant decrease of the seasonal term in the residual coordinate time series, which called our attention to perform a repeated comparison of the GPS derived annual periodicity

  5. SYSTEM AND PROCESS FOR PRODUCTION OF METHANOL FROM COMBINED WIND TURBINE AND FUEL CELL POWER

    Science.gov (United States)

    The paper examines an integrated use of ultra-clean wind turbines and high temperature fuel cells to produce methanol, especially for transportation purposes. The principal utility and application of the process is the production of transportation fuel from domestic resources to ...

  6. SYSTEM AND PROCESS FOR PRODUCTION OF METHANOL FROM COMBINED WIND TURBINE AND FUEL CELL POWER

    Science.gov (United States)

    The paper examines an integrated use of ultra-clean wind turbines and high temperature fuel cells to produce methanol, especially for transportation purposes. The principal utility and application of the process is the production of transportation fuel from domestic resources to ...

  7. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    Science.gov (United States)

    Collette, R.; King, J.; Buesch, C.; Keiser, D. D.; Williams, W.; Miller, B. D.; Schulthess, J.

    2016-07-01

    The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends when comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. The results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.

  8. Processes for converting biomass-derived feedstocks to chemicals and liquid fuels

    Energy Technology Data Exchange (ETDEWEB)

    Held, Andrew; Woods, Elizabeth; Cortright, Randy; Gray, Matthew

    2016-07-05

    The present invention provides processes, methods, and systems for converting biomass-derived feedstocks to liquid fuels and chemicals. The method generally includes the reaction of a hydrolysate from a biomass deconstruction process with hydrogen and a catalyst to produce a reaction product comprising one of more oxygenated compounds. The process also includes reacting the reaction product with a condensation catalyst to produce C.sub.4+ compounds useful as fuels and chemicals.

  9. Processes for converting biomass-derived feedstocks to chemicals and liquid fuels

    Science.gov (United States)

    Held, Andrew; Woods, Elizabeth; Cortright, Randy; Gray, Matthew

    2016-07-05

    The present invention provides processes, methods, and systems for converting biomass-derived feedstocks to liquid fuels and chemicals. The method generally includes the reaction of a hydrolysate from a biomass deconstruction process with hydrogen and a catalyst to produce a reaction product comprising one of more oxygenated compounds. The process also includes reacting the reaction product with a condensation catalyst to produce C.sub.4+ compounds useful as fuels and chemicals.

  10. Literature on fabrication of tungsten for application in pyrochemical processing of spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Edstrom, C.M.; Phillips, A.G.; Johnson, L.D.; Corle, R.R.

    1980-10-11

    The pyrochemical processing of nuclear fuels requires crucibles, stirrers, and transfer tubing that will withstand the temperature and the chemical attack from molten salts and metals used in the process. This report summarizes the literature that pertains to fabrication (joining, chemical vapor deposition, plasma spraying, forming, and spinning) is the main theme. This report also summarizes a sampling of literature on molbdenum and the work previously performed at Argonne National Laboratory on other container materials used for pyrochemical processing of spent nuclear fuels.

  11. Processes for converting biomass-derived feedstocks to chemicals and liquid fuels

    Science.gov (United States)

    Held, Andrew; Woods, Elizabeth; Cortright, Randy; Gray, Matthew

    2017-05-23

    The present invention provides processes, methods, and systems for converting biomass-derived feedstocks to liquid fuels and chemicals. The method generally includes the reaction of a hydrolysate from a biomass deconstruction process with hydrogen and a catalyst to produce a reaction product comprising one of more oxygenated compounds. The process also includes reacting the reaction product with a condensation catalyst to produce C.sub.4+ compounds useful as fuels and chemicals.

  12. Chemical Engineering Division fuel cycle programs. Quarterly progress report, April-June 1979. [Pyrochemical/dry processing; waste encapsulation in metal; transport in geologic media

    Energy Technology Data Exchange (ETDEWEB)

    Steindler, M.J.; Ader, M.; Barletta, R.E.

    1980-09-01

    For pyrochemical and dry processing materials development included exposure to molten metal and salt of Mo-0.5% Ti-0.07% Ti-0.01% C, Mo-30% W, SiC, Si/sub 2/ON/sub 2/, ZrB/sub 2/-SiC, MgAl/sub 2/O/sub 4/, Al/sub 2/O/sub 3/, AlN, HfB/sub 2/, Y/sub 2/O/sub 3/, BeO, Si/sub 3/N/sub 4/, nickel nitrate-infiltrated W, W-coated Mo, and W-metallized alumina-yttria. Work on Th-U salt transport processing included solubility of Th in liquid Cd, defining the Cd-Th and Cd-Mg-Th phase diagrams, ThO/sub 2/ reduction experiments, and electrolysis of CaO in molten salt. Work on pyrochemical processes and associated hardware for coprocessing U and Pu in spent FBR fuels included a second-generation computer model of the transport process, turntable transport process design, work on the U-Cu-Mg system, and U and Pu distribution coefficients between molten salt and metal. Refractory metal vessels are being service-life tested. The chloride volatility processing of Th-based fuel was evaluated for its proliferation resistance, and a preliminary ternary phase diagram for the Zn-U-Pu system was computed. Material characterization and process analysis were conducted on the Exportable Pyrochemical process (Pyro-Civex process). Literature data on oxidation of fissile metals to oxides were reviewed. Work was done on chemical bases for the reprocessing of actinide oxides in molten salts. Flowsheets are being developed for the processing of fuel in molten tin. Work on encapsulation of solidified radioactive waste in metal matrix included studies of leach rate of crystalline waste materials and of the impact resistance of metal-matrix waste forms. In work on the transport properties of nuclear waste in geologic media, adsorption of Sr on oolitic limestone was studied, as well as the migration of Cs in basalt. Fitting of data on the adsorption of iodate by hematite to a mathematical model was attempted.

  13. Fundamental phenomena on fuel decomposition and boundary layer combustion processes with applications to hybrid rocket motors

    Science.gov (United States)

    Kuo, Kenneth K.; Lu, Y. C.; Chiaverini, Martin J.; Harting, George C.

    1994-11-01

    An experimental study on the fundamental processes involved in fuel decomposition and boundary layer combustion in hybrid rocket motors is being conducted at the High Pressure Combustion Laboratory of the Pennsylvania State University. This research should provide a useful engineering technology base in the development of hybrid rocket motors as well as a fundamental understanding of the complex processes involved in hybrid propulsion. A high pressure slab motor has been designed and manufactured for conducting experimental investigations. Oxidizer (LOX or GOX) supply and control systems have been designed and partly constructed for the head-end injection into the test chamber. Experiments using HTPB fuel, as well as fuels supplied by NASA designated industrial companies will be conducted. Design and construction of fuel casting molds and sample holders have been completed. The portion of these items for industrial company fuel casting will be sent to the McDonnell Douglas Aerospace Corporation in the near future. The study focuses on the following areas: observation of solid fuel burning processes with LOX or GOX, measurement and correlation of solid fuel regression rate with operating conditions, measurement of flame temperature and radical species concentrations, determination of the solid fuel subsurface temperature profile, and utilization of experimental data for validation of a companion theoretical study (Part 2) also being conducted at PSU.

  14. Fundamental phenomena on fuel decomposition and boundary layer combustion processes with applications to hybrid rocket motors

    Science.gov (United States)

    Kuo, Kenneth K.; Lu, Y. C.; Chiaverini, Martin J.; Harting, George C.

    1994-01-01

    An experimental study on the fundamental processes involved in fuel decomposition and boundary layer combustion in hybrid rocket motors is being conducted at the High Pressure Combustion Laboratory of the Pennsylvania State University. This research should provide a useful engineering technology base in the development of hybrid rocket motors as well as a fundamental understanding of the complex processes involved in hybrid propulsion. A high pressure slab motor has been designed and manufactured for conducting experimental investigations. Oxidizer (LOX or GOX) supply and control systems have been designed and partly constructed for the head-end injection into the test chamber. Experiments using HTPB fuel, as well as fuels supplied by NASA designated industrial companies will be conducted. Design and construction of fuel casting molds and sample holders have been completed. The portion of these items for industrial company fuel casting will be sent to the McDonnell Douglas Aerospace Corporation in the near future. The study focuses on the following areas: observation of solid fuel burning processes with LOX or GOX, measurement and correlation of solid fuel regression rate with operating conditions, measurement of flame temperature and radical species concentrations, determination of the solid fuel subsurface temperature profile, and utilization of experimental data for validation of a companion theoretical study (Part 2) also being conducted at PSU.

  15. Aqueous processing of U-10Mo scrap for high performance research reactor fuel

    Science.gov (United States)

    Youker, Amanda J.; Stepinski, Dominique C.; Maggos, Laura E.; Bakel, Allen J.; Vandegrift, George F.

    2012-08-01

    The Global Threat Reduction Initiative (GTRI) Conversion program, which is part of the US government's National Nuclear Security Administration (NNSA), supports the conversion of civilian use of highly enriched uranium (HEU) to low enriched uranium (LEU) for reactor fuel and targets. The reason for conversion is to eliminate the use of any material that may pose a threat to the United States or other foreign countries. High performance research reactors (HPRRs) cannot make the conversion to a standard LEU fuel because they require a more dense fuel to meet their performance requirements. As a result, a more dense fuel consisting of a monolithic uranium-molybdenum alloy containing 10% (w/w) Mo with Al cladding and a Zr bonding-layer is being considered. Significant losses are expected in the fabrication of this fuel, so a means to recycle the scrap pieces is needed. Argonne National Laboratory has developed an aqueous-processing flowsheet for scrap recovery in the fuel fabrication process for high-density LEU-monolithic fuel based on data found in the literature. Experiments have been performed to investigate dissolution conditions for solutions containing approximately 20 g-U/L and 50 g-U/L with and without Fe(NO3)3. HNO3 and HF concentrations have been optimized for timely dissolution of the fuel scrap and prevention of the formation of the U-Zr2 intermetallic, explosive complex, while meeting the requirements needed for further processing.

  16. Monitoring of endoscope reprocessing with an adenosine triphosphate (ATP) bioluminescence method.

    Science.gov (United States)

    Parohl, Nina; Stiefenhöfer, Doris; Heiligtag, Sabine; Reuter, Henning; Dopadlik, Dana; Mosel, Frank; Gerken, Guido; Dechêne, Alexander; Heintschel von Heinegg, Evelyn; Jochum, Christoph; Buer, Jan; Popp, Walter

    2017-01-01

    Background: The arising challenges over endoscope reprocessing quality proposes to look for possibilities to measure and control the process of endoscope reprocessing. Aim: The goal of this study was to evaluate the feasibility of monitoring endoscope reprocessing with an adenosine triphosphate (ATP) based bioluminescence system. Methods: 60 samples of eight gastroscopes have been assessed from routine clinical use in a major university hospital in Germany. Endoscopes have been assessed with an ATP system and microbial cultures at different timepoints during the reprocessing. Findings: After the bedside flush the mean ATP level in relative light units (RLU) was 19,437 RLU, after the manual cleaning 667 RLU and after the automated endoscope reprocessor (AER) 227 RLU. After the manual cleaning the mean total viable count (TVC) per endoscope was 15.3 CFU/10 ml, and after the AER 5.7 CFU/10 ml. Our results show that there are reprocessing cycles which are not able to clean a patient used endoscope. Conclusion: Our data suggest that monitoring of flexible endoscope with ATP can identify a number of different influence factors, like the endoscope condition and the endoscopic procedure, or especially the quality of the bedside flush and manual cleaning before the AER. More process control is one option to identify and improve influence factors to finally increase the overall reprocessing quality, best of all by different methods. ATP measurement seems to be a valid technique that allows an immediate repeat of the manual cleaning if the ATP results after manual cleaning exceed the established cutoff of 200 RLU.

  17. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Benedict, R.W.; Goff, K.M.

    1993-01-01

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

  18. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Benedict, R.W.; Goff, K.M.

    1993-03-01

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

  19. Test program on the release characteristics of Kr-85 from remote fuel fabrication process

    Energy Technology Data Exchange (ETDEWEB)

    Park, Geun Il; Lee, J. W.; Kim, S. S. [and others

    2003-01-01

    In order to investigate the release kinetics of Kr-85 fission gas during DUPIC fuel fabrication process using spent fuel materials, the test equipment and its procedure was developed. The purpose of this test involves the measurement of Kr-85 released during OREOX process in DUPIC fuel fabrication as well as the analysis of fission- gas release kinetics with the variation of fuel fabrication conditions. Gas monitoring system installed inside glove box was located at out-cell of DFDF(DUPIC Fuel Fabrication Facility) at which OREOX and tube furnaces have already installed inside hot cell. The use of glove box is aimed for preventing a gas release from sampling gas line under negative pressure. Based on the allowable discharge concentration of Kr-85 to environment and the preliminary analysis assuming total released amount a year, environmental impact according to Kr-85 measuring test would be minimal.

  20. Numerical Studies on Controlling Gaseous Fuel Combustion by Managing the Combustion Process of Diesel Pilot Dose in a Dual-Fuel Engine

    Directory of Open Access Journals (Sweden)

    Mikulski Maciej

    2015-06-01

    Full Text Available Protection of the environment and counteracting global warming require finding alternative sources of energy. One of the methods of generating energy from environmentally friendly sources is increasing the share of gaseous fuels in the total energy balance. The use of these fuels in compression-ignition (CI engines is difficult due to their relatively high autoignition temperature. One solution for using these fuels in CI engines is operating in a dualfuel mode, where the air and gas mixture is ignited with a liquid fuel dose. In this method, a series of relatively complex chemical processes occur in the engine's combustion chamber, related to the combustion of individual fuel fractions that interact with one another. Analysis of combustion of specific fuels in this type of fuel injection to the engine is difficult due to the fact that combustion of both fuel fractions takes place simultaneously. Simulation experiments can be used to analyse the impact of diesel fuel combustion on gaseous fuel combustion. In this paper, we discuss the results of simulation tests of combustion, based on the proprietary multiphase model of a dual-fuel engine. The results obtained from the simulation allow for analysis of the combustion process of individual fuels separately, which expands the knowledge obtained from experimental tests on the engine.

  1. Integrated scheme of long-term for spent fuel management of power nuclear reactors; Esquema integrado de largo plazo para la administracion de combustible gastado de reactores nucleares de potencia

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Palacios H, J. C.; Martinez C, E., E-mail: ramon-ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    After of irradiation of the nuclear fuel in the reactor core, is necessary to store it for their cooling in the fuel pools of the reactor. This is the first step in a processes series before the fuel can reach its final destination. Until now there are two options that are most commonly accepted for the end of the nuclear fuel cycle, one is the open nuclear fuel cycle, requiring a deep geological repository for the fuel final disposal. The other option is the fuel reprocessing to extract the plutonium and uranium as valuable materials that remaining in the spent fuel. In this study the alternatives for the final part of the fuel cycle, which involves the recycling of plutonium and the minor actinides in the same reactor that generated them are shown. The results shown that this is possible in a thermal reactor and that there are significant reductions in actinides if they are recycled into reactor fuel. (Author)

  2. Advanced Safeguards Approaches for New TRU Fuel Fabrication Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Ehinger, Michael H.; Boyer, Brian; Therios, Ike; Bean, Robert; Dougan, A.; Tolk, K.

    2007-12-15

    This second report in a series of three reviews possible safeguards approaches for the new transuranic (TRU) fuel fabrication processes to be deployed at AFCF – specifically, the ceramic TRU (MOX) fuel fabrication line and the metallic (pyroprocessing) line. The most common TRU fuel has been fuel composed of mixed plutonium and uranium dioxide, referred to as “MOX”. However, under the Advanced Fuel Cycle projects custom-made fuels with higher contents of neptunium, americium, and curium may also be produced to evaluate if these “minor actinides” can be effectively burned and transmuted through irradiation in the ABR. A third and final report in this series will evaluate and review the advanced safeguards approach options for the ABR. In reviewing and developing the advanced safeguards approach for the new TRU fuel fabrication processes envisioned for AFCF, the existing international (IAEA) safeguards approach at the Plutonium Fuel Production Facility (PFPF) and the conceptual approach planned for the new J-MOX facility in Japan have been considered as a starting point of reference. The pyro-metallurgical reprocessing and fuel fabrication process at EBR-II near Idaho Falls also provided insight for safeguarding the additional metallic pyroprocessing fuel fabrication line planned for AFCF.

  3. Implications of sedimentological and hydrological processes on the distribution of radionuclides in a salt marsh near Sellafield, Cumbria

    Energy Technology Data Exchange (ETDEWEB)

    Carr, A.P.; Blackley, M.W.L.

    1985-01-01

    The report examines sedimentological and hydrological processes affecting a salt marsh in the Ravenglass estuary, which is situated south of the Sellafield nuclear-fuel-reprocessing plant. The results are discussed in the context of the distribution of low-level radioactive effluent at the site.

  4. Economical process for growing seaweed as biomass fuel source

    Energy Technology Data Exchange (ETDEWEB)

    Lagovskiy, V.

    1985-10-10

    Calculations made by researchers of Moscow State University have shown that the Aral Sea is capable of providing energy for almost the entire country. An experimental unit called Biosolar, for growing such energy already exists. Up to 40 liters of fuel gas a day can be gathered from a single square meter of plant beds. Seaweed yields biomass, which is placed in special vats. There it is eaten by bacteria, which release methane.

  5. Optimization of fuel supply map during starting process of electronic controlled diesel engine

    Institute of Scientific and Technical Information of China (English)

    Jinguang LIANG; Xiumin YU; Yue GAO; Yunkai WANG; Hongyang YU; Baoli GONG

    2008-01-01

    Tests were conducted to study influence of fuel supply map during the starting process of an electronic con-trolled diesel engine using an electronic controlled diesel engine which was made up of a CA498Z diesel engine, a VP37 elec-tronic controlled distributor injection pump management system and a VS100 calibration system. The calibration pro-cess of starting fuel supply map was educed under the principle of low HC emission and rapid starting velocity. The cal-ibration methods of starting fuel supply map were obtained.

  6. Fuel ethanol production from lignocellulose: a challenge for metabolic engineering and process integration

    DEFF Research Database (Denmark)

    Zaldivar, Jesus; Nielsen, Jens; Olsson, Lisbeth

    2001-01-01

    With industrial development growing rapidly, there is a need for environmentally sustainable energy sources. Bioethanol (ethanol from biomass) is an attractive, sustainable energy source to fuel transportation. Based on the premise that fuel bioethanol can contribute to a cleaner environment...... and with the implementation of environmental protection laws in many countries, demand for this fuel is increasing. Efficient ethanol production processes and cheap substrates are needed. Current ethanol production processes using crops such as sugar cane and corn are well-established; however, utilization of a cheaper...

  7. Structural damage and chemical contaminants on reprocessed arthroscopic shaver blades.

    Science.gov (United States)

    Kobayashi, Masahiko; Nakagawa, Yasuaki; Okamoto, Yukihiro; Nakamura, Shinichiro; Nakamura, Takashi

    2009-02-01

    In response to socioeconomic pressure to cut budgets in medicine, single-use surgical instruments are often reprocessed despite potential biological hazard. To evaluate the quality and contaminants of reprocessed shaver blades. Reprocessed shaver blades have mechanical damage and chemical contamination. Controlled laboratory study. Seven blades and 3 abraders were reprocessed 1 time or 3 times and then were assessed. In the first part of the study, structural damage on the blades after 3 reprocessings was compared to that after 1 reprocessing using optical microscopy. In the second part, surface damage was observed using optical microscopy and scanning electron microscopy; elemental and chemical analyses of contaminants found by the microscopy were performed using scanning electron microscopy/energy dispersive x-ray spectroscopy, scanning Auger microscopy, and Fourier transform infrared spectroscopy. Optical microscopic examination revealed abrasion on the surface of the inner blade and cracks on the inner tube after 1 reprocessing. These changes were more evident after 3 reprocessings. Scanning electron microscopy/energy dispersive x-ray spectroscopy of the inner cutter of the blade reprocessed once showed contaminants containing calcium, carbon, oxygen, and silicon, and Fourier transform infrared spectroscopy demonstrated biological protein consisting mainly of collagen, some type of salts, and polycarbonate used in plastic molding. Scanning electron microscopy/energy dispersive x-ray spectroscopy of the inner cutter of the reprocessed abrader revealed contaminants containing carbon, calcium, phosphorous, and oxygen, and Fourier transform infrared spectroscopy showed H2O, hydroxyapatite, and hydroxyl proteins. Scanning Auger microscopy showed that the tin-nickel plating on the moving blade and abrader was missing in some locations. This is the first study to evaluate both mechanical damage and chemical contaminants containing collagen, hydroxyapatite, and salts

  8. Reprocessing the Southern Hemisphere ADditional OZonesondes (SHADOZ) Database for Long-Term Trend Analyses

    Science.gov (United States)

    Witte, J. C.; Thompson, A. M.; Coetzee, G.; Fujiwara, M.; Johnson, B. J.; Sterling, C. W.; Cullis, P.; Ashburn, C. E.; Jordan, A. F.

    2015-12-01

    SHADOZ is a large archive of tropical balloon-bone ozonesonde data at NASA/Goddard Space Flight Center with data from 14 tropical and subtropical stations provided by collaborators in Europe, Asia, Latin America and Africa . The SHADOZ time series began in 1998, using electrochemical concentration cell (ECC) ozonesondes. Like many long-term sounding stations, SHADOZ is characterized by variations in operating procedures, launch protocols, and data processing such that biases within a data record and among sites appear. In addition, over time, the radiosonde and ozonesonde instruments and data processing protocols have changed, adding to the measurement uncertainties at individual stations and limiting the reliability of ozone profile trends and continuous satellite validation. Currently, the ozonesonde community is engaged in reprocessing ECC data, with an emphasis on homogenization of the records to compensate for the variations in instrumentation and technique. The goals are to improve the information and integrity of each measurement record and to support calculation of more reliable trends. We illustrate the reprocessing activity of SHADOZ with selected stations. We will (1) show reprocessing steps based on the recent WMO report that provides post-processing guidelines for ozonesondes; (2) characterize uncertainties in various parts of the ECC conditioning process; and (3) compare original and reprocessed data to co-located ground and satellite measurements of column ozone.

  9. Development of Demo of Solution Measurement and Monitoring System in Reprocessing Plants

    Institute of Scientific and Technical Information of China (English)

    LIANG; Qing-lei; CHANG; Li; LI; Jing-huai; LU; Jie; TIAN; Yuan

    2015-01-01

    There are numerous unattended measurement and monitoring systems at reprocessing plants,and the most important one is the solution measurement and monitoring system,which can monitor the stable operation of the process and account the nuclear material of the entire

  10. Development and control of the process for the manufacture of zircaloy-4 tubing for LWBR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Eyler, J.H.

    1981-01-01

    The technical requirements for the Light Water Breeder Reactor (LWBR) fuel elements (fuel rods) imposed certain unique requirements for the low hafnium Zircaloy-4 tubing used as fuel rod cladding. This report describes, in detail, the tube manufacturing process, the product and process controls used, the inspections and tests performed, and the efforts involved in refining a commercial tube reducing process to produce tubes that would satisfy the requirements for LWBR fuel rod cladding.

  11. Physical and economical aspects of Pu multiple recycling on the basis of REMIX reprocessing technology in thermal reactors

    Directory of Open Access Journals (Sweden)

    Teplov Pavel S.

    2016-01-01

    Full Text Available The basic strategy of Russian nuclear energy is propagation of a closed fuel cycle on the basis of fast breeder and thermal reactors, as well as the solution of the spent nuclear fuel accumulation and resource problems. The three variants of multiple Pu and U recycling in Russian pressurized water reactor concept reactors on the basis of REgenerated MIXture of U, Pu oxides (REMIX reprocessing technology are considered in this work. The REMIX fuel is fabricated from an unseparated mixture of uranium and plutonium obtained during spent fuel reprocessing with further makeup by enriched natural U or reactor grade Pu. This makes it possible to recycle several times the total amount of Pu obtained from the spent fuel. The main difference in Pu recycling is the concept of 100% or partial fuel loading of the core. The third variant is heterogeneous composition of enriched uranium and uranium–plutonium mixed oxide fuel pins in one fuel assembly. It should be noted that all fuel assemblies with Pu require the involvement of expensive technologies during manufacturing. These three variants of the full core loadings can be balanced on zero Pu accumulation in the cycle. The various physical and economical aspects of Pu and U multiple recycling in selected variants are observed in the given work.

  12. TRISO-Coated Fuel Processing to Support High Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.

    2002-10-01

    The initial objective of the work described herein was to identify potential methods and technologies needed to disassemble and dissolve graphite-encapsulated, ceramic-coated gas-cooled-reactor spent fuels so that the oxide fuel components can be separated by means of chemical processing. The purpose of this processing is to recover (1) unburned fuel for recycle, (2) long-lived actinides and fission products for transmutation, and (3) other fission products for disposal in acceptable waste forms. Follow-on objectives were to identify and select the most promising candidate flow sheets for experimental evaluation and demonstration and to address the needs to reduce technical risks of the selected technologies. High-temperature gas-cooled reactors (HTGRs) may be deployed in the next -20 years to (1) enable the use of highly efficient gas turbines for producing electricity and (2) provide high-temperature process heat for use in chemical processes, such as the production of hydrogen for use as clean-burning transportation fuel. Also, HTGR fuels are capable of significantly higher burn-up than light-water-reactor (LWR) fuels or fast-reactor (FR) fuels; thus, the HTGR fuels can be used efficiently for transmutation of fissile materials and long-lived actinides and fission products, thereby reducing the inventory of such hazardous and proliferation-prone materials. The ''deep-burn'' concept, described in this report, is an example of this capability. Processing of spent graphite-encapsulated, ceramic-coated fuels presents challenges different from those of processing spent LWR fuels. LWR fuels are processed commercially in Europe and Japan; however, similar infrastructure is not available for processing of the HTGR fuels. Laboratory studies on the processing of HTGR fuels were performed in the United States in the 1960s and 1970s, but no engineering-scale processes were demonstrated. Currently, new regulations concerning emissions will impact the

  13. Advanced diagnostics in oxy-fuel combustion processes

    Energy Technology Data Exchange (ETDEWEB)

    Brix, J.; Clausen, Soennik; Degn Jensen, A. (Technical Univ. of Denmark. CHEC Research Centre, Kgs. Lyngby (Denmark)); Boeg Toftegaard, M. (DONG Energy Power, Hvidovre (Denmark))

    2012-07-01

    This report sums up the findings in PSO-project 010069, ''Advanced Diagnostics in Oxy-Fuel Combustion Processes''. Three areas of optic diagnostics are covered in this work: - FTIR measurements in a 30 kW swirl burner. - IR measurements in a 30 kW swirl burner. - IR measurements in a laboratory scale fixed bed reactor. The results obtained in the swirl burner have proved the FTIR method as a valuable technique for gas phase temperature measurements. When its efficacy is evaluated against traditional thermocouple measurements, two cases, with and without probe beam stop, must however be treated separately. When the FTIR probe is operated with the purpose of gas phase concentration measurements the probe needs to operate with a beam stop mounted in front of it. With this beam stop in place it was shown that the measured gas phase temperature was affected by cooling, induced by the cooled beam stop. Hence, for a more accurate determination of gas phase temperatures the probe needed to operate without the beam stop. When this was the case, the FTIR probe showed superior to traditional temperature measurements using a thermocouple as it could measure the fast temperature fluctuations. With the beam stop in place the efficacy of the FTIR probe for gas temperature determination was comparable to the use of a traditional thermocouple. The evaluation of the FTIR technique regarding estimation of gas phase concentrations of H{sub 2}O, CO{sub 2} and CO showed that the method is reliable though it cannot be stated as particularly accurate. The accuracy of the method is dependent on the similarity of the reference emission spectra of the gases with those obtained in the experiments, as the transmittance intensity is not a linear function of concentration. The length of the optical path also affects the steadiness of the measurements. The length of the optical path is difficult to adjust on the small scales that are the focus of this work. However

  14. Nuclear fuels - Present and future

    Science.gov (United States)

    Olander, D.

    2009-06-01

    The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

  15. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities, Sections 15-19

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.

    1982-09-01

    Information is presented under the following section headings: fuel reprocessing; spent fuel and high-level and transuranic waste storage; spent fuel and high-level and transuranic waste disposal; low-level and intermediate-level waste disposal; and, transportation of radioactive materials in the nuclear fuel cycle. In each of the first three sections a description is given on the mainline process, effluent processing and waste management systems, plant layout, and alternative process schemes. Safety information and a summary are also included in each. The section on transport of radioactive materials includes information on the transportation of uranium ore, uranium ore concentrate, UF/sub 6/, PuO/sub 2/ powder, unirradiated uranium and mixed-oxide fuel assemblies, spent fuel, solidified high-level waste, contact-handled transuranic waste, remote-handled transuranic waste, and low and intermediate level nontransuranic waste. A glossary is included. (JGB)

  16. High liquid fuel yielding biofuel processes and a roadmap for the future transportation

    Science.gov (United States)

    Singh, Navneet R.

    In a fossil-fuel deprived world when crude oil will be scarce and transportation need cannot be met with electricity and transportation liquid fuel must be produced, biomass derived liquid fuels can be a natural replacement. However, the carbon efficiency of the currently known biomass to liquid fuel conversion processes ranges from 35-40%, yielding 90 ethanol gallon equivalents (ege) per ton of biomass. This coupled with the fact that the efficiency at which solar energy is captured by biomass (syngas derived from coal gasification (H2Bioil-C) or a natural gas reformer (H 2Bioil-NG) is used to supply the hydrogen and process heat for the biomass fast-hydropyrolysis/hydrodeoxygenation. Another off-shoot of the H2Bioil process is the H2Bioil-B process, where hydrogen required for the hydropyrolysis is obtained from gasification of a fraction of the biomass. H2Bioil-B achieves the highest liquid fuel yield (126-146 ege/ton of biomass) reported in the literature for any self-contained conversion of biomass to biofuel. Finally, an integration of the H2Bioil process with the H2CAR process is suggested which can achieve 100% carbon efficiency (330 ege/ton of biomass) at the expense of 0.24 kg hydrogen/liter of oil. A sun-to-fuel efficiency analysis shows that extracting CO2 from air and converting it to liquid fuel is at least two times more efficient than growing dedicated fuel crops and converting them to liquid fuel even for the highest biomass growth rates feasible by algae. This implies that liquid fuel should preferably be produced from sustainably available waste (SAW) biomass first and if the SAW biomass is unable to meet the demand for liquid fuel, then, CO2 should be extracted from air and converted to liquid fuel, rather than growing biomass. Furthermore, based on the Sun-to-Wheels recovery for different transportation pathways, synergistic and complementary use of electricity, hydrogen and biomass, all derived from solar energy, is presented in an energy

  17. Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P. O. 1236909 Final report

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

    1978-12-01

    The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel.

  18. Spent Nuclear Fuel (SNF) Process Validation Technical Support Plan

    Energy Technology Data Exchange (ETDEWEB)

    SEXTON, R.A.

    2000-03-13

    The purpose of Process Validation is to confirm that nominal process operations are consistent with the expected process envelope. The Process Validation activities described in this document are not part of the safety basis, but are expected to demonstrate that the process operates well within the safety basis. Some adjustments to the process may be made as a result of information gathered in Process Validation.

  19. Scrap tire process turns waste into fuel. [USA - New York

    Energy Technology Data Exchange (ETDEWEB)

    Tesla, M.R. (New York State Electric and Gas Corporation, Binghamton, NY (United States). Alternate Fuels Development)

    1994-05-01

    New York State Electric Gas Corp. (NYSEG) has in the last three years burned more than 1.3 million tyres to generate power, saving more than 16,000 tons of coal. The article describes how at the 73 MW Jennison power station scrap tires are chopped into 2x2 in. chips prior to mixing with coal near the coal pile (in a 20.80 tire chip to coal ratio) and combustion in stoker boilers with travelling chain grates. NYSEG has also test burned other alternative fuels such as creosote-treated wood and coal tar soils. 6 photos.

  20. Chemical Engineering Division Fuel Cycle Programs. Quarterly progress report, April-June 1978. [Advanced solvent extraction; accidents; pyrochemical; radwaste in metal matrix; waste migration

    Energy Technology Data Exchange (ETDEWEB)

    Steindler, M. J.; Ader, M.; Barletta, R. E.

    1979-12-01

    Fuel cycle studies reported include development of centrifugal contactors for Purex processes. Tricaprylmethyl-ammonium nitrate and di-n-amyl-n-amylphosphonate are being evaluated as Thorex extractants. Dispersion of uranium and plutonium by fires, and mechanisms for subdividing and dispersing liquids and solids were reviewed. In the pyrochemical and dry processing program, a facility for testing containment materials is under construction; a flowsheet for carbide fuel processing has been designed and studies of carbide reactions in bismuth are underway; salt transport processes are being studied; process-size refractory metal vessels are being fabricated; the feasibility of AIROX reprocessing is being determined; the solubility of UO/sub 2/, UO/sub 2/ + fission products, and PuO/sub 2/ in molten alkali metal nitrates, has been investigated; a flowsheet was developed for reprocessing actinide oxides in molten salts; preparation of Th-U carbide from the oxide is being studied; new flowsheets based on the Dow Aluminum Pyrometallurgical process for reprocessing of spent uranium metal fuel have been prepared; the chloride volitility processing of thorium-based fuels is being studied; the reprocessing of (Th,U)O/sub 2/ solid solution in KCl-LiCl-ThCl/sub 4/-Th is being studied; and a flowsheet for processing spent nuclear fuel in molten tin has been constructed. Leach rates of simulated encapsulated waste forms in a metal matrix were studied. Nine criteria for handling waste cladding hulls were established. Strontium and tin migration in glauconite columns was measured. Radioactive Sr in a stream of water moved through oolitic limestone as rapidly as water, but in a stream of water equilibrated with the limestone, Sr moved through the limestone one-tenth as fast. Migration of trace quantities of Cs and I through kaolinite was studied. 88 figures, 53 tables.

  1. Managing probe applications effectively? The process of pre-commercial fuel cell applications

    Energy Technology Data Exchange (ETDEWEB)

    Hellman, H.L. [Delft Univ. of Technology, Delft (Netherlands)

    2007-07-01

    There has been a high degree of uncertainty regarding market adoption of proton exchange membrane (PEM) fuel cell technologies, making it challenging for fuel cell firms to determine which markets to pursue at which point in time. Probes, such as prototypes, demonstrations and field trials, are market experiments conducted by fuel cell firms and industry that are central to the commercialization of fuel cell technology. However, demonstration projects are extremely resource and time intensive, and outcomes are uncertain. This paper reviewed innovation management literature on the characteristics of the technology application process. Case study research of four fuel cell firms were discussed in terms of probe objectives and outcomes. An analysis of probe application patterns over time was also presented. Several fuel cell projects were also analyzed in terms of motivations and lessons learned. The paper focused on young independent fuel cell developers that had limited resources to allocate and waste resources on ineffective probe applications. The findings complement prior research that have found probe applications a valuable tool for learning, stimulating market demand and network formation. The paper made several recommendations regarding the main objectives of the fuel cell project, to set realistic goals and align expectations between consortia members. In addition, it was found that a short term demonstration project may appear to be an attractive application, but a project with a longer term vision and a commercial value proposition was more likely to attract stakeholders. 16 refs., 3 figs.

  2. An integrated MEMS infrastructure for fuel processing: hydrogen generation and separation for portable power generation

    Science.gov (United States)

    Varady, M. J.; McLeod, L.; Meacham, J. M.; Degertekin, F. L.; Fedorov, A. G.

    2007-09-01

    Portable fuel cells are an enabling technology for high efficiency and ultra-high density distributed power generation, which is essential for many terrestrial and aerospace applications. A key element of fuel cell power sources is the fuel processor, which should have the capability to efficiently reform liquid fuels and produce high purity hydrogen that is consumed by the fuel cells. To this end, we are reporting on the development of two novel MEMS hydrogen generators with improved functionality achieved through an innovative process organization and system integration approach that exploits the advantages of transport and catalysis on the micro/nano scale. One fuel processor design utilizes transient, reverse-flow operation of an autothermal MEMS microreactor with an intimately integrated, micromachined ultrasonic fuel atomizer and a Pd/Ag membrane for in situ hydrogen separation from the product stream. The other design features a simpler, more compact planar structure with the atomized fuel ejected directly onto the catalyst layer, which is coupled to an integrated hydrogen selective membrane.

  3. Application of steric exclusion chromatography for the separation of degradation products of the solvent used for the reprocessing of the nuclear fuels; Application de la chromatographie d`exclusion sterique a la separation de produits de degradation du solvant du retraitement des combustibles nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Pozo, C.

    1993-08-01

    The solvent, used in France in Purex reprocessing plants at La Hague is tributylphosphate (TBP) diluted to 30% with a mixture of branched alkanes, for which the main component is branched dodecane (70%). In order to minimize volumes of organic wastes, we have to maintain Purex solvent qualities and to get rid of degradation products. The subject of this memoir concerns among all the degradation products the heaviest molecules. The separation and the identification of these products have been carried out by preparative steric exclusion chromatography, followed by the analysis of the samples by various analytical methods. An inactive residue containing heavy degradation products was prepared according to the process used in the UP3 La Hague plant. The Analysis of this residue using steric exclusion chromatography and GPC/MS methods, shows the presence of three families of compounds heavier than TBP: the ``dimers of TBP`` (provided from the addition of two molecules of TBP), the ``TBP-alkanes`` (the main molecule is the result of the addition of dodecane with TBP), and ``the functionalized TBP`` (hydroxyled TBP, nitrous TBP, nitrated TBP). Plutonium (IV) retention tests were made on the various fractions generated by steric chromatography. They showed that ``the dimers of TBP`` and ``the functionalized TBP`` families are responsible for that retention. These results confirm the good efficiency of the solvent distillation system operated in UP3 plant which allow the elimination of heavy degradation products of the solvent with the residue and then restore excellent extracting properties for the recycled solvent. (author). 35 figs., 69 refs., 15 tabs.

  4. Gasification process of refuse derived fuel in circulating fluidized bed

    Energy Technology Data Exchange (ETDEWEB)

    Ichikawa, S.; Kinoshita, Y.; Lee, C.W.; Itaya, Y.; Mori, S. [Nagoya Univ., Nagoya (Japan). Dept. of Chemical Engineering

    2002-07-01

    This paper presents a fuel gas production system involving gasification of refuse-derived fuel (RDF) in a circulating fluidized bed (CFB). Although RDF is considered to be a viable source of energy, combustion of RDF has not spread widely because of a lack of conventional incinerators, erosion due to hydrogen chloride, and emissions of dioxin. This paper presents the results of an experimental study of the pyrolysis behaviour of 3 kinds of RDF and the particle motion in a cold model CFB. The objective was to clarify operating parameters for optimum control. It was shown that an increase in combustion temperature improves the yield of the combustible gas components and the energy recycling efficiency from the RDF. The highest heating value of pyrolysis gas was obtained at 873 to 973 degrees K. The gas flow rate in the pneumatic valve of the CFB was an important control factor for the circulation flux and solids holdup in the riser. High holdups were observed when minute silica sand particles were used in the CFB. 15 refs., 1 tab., 8 figs.

  5. A review on the status of development in thorium-based nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Woo; Na, S. H.; Lee, Y. W.; Kim, H. S.; Kim, S. H.; Joung, C.Y

    2000-02-01

    Thorium as an alternative nuclear energy source had been widely investigated in the 1950s-1960s because it is more abundant than uranium, but the studies of thorium nuclear fuel cycle were discontinued by political and economic reasons in the 1970s. Recently, however, renewed interest was vested in thorium-based nuclear fuel cycle because it may generate less long-lived minor actinides and has a lower radiotoxicity of high level wastes after reprocessing compared with the thorium fuel cycle. In this state-of the art report, thorium-based nuclear cycle. In this state-of the art report, thorium-based nuclear fuel cycle and fuel fabrication processes developed so far with different reactor types are reviewed and analyzed to establish basic technologies of thorium fuel fabrication which could meet our situation. (author)

  6. Fuel and canister process report for the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Werme, Lars (ed.)

    2006-10-15

    This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The report is written by, and for, experts in the relevant scientific fields. It should though be possible for a generalist in the area of long-term safety assessments of geologic nuclear waste repositories to comprehend the contents of the report. The report is an important part of the documentation of the SR-Can project and an essential reference within the project, providing a scientifically motivated plan for the handling of geosphere processes. It is, furthermore, foreseen that the report will be essential for reviewers scrutinising the handling of geosphere issues in the SR-Can assessment. Several types of fuel will be emplaced in the repository. For the reference case with 40 years of reactor operation, the fuel quantity from boiling water reactors, BWR fuel, is estimated at 7,000 tonnes, while the quantity from pressurized water reactors, PWR fuel, is estimated at about 2,300 tonnes. In addition, 23 tonnes of mixed-oxide fuel (MOX) fuel of German origin from BWR and PWR reactors and 20 tonnes of fuel from the decommissioned heavy water reactor in Aagesta will be disposed of. To allow for future changes in the Swedish nuclear programme, the safety assessment assumes a total of 6,000 canister corresponding to 12,000 tonnes of fuel.

  7. Nitrogen Trifluoride-Based Fluoride- Volatility Separations Process: Initial Studies

    Energy Technology Data Exchange (ETDEWEB)

    McNamara, Bruce K.; Scheele, Randall D.; Casella, Andrew M.; Kozelisky, Anne E.

    2011-09-28

    This document describes the results of our investigations on the potential use of nitrogen trifluoride as the fluorinating and oxidizing agent in fluoride volatility-based used nuclear fuel reprocessing. The conceptual process uses differences in reaction temperatures between nitrogen trifluoride and fuel constituents that produce volatile fluorides to achieve separations and recover valuable constituents. We provide results from our thermodynamic evaluations, thermo-analytical experiments, kinetic models, and provide a preliminary process flowsheet. The evaluations found that nitrogen trifluoride can effectively produce volatile fluorides at different temperatures dependent on the fuel constituent.

  8. Activity of fuel batches processed through Hanford separations plants, 1944 through 1989

    Energy Technology Data Exchange (ETDEWEB)

    Watrous, R.A.; Wootan, D.W.

    1997-07-29

    This document provides a printout of the ``Fuel Activity Database`` (version U6) generated by the Hanford DKPRO code and transmitted to the Los Alamos National Laboratory for input to their ``Hanford Defined Waste`` model of waste tank inventories. This fuel activity file consists of 1,276 records--each record representing the activity associated with a batch of spent reactor fuel processed by month (or shorter period) through individual Hanford separations plants between 1944 and 1989. Each record gives the curies for 46 key radionuclides, decayed to a common reference date of January 1, 1994.

  9. Optimum catalytic process for alcohol fuels from syngas: Seventh quarterly technical progress report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1988-10-29

    The objectives of this contract are to discover and evaluate the catalytic properties of novel homogeneous, heterogeneous, or combination catalytic systems for the production of alcohol fuel extenders from syngas, to evaluate analytically and on the bench scale novel reactor concepts for use in converting syngas to liquid fuel products, and to develop on the bench scale the best combination of chemistry, reactor, and total process configuration to achieve the minimum product cost for conversion of syngas to liquid fuel products. Catalysts investigated include: ruthenium and molybdenum sulfides impregnated with cobalt. 7 figs., 9 tabs.

  10. Fluid dynamics simulations of a fuel processing system; Stroemungsmechanische Modellierung eines Brenngaserzeugungssystems

    Energy Technology Data Exchange (ETDEWEB)

    Scharf, Florian

    2012-07-01

    The present thesis deals with the topic of the complete fluid dynamic modelling of the key components of a fuel processing system and based on it the development of the next generation reactors. Fuel cell auxilliary power units enable an energy efficient power generation for mobile applications with higher on-board power consumption. Enabling the operation of the fuel cell with the available middle-destillate on-board, the fuel is transformed to an hydrogen-rich gas in a fuel processing system consisting of the key components autothermal reformer, water-shift reactor and catalytic burner. The modules of the fuel processing system are thereby integrated within the reactors to obtain a lightweighted and compact overall system. The complete numerical description of theses systems are based on chemical-reaction models, vaporization models of fuel and water in the integrated reactant treatment and as well on models for the integrated heat-exchanger. The strong interaction between the single reactor zones require therefore modelling of the key components as an overall reactor system. The methodology of the present thesis is based on a tight integration of CFD simulations with experimental analysis and the construction of the reactors. The results of carried out prototype testings, post-mortem-analysis and laboratory experiments are taken as basis for the CFD modelling and the reactor construction. As numerical tool for fluid dynamic modelling the CFD software FLUENT was used. In the context of this thesis the CFD modelling library was extended with the purpose of the complete description of key components in the CFD overall model. In these CFD overall models the thermal interactions between the reactor zones as well as the influence of the pulsating fuel injection and the chemical reactions are taken into account. For this purpose a similarity theoretical CFD modell of the pulsating fuel injection was generated based on results from prototype testings and high

  11. Influence of high injection pressure on fuel injection perfomances and diesel engine worcking process

    Directory of Open Access Journals (Sweden)

    Shatrov Mikhail G.

    2015-01-01

    Full Text Available In MADI, investigations are carried out in the field of diesel engine working process perfection for complying with prospective ecological standards such as Euro-6 and Tier-4. The article describes the results of the first stage of experimental research of the influence of injection pressure up to 3000 bar on working processes of diesel engine and its fuel system. Justification of the design of a Common Rail injector for fuel injection under 3000 bar pressure is presented. The influence of raising injection pressure (up to 3000 bar on the fuel spray propagation dynamics is demonstrated. The combined influence of injection pressure (up to 3000 bar and air boost pressure on fuel spray propagation dynamics is shown, including on engine emission and noise.

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

    Science.gov (United States)

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

    2016-05-01

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

  13. Simulation of primary fuel atomization processes at subcritical pressures.

    Energy Technology Data Exchange (ETDEWEB)

    Arienti, Marco

    2013-06-01

    This report documents results from an LDRD project for the first-principles simulation of the early stages of spray formation (primary atomization). The first part describes a Cartesian embedded-wall method for the calculation of flow internal to a real injector in a fully coupled primary calculation. The second part describes the extension to an all-velocity formulation by introducing a momentum-conservative semi-Lagrangian advection and by adding a compressible term in the Poissons equation. Accompanying the description of the new algorithms are verification tests for simple two-phase problems in the presence of a solid interface; a validation study for a scaled-up multi-hole Diesel injector; and demonstration calculations for the closing and opening transients of a single-hole injector and for the high-pressure injection of liquid fuel at supersonic velocity.

  14. An Investigation of Anaerobic Processes in Fuel/Natural Seawater Environments

    Science.gov (United States)

    2012-02-08

    crude oil remaining. Biodiesel is produced from vegetable oils by converting the triglyceride oils to methyl (or ethyl) esters with a process known...separated esters and glycerin. Biodiesel contains no sulfur. In the United States the term "biodiesel" is standardized as fatty acid methyl ester (FAME...particular interest to the defense community. JP-5 (for Jet Propellant) is a kerosene -based fuel for use in aircraft turbine engines. Since the fuel is the

  15. Analysis of nuclear material flow for experimental DUPIC fuel fabrication process at DFDF

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Lee, J. W.; Yang, M. S.; Baik, S. Y.; Lee, E. P

    1999-08-01

    This report describes facilities necessary for manufacturing experiment for DUPIC fuel, manufacturing process and equipment. Nuclear material flows among facilities, in PIEF and IMEF, for irradiation test, for post examination of DUPIC fuel, for quality control, for chemical analysis and for treatment of radioactive waste have been analyzed in details. This may be helpful for DUPIC project participants and facility engineers working in related facilities to understand overall flow for nuclear material and radioactive waste. (Author). 14 refs., 15 tabs., 41 figs.

  16. Experiment on the improvement of OREOX process for fabrication of dry recycling nuclear fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Kim, S. S.; Park, G. I. [and others

    2004-01-01

    The OREOX(Oxidation and REduction of OXide fuel) process has been performed to fabricate dry recycling(DUPIC ; Direct Use of spent PWR fuel In CANDU reactor) nuclear fuel pellets by using spent PWR fuel. Generally, sinterable DUPIC powder has been manufactured from spent PWR fuel pellets by the 3 cycles of oxidation and reduction treatment. The OREOX process is one of the most important processes for DUPIC pellet fabrication. A lot of time more than 37 hours as well as a lot of reaction gas is required to perform 3 cycles of OREOX treatments. In this experiment, 1 cycle OREOX process was adopted to improve the powdering process of DUPIC pellet manufacturing processes. As a result of experiment, the densities of pellets sintered at 1800 .deg. C for 10 hours ranged from 10.15 to 10.22 g/cm{sup 3}(93.8{approx}94.5 % of T.D.). The pellets were sintered again to increase the sintered density. The sintered densities of pellets re-sintered at 1850 .deg. C for 7 hours ranged from 10.27 to 10.33 g/cm{sup 3}(94.9{approx} 95.5 % of T.D)

  17. Properties of endotracheal tubes reprocessed by two procedures

    Directory of Open Access Journals (Sweden)

    Elisa Elisa

    2011-04-01

    component. SEM analysis detected some large particles and fissures. EDX analysis on the large particles detected sodium and calcium signals. Altogether, signs of contamination and material damage were very strong. Conclusion Both reprocessing methods of reused EITs gave comparable results on sterility and mechanical behavior, but reprocessing may cause decreased surface and matrix quality.

  18. Safeguards instruments for Large-Scale Reprocessing Plants

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, E.A. [Los Alamos National Lab., NM (United States); Case, R.S.; Sonnier, C. [Sandia National Labs., Albuquerque, NM (United States)

    1993-06-01

    Between 1987 and 1992 a multi-national forum known as LASCAR (Large Scale Reprocessing Plant Safeguards) met to assist the IAEA in development of effective and efficient safeguards for large-scale reprocessing plants. The US provided considerable input for safeguards approaches and instrumentation. This paper reviews and updates instrumentation of importance in measuring plutonium and uranium in these facilities.

  19. 9 CFR 114.18 - Reprocessing of biological products.

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Reprocessing of biological products. 114.18 Section 114.18 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION SERVICE... REQUIREMENTS FOR BIOLOGICAL PRODUCTS § 114.18 Reprocessing of biological products. The Administrator...

  20. Fuel-cycle facilities: preliminary safety and environmental information document. Volume VII

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    Information is presented concerning the mining and milling of uranium and thorium; uranium hexafluoride conversion; enrichment; fuel fabrication; reprocessing; storage options; waste disposal options; transportation; heavy-water-production facilities; and international fuel service centers.

  1. Evaluation of thorium based nuclear fuel. Chemical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Konings, R.J.M.; Blankenvoorde, P.J.A.M.; Cordfunke, E.H.P.; Bakker, K.

    1995-07-01

    This report describes the chemical aspects of a thorium-based fuel cycle. It is part of a series devoted to the study of thorium-based fuel as a means to achieve a considerable reduction of the radiotoxicity of the waste from nuclear power production. Therefore special emphasis is placed on fuel (re-)fabrication and fuel reprocessing in the present work. (orig.).

  2. LMFBR fuel cycle studies progress report for August 1972. No. 42

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1972-10-01

    This report continues a series outlining progress in the development of methods for the reprocessing of LMFBR fuels. Development work is reported on problems of irradiated fuel transport to the processing facility, the dissolution of the fuel and the chemical recovery of PuO2-UO2 values, the containment of volatile fission products, product purification, conversion of fuel processing plant product nitrate solutions to solids suitable for shipping and for subsequent fuel fabrication. Pertinent experimental results are presented for the information of those immediately concerned with the field. Detailed description of experimental work and data are included in the topical reports and in the Chemical Technology Division Annual Reports.

  3. Bed models for solid fuel conversion process in grate-fired boilers

    DEFF Research Database (Denmark)

    Costa, M.; Massarotti, N.; Indrizzi, V.

    2013-01-01

    to describe the thermo-chemical conversion process of a solid fuel bed in a grate-fired boiler is presented. In this work both models consider the incoming solid fuel as subjected to drying, pyrolysis, gasification and combustion. In the first approach the biomass bed is treated as a 0D system, but the thermo......Because of the complexity to describe and solve thermo-chemical processes occurring in a fuel bed in grate-fired boiler, it is often necessary to simplify the process and use modeling techniques based on overall mass, energy and species conservation. A comparison between two numerical models......-chemical processes are divided in two successive sections: drying and conversion (which includes pyrolysis, gasification and combustion). The second model is an empirical 1D approach. The two models need input data such as composition, temperature and feeding rate of biomass and primary air. Temperature, species...

  4. Evaluation of Co-precipitation Processes for the Synthesis of Mixed-Oxide Fuel Feedstock Materials

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Emory D [ORNL; Voit, Stewart L [ORNL; Vedder, Raymond James [ORNL

    2011-06-01

    The focus of this report is the evaluation of various co-precipitation processes for use in the synthesis of mixed oxide feedstock powders for the Ceramic Fuels Technology Area within the Fuels Cycle R&D (FCR&D) Program's Advanced Fuels Campaign. The evaluation will include a comparison with standard mechanical mixing of dry powders and as well as other co-conversion methods. The end result will be the down selection of a preferred sequence of co-precipitation process for the preparation of nuclear fuel feedstock materials to be used for comparison with other feedstock preparation methods. A review of the literature was done to identify potential nitrate-to-oxide co-conversion processes which have been applied to mixtures of uranium and plutonium to achieve recycle fuel homogeneity. Recent studies have begun to study the options for co-converting all of the plutonium and neptunium recovered from used nuclear fuels, together with appropriate portions of recovered uranium to produce the desired mixed oxide recycle fuel. The addition of recycled uranium will help reduce the safeguard attractiveness level and improve proliferation resistance of the recycled fuel. The inclusion of neptunium is primarily driven by its chemical similarity to plutonium, thus enabling a simple quick path to recycle. For recycle fuel to thermal-spectrum light water reactors (LWRs), the uranium concentration can be {approx}90% (wt.), and for fast spectrum reactors, the uranium concentration can typically exceed 70% (wt.). However, some of the co-conversion/recycle fuel fabrication processes being developed utilize a two-step process to reach the desired uranium concentration. In these processes, a 50-50 'master-mix' MOX powder is produced by the co-conversion process, and the uranium concentration is adjusted to the desired level for MOX fuel recycle by powder blending (milling) the 'master-mix' with depleted uranium oxide. In general, parameters that must be

  5. Economic Evaluation on the MOX Fuel in the Closed Fuel Cycle

    Directory of Open Access Journals (Sweden)

    Youqi Zheng

    2012-01-01

    Full Text Available The mixed oxide (MOX fuel is one of the most important fuels for the advanced reactors in the future. It is flexible to be applied either in the thermal reactor like pressurized water reactor (PWR or in the fast reactor (FR. This paper compares the two approaches from the view of fuel cost. Two features are involved. (1 The cost of electricity (COE is investigated based on the simulation of realistic operation of a practical PWR power plant and a typical fast breeder reactor design. (2 A new economic analysis model is established, considering the discount rate and the revenue of the reprocessed plutonium besides the traditional costs in the processes of fuel cycle. The sensitivity of COE to the changing parameters is also analyzed. The results show that, in the closed fuel cycle, the fuel cost of applying MOX fuels in the FBR is about 25% lower than that in the PWR at the current operating and fuel cycle level.

  6. Integrated process for the catalytic conversion of biomass-derived syngas into transportation fuels

    Energy Technology Data Exchange (ETDEWEB)

    Dagle, Vanessa Lebarbier; Smith, Colin; Flake, Matthew; Albrecht, Karl O.; Gray, Michel J.; Ramasamy, Karthikeyan K.; Dagle, Robert A.

    2016-01-01

    Efficient synthesis of renewable fuels that will enable cost competitiveness with petroleum-derived fuels remains a grand challenge for U.S. scientists. In this paper, we report on an integrated catalytic approach for producing transportation fuels from biomass-derived syngas. The composition of the resulting hydrocarbon fuel can be modulated to meet specified requirements. Biomass-derived syngas is first converted over an Rh-based catalyst into a complex aqueous mixture of condensable C2+ oxygenated compounds (predominantly ethanol, acetic acid, acetaldehyde, ethyl acetate). This multi-component aqueous mixture then is fed to a second reactor loaded with a ZnxZryOz mixed oxide catalyst, which has tailored acid-base sites, to produce an olefin mixture rich in isobutene. The olefins then are oligomerized using a solid acid catalyst (e.g., Amberlyst-36) to form condensable olefins with molecular weights that can be targeted for gasoline, jet, and/or diesel fuel applications. The product rich in long-chain olefins (C7+) is finally sent to a fourth reactor that is needed for hydrogenation of the olefins into paraffin fuels. Simulated distillation of the hydrotreated oligomerized liquid product indicates that ~75% of the hydrocarbons present are in the jet-fuel range. Process optimization for the oligomerization step could further improve yield to the jet-fuel range. All of these catalytic steps have been demonstrated in sequence, thus providing proof-of-concept for a new integrated process for the production of drop-in biofuels. This unique and flexible process does not require external hydrogen and also could be applied to non-syngas derived feedstock, such as fermentation products (e.g., ethanol, acetic acid, etc.), other oxygenates, and mixtures thereof containing alcohols, acids, aldehydes and/or esters.

  7. Fuel-Flexible Combustion System for Refinery and Chemical Plant Process Heaters

    Energy Technology Data Exchange (ETDEWEB)

    Benson, Charles; Wilson, Robert

    2014-04-30

    This project culminated in the demonstration of a full-scale industrial burner which allows a broad range of “opportunity” gaseous fuels to be cost-effectively and efficiently utilized while generating minimal emissions of criteria air pollutants. The burner is capable of maintaining a stable flame when the fuel composition changes rapidly. This enhanced stability will contribute significantly to improving the safety and reliability of burner operation in manufacturing sites. Process heating in the refining and chemicals sectors is the primary application for this burner. The refining and chemical sectors account for more than 40% of total industrial natural gas use. Prior to the completion of this project, an enabling technology did not exist that would allow these energy-intensive industries to take full advantage of opportunity fuels and thereby reduce their natural gas consumption. Opportunity gaseous fuels include biogas (from animal and agricultural wastes, wastewater plants, and landfills) as well as syngas (from the gasification of biomass, municipal solid wastes, construction wastes, and refinery residuals). The primary challenge to using gaseous opportunity fuels is that their composition and combustion performance differ significantly from those of conventional fuels such as natural gas and refinery fuel gas. An effective fuel-flexible burner must accept fuels that range widely in quality and change in composition over time, often rapidly. In Phase 1 of this project, the team applied computational fluid dynamics analysis to optimize the prototype burner’s aerodynamic, combustion, heat transfer, and emissions performance. In Phase 2, full-scale testing and refinement of two prototype burners were conducted in test furnaces at Zeeco’s offices in Broken Arrow, OK. These tests demonstrated that the full range of conventional and opportunity fuels could be utilized by the project’s burner while achieving robust flame stability and very low levels of

  8. Solar photochemical process engineering for production of fuels and chemicals

    Science.gov (United States)

    Biddle, J. R.; Peterson, D. B.; Fujita, T.

    1985-01-01

    The engineering costs and performance of a nominal 25,000 scmd (883,000 scfd) photochemical plant to produce dihydrogen from water were studied. Two systems were considered, one based on flat-plate collector/reactors and the other on linear parabolic troughs. Engineering subsystems were specified including the collector/reactor, support hardware, field transport piping, gas compression equipment, and balance-of-plant (BOP) items. Overall plant efficiencies of 10.3 and 11.6 percent are estimated for the flat-plate and trough systems, respectively, based on assumed solar photochemical efficiencies of 12.9 and 14.6 percent. Because of the opposing effects of concentration ratio and operating temperature on efficiency, it was concluded that reactor cooling would be necessary with the trough system. Both active and passive cooling methods were considered. Capital costs and energy costs, for both concentrating and non-concentrating systems, were determined and their sensitivity to efficiency and economic parameters were analyzed. The overall plant efficiency is the single most important factor in determining the cost of the fuel.

  9. Novel materials process for alcohol based fuel cells. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hyde, K.; Smith, R.

    2005-07-01

    At present, the unit cost of producing alcohol fuel cells, in particular the cost of the ion-exchange membrane and the platinum catalyst, is limiting the sales. Since the cost of platinum cannot be reduced, an effective means of making the cells more attractive would be to increase the power output per unit area of membrane other than by operating at elevated temperatures. To replace the expensive Nafion, ITM and Cranfield University have developed a new membrane based on ionic hydrophilic polymers. Both acidic and alkaline-based membranes have been produced, the latter may well avoid the use of platinum thus gaining a further cost bonus. Conductivity of the new styrene-sulphonic acid graft membranes is more than double that of Nafion. Similarly, in cross-over tests, the new cells outperformed the Nafion cells. Palladium was investigated as a cheaper alternative to platinum. Based on this study, ITM have applied for five new patents. The study was conducted by ITM Power Plc under contract to the DTI.

  10. Fuel and canister process report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    Werme, Lars; Lilja, Christina (eds.)

    2010-12-15

    This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

  11. /sup 238/Pu fuel form processes quarterly report, April-June 1980

    Energy Technology Data Exchange (ETDEWEB)

    Folger, R. L.

    1980-06-01

    Savannah River Laboratory (SRL) completed the development of a production process to fabricate /sup 238/PuO/sub 2/ fuel forms for the GPHS. The fabrication flowsheet was based on a flowsheet originally developed at Los Alamos National Scientific Laboratory (LANSL). A summary report of the SRL process development effort is presented.

  12. CONCEPTUAL PROCESS DESCRIPTION FOR THE MANUFACTURE OF LOW-ENRICHED URANIUM-MOLYBDENUM FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Daniel M. Wachs; Curtis R. Clark; Randall J. Dunavant

    2008-02-01

    The National Nuclear Security Agency Global Threat Reduction Initiative (GTRI) is tasked with minimizing the use of high-enriched uranium (HEU) worldwide. A key component of that effort is the conversion of research reactors from HEU to low-enriched uranium (LEU) fuels. The GTRI Convert Fuel Development program, previously known as the Reduced Enrichment for Research and Test Reactors program was initiated in 1978 by the United States Department of Energy to develop the nuclear fuels necessary to enable these conversions. The program cooperates with the research reactors’ operators to achieve this goal of HEU to LEU conversion without reduction in reactor performance. The programmatic mandate is to complete the conversion of all civilian domestic research reactors by 2014. These reactors include the five domestic high-performance research reactors (HPRR), namely: the High Flux Isotope Reactor at the Oak Ridge National Laboratory, the Advanced Test Reactor at the Idaho National Laboratory, the National Bureau of Standards Reactor at the National Institute of Standards and Technology, the Missouri University Research Reactor at the University of Missouri–Columbia, and the MIT Reactor-II at the Massachusetts Institute of Technology. Characteristics for each of the HPRRs are given in Appendix A. The GTRI Convert Fuel Development program is currently engaged in the development of a novel nuclear fuel that will enable these conversions. The fuel design is based on a monolithic fuel meat (made from a uranium-molybdenum alloy) clad in Al-6061 that has shown excellent performance in irradiation testing. The unique aspects of the fuel design, however, necessitate the development and implementation of new fabrication techniques and, thus, establishment of the infrastructure to ensure adequate fuel fabrication capability. A conceptual fabrication process description and rough estimates of the total facility throughput are described in this document as a basis for

  13. CFBC evaluation of fuels processed from Illinois coals. Technical report, December 1, 1991--February 29, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Rajan, S.

    1992-08-01

    The main thrust of this research project is the combustion testing and evaluation of two fuels processed from Illinois high sulfur coals. These fuels are (a) flotation slurry fuel beneficiated from coal fines containing 30% and 80% solids, and (b) coal-sorbent pellets made from coal fines using corn starch as a binder. Combustion data from these two fuels are to be compared with corresponding data obtained from a standard coal from the IBCSP coal bank. Parameters to be evaluated are SO{sub 2}, NO{sub x} emissions, combustion efficiency and ash composition, insofar as its influences disposal techniques. During the last quarter, the equipment was serviced and brought on line, and combustion tests were initiated.

  14. NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

    1998-08-01

    An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

  15. Effect of Process Variables During the Head-End Treatment of Spent Oxide Fuel

    Energy Technology Data Exchange (ETDEWEB)

    K.J. Bateman; C.D. Morgan; J.F. Berg; D.J. Brough; P.J. Crane; D.G. Cummings; J.J. Giglio; M.W. Huntley; M.J. Rodriquez; J.D. Sommers; R.P. Lind; D.A. Sell

    2006-08-01

    The development of a head-end processing step for spent oxide fuel that applies to both aqueous and pyrometallurgical technologies is being performed by the Idaho National Laboratory, the Oak Ridge National Laboratory, and the Korean Atomic Energy Research Institute through a joint International Nuclear Energy Research Initiative. The processing step employs high temperatures and oxidative gases to promote the oxidation of UO2 to U3O8. Potential benefits of the head-end step include the removal or reduction of fission products as well as separation of the fuel from cladding. The effects of temperature, pressure, oxidative gas, and cladding have been studied with irradiated spent oxide fuel to determine the optimum conditions for process control. Experiments with temperatures ranging from 500oC to 1250oC have been performed on spent fuel using either air or oxygen gas for the oxidative reaction. Various flowrates and applications have been tested with the oxidative gases to discern the effects on the process. Tests have also been performed under vacuum conditions, following the oxidation cycle, at high temperatures to improve the removal of fission products. The effects of cladding on fission product removal have also been investigated with released fuel under vacuum and high temperature conditions. Results from these experiments will be presented as well as operating conditions based on particle size and decladding characteristics.

  16. Effects of Catalysts on Emissions of Pollutants from Combustion Processes of Liquid Fuels

    Science.gov (United States)

    Bok, Agnieszka; Guziałowska-Tic, Joanna; Tic, Wilhelm Jan

    2014-12-01

    The dynamic growth of the use of non-renewable fuels for energy purposes results in demand for catalysts to improve their combustion process. The paper describes catalysts used mainly in the processes of combustion of motor fuels and fuel oils. These catalysts make it possible to raise the efficiency of oxidation processes simultanously reducing the emission of pollutants. The key to success is the selection of catalyst compounds that will reduce harmful emissions of combustion products into the atmosphere. Catalysts are introduced into the combustion zone in form of solutions miscible with fuel or with air supplied to the combustion process. The following compounds soluble in fuel are inclused in the composition of the described catalysts: organometallic complexes, manganese compounds, salts originated from organic acids, ferrocen and its derivatives and sodium chloride and magnesium chloride responsible for burning the soot (chlorides). The priority is to minimize emissions of volatile organic compounds, nitrogen oxides, sulphur oxides, and carbon monoxide, as well as particulate matter.

  17. Solid recovered fuel production from biodegradable waste in grain processing industry.

    Science.gov (United States)

    Kliopova, Irina; Staniskis, Jurgis Kazimieras; Petraskiene, Violeta

    2013-04-01

    Management of biodegradable waste is one of the most important environmental problems in the grain-processing industry since this waste cannot be dumped anymore due to legal requirements. Biodegradable waste is generated in each stage of grain processing, including the waste-water and air emissions treatment processes. Their management causes some environmental and financial problems. The majority of Lithuanian grain-processing enterprises own and operate composting sites, but in Lithuania the demand for compost is not given. This study focused on the analysis of the possibility of using biodegradable waste for the production of solid recovered fuel, as a local renewable fuel with the purpose of increasing environmental performance and decreasing the direct costs of grain processing. Experimental research with regard to a pilot grain-processing plant has proven that alternative fuel production will lead to minimizing of the volume of biodegradable waste by 75% and the volume of natural gas for heat energy production by 62%. Environmental indicators of grain processing, laboratory analysis of the chemical and physical characteristics of biodegradable waste, mass and energy balances of the solid recovered fuel production, environmental and economical benefits of the project are presented and discussed herein.

  18. Pyro-chemistry within Europart assessment of the studies on spent fuel treatment processes collective work

    Energy Technology Data Exchange (ETDEWEB)

    Bourg, S. [CEA Valrho, 30 - Marcoule (France); Madic, C. [CEA Saclay, 91 - Gif sur Yvette (France); Caravaca, C. [CIEMAT (Spain); Finne, J. [Electricite de France (EDF), 75 - Paris (France); Angelis, G. de [ENEA, Bologna (Italy); Malmbeck, R. [ITU, JRC (Germany); Lewin, B.G. [NEXIA Solutions (United Kingdom); Uhlir, J. [Nuclear Research Institute (NRI) (Czech Republic); Inoue, T. [CRIEPI (Japan); Luca, V. [ANSTO (Australia)

    2007-07-01

    In the continuation of PYROREP (FP5), the pyrochemical domain of EUROPART (FP6) involves today 7 European participants and 2 international partners in order to assess treatment processes which could be used in a broad variety of fuel cycles: molten salt reactor spent fuel, metallic, oxide, carbide or nitride fuels from any reactors. Indeed, a pyrochemical process is not sensitive to radiolysis, and obviously to temperature effects, and could afford to treat fuels after a short cooling period, even at high concentrations. Within EUROPART, the chemistry of actinides and of some fission products was studied in depth in either chloride or fluoride molten salts. Laboratory-scale demonstration experiments were used in order to assess electrolytic processes in molten chlorides and liquid-liquid reductive extraction in molten fluoride/liquid aluminum. An electrolysis process is also under development in molten fluoride. To complete the studies, the development of devices and the modeling of processes are taken into consideration. In order to be fitted for acceptance, a process has to produce minimal quantities of wastes, compatible with storage constraints. Therefore, decontamination techniques, as well as materials suitable for disposal, have to be developed. Fission products precipitation by carbonate and phosphate, and filtration on zeolites were studied. Some original confinement matrices such as sodalite or pollucite were also studied. Integration studies have been carried out in order to evaluate and to compare some selected process flowsheets and possibly to redirect development programs. (authors)

  19. Aqueous processing of U-10Mo scrap for high performance research reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Youker, Amanda J., E-mail: youker@anl.gov [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Stepinski, Dominique C.; Maggos, Laura E.; Bakel, Allen J.; Vandegrift, George F. [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer GTRI program supports conversion from HEU to LEU. Black-Right-Pointing-Pointer High performance research reactors require a dense LEU fuel such as U-10Mo foils. Black-Right-Pointing-Pointer Dissolution conditions for U-10Mo foils in acidic media have been optimized. Black-Right-Pointing-Pointer Solvent-extraction processing can be used to recover U lost in fuel fabrication. Black-Right-Pointing-Pointer Flowsheets were developed using Argonne-design contactors but other contactors can be used as well. - Abstract: The Global Threat Reduction Initiative (GTRI) Conversion program, which is part of the US government's National Nuclear Security Administration (NNSA), supports the conversion of civilian use of highly enriched uranium (HEU) to low enriched uranium (LEU) for reactor fuel and targets. The reason for conversion is to eliminate the use of any material that may pose a threat to the United States or other foreign countries. High performance research reactors (HPRRs) cannot make the conversion to a standard LEU fuel because they require a more dense fuel to meet their performance requirements. As a result, a more dense fuel consisting of a monolithic uranium-molybdenum alloy containing 10% (w/w) Mo with Al cladding and a Zr bonding-layer is being considered. Significant losses are expected in the fabrication of this fuel, so a means to recycle the scrap pieces is needed. Argonne National Laboratory has developed an aqueous-processing flowsheet for scrap recovery in the fuel fabrication process for high-density LEU-monolithic fuel based on data found in the literature. Experiments have been performed to investigate dissolution conditions for solutions containing approximately 20 g-U/L and 50 g-U/L with and without Fe(NO{sub 3}){sub 3}. HNO{sub 3} and HF concentrations have been optimized for timely dissolution of the fuel scrap and prevention of the formation of the U-Zr{sub 2} intermetallic, explosive complex, while

  20. Low Emissions Burner Technology for Metal Processing Industry using Byproducts and Biomass Derived Liquid Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Agrawal, Ajay; Taylor, Robert

    2013-09-30

    path forward to utilize both fossil and alternative liquid fuels in the same combustion system. In particular, experiments show that straight VO can be cleanly combusted without the need for chemical processing or preheating steps, which can result in significant economic and environmental benefits. Next, low-emission combustion of glycerol/methane was achieved by utilizing FB injector to yield fine droplets of highly viscous glycerol. Heat released from methane combustion further improves glycerol pre-vaporization and thus its clean combustion. Methane addition results in an intensified reaction zone with locally high temperatures near the injector exit. Reduction in methane flow rate elongates the reaction zone, which leads to higher CO emissions and lower NOx emissions. Similarly, higher air to liquid (ALR) mass ratio improves atomization and fuel pre-vaporization and shifts the flame closer to the injector exit. In spite of these internal variations, all fuel mixes of glycerol with methane produced similar CO and NOx emissions at the combustor exit. Results show that FB concept provides low emissions with the flexibility to utilize gaseous and highly viscous liquid fuels, straight VO and glycerol, without preheating or preprocessing the fuels. Following these initial experiments in quartz combustor, we demonstrated that glycerol combustion can be stably sustained in a metal combustor. Phase Doppler Particle Analyzer (PDPA) measurements in glycerol/methane flames resulted in flow-weighted Sauter Mean Diameter (SMD) of 35 to 40 μm, depending upon the methane percentage. This study verified that lab-scale dual-fuel burner using FB injector can successfully atomize and combust glycerol and presumably other highly viscous liquid fuels at relatively low HRR (<10 kW). For industrial applications, a scaled-up glycerol burner design thus seemed feasible.

  1. Preliminary evaluation of alternative waste form solidification processes. Volume I. Identification of the processes.

    Energy Technology Data Exchange (ETDEWEB)

    Treat, R.L.; Nesbitt, J.F.; Blair, H.T.; Carter, J.G.; Gorton, P.S.; Partain, W.L.; Timmerman, C.L.

    1980-04-01

    This document contains preconceptual design data on 11 processes for the solidification and isolation of nuclear high-level liquid wastes (HLLW). The processes are: in-can glass melting (ICGM) process, joule-heated glass melting (JHGM) process, glass-ceramic (GC) process, marbles-in-lead (MIL) matrix process, supercalcine pellets-in-metal (SCPIM) matrix process, pyrolytic-carbon coated pellets-in-metal (PCCPIM) matrix process, supercalcine hot-isostatic-pressing (SCHIP) process, SYNROC hot-isostatic-pressing (SYNROC HIP) process, titanate process, concrete process, and cermet process. For the purposes of this study, it was assumed that each of the solidification processes is capable of handling similar amounts of HLLW generated in a production-sized fuel reprocessing plant. It was also assumed that each of the processes would be enclosed in a shielded canyon or cells within a waste facility located at the fuel reprocessing plant. Finally, it was assumed that all of the processes would be subject to the same set of regulations, codes and standards. Each of the solidification processes converts waste into forms that may be acceptable for geological disposal. Each process begins with the receipt of HLLW from the fuel reprocessing plant. In this study, it was assumed that the original composition of the HLLW would be the same for each process. The process ends when the different waste forms are enclosed in canisters or containers that are acceptable for interim storage. Overviews of each of the 11 processes and the bases used for their identification are presented in the first part of this report. Each process, including its equipment and its requirements, is covered in more detail in Appendices A through K. Pertinent information on the current state of the art and the research and development required for the implementation of each process are also noted in the appendices.

  2. A Green Approach to SNF Reprocessing: Are Common Household Reagents the Answer?

    Energy Technology Data Exchange (ETDEWEB)

    Peper, Shane M.; McNamara, Bruce K.; O' Hara, Matthew J.; Douglas, Matthew

    2008-04-03

    It has been discovered that UO2, the principal component of spent nuclear fuel (SNF), can efficiently be dissolved at room temperature using a combination of common household reagents, namely hydrogen peroxide, baking soda, and ammonia. This rather serendipitous discovery opens up the possibility, for the first time, of considering a non-acidic process for recycling U from SNF. Albeit at the early stages of development, our unconventional dissolution approach possesses many attractive features that could make it a reality in the future. With dissolution byproducts of water and oxygen, our approach poses a minimal threat to the environment. Moreover, the use of common household reagents to afford actinide oxide dissolution suggests a certain degree of economic favorability. With the use of a “closed” digestion vessel as a reaction chamber, our approach has substantial versatility with the option of using either aqueous or gaseous reactant feeds or a combination of both. Our approach distinguishes itself from all existing reprocessing technologies in two important ways. First and foremost, it is an alkaline rather than an acidic process, using mild non-corrosive chemicals under ambient conditions to effect actinide separations. Secondly, it does not dissolve the entire SNF matrix, but rather selectively solubilizes U and other light actinides for subsequent separation, resulting in potentially faster head-end dissolution and fewer downstream separation steps. From a safeguards perspective, the use of oxidizing alkaline solutions to effect actinide separations also potentially offers a degree of inherent proliferation resistance, by allowing the U to be selectively removed from the remaining dissolver solution while keeping Pu grouped with the other minor actinides and fission products. This paper will describe the design and general experimental setup of a “closed” digestion vessel for performing uranium oxide dissolutions under alkaline conditions using

  3. Biorefineries to integrate fuel, energy and chemical production processes

    Directory of Open Access Journals (Sweden)

    Enrica Bargiacchi

    2007-12-01

    Full Text Available The world of renewable energies is in fast evolution and arouses political and public interests, especially as an opportunity to boost environmental sustainability by mitigation of greenhouse gas emissions. This work aims at examining the possibilities related to the development of biorefineries, where biomass conversion processes to produce biofuels, electricity and biochemicals are integrated. Particular interest is given to the production processes of biodiesel, bioethanol and biogas, for which present world situation, problems, and perspectives are drawn. Potential areas for agronomic and biotech researches are also discussed. Producing biomass for biorefinery processing will eventually lead to maximize yields, in the non food agriculture.

  4. Sol-gel process preparation and evaluation of the analytical performances of an hydrazine specific chemical sensor; Preparation par procede sol-gel et evaluation des performances analytiques d`un capteur chimique specifique de l`hydrazine

    Energy Technology Data Exchange (ETDEWEB)

    Gojon, C

    1996-12-01

    The realisation of optical fibers active chemical collector to analyze hydrazine in line, in the spent fuel reprocessing process is the subject of this work. The p.dimethyl-amino-benzaldehyde has been chosen as reagent for its chemical and optical properties. 186 refs.

  5. Catalysts and process for liquid hydrocarbon fuel production

    Energy Technology Data Exchange (ETDEWEB)

    White, Mark G.; Ranaweera, Samantha A.; Henry, William P.

    2016-08-02

    The present invention provides a novel process and system in which a mixture of carbon monoxide and hydrogen synthesis gas, or syngas, is converted into hydrocarbon mixtures composed of high quality distillates, gasoline components, and lower molecular weight gaseous olefins in one reactor or step. The invention utilizes a novel supported bimetallic ion complex catalyst for conversion, and provides methods of preparing such novel catalysts and use of the novel catalysts in the process and system of the invention.

  6. Treatment of dysfunctionally stored experiences with the method Eye Movement Desensitization and Reprocessing – EMDR

    OpenAIRE

    Robert Cvetek

    2002-01-01

    In this paper a new therapeutic method called EMDR (Eye Movement Desensitization and Reprocessing) is described. The method was formed mainly for treatment of posttraumatic stress disorder, but there are also some reports about success with other mental disorders. The theoretical base of EMDR and especially the accelerated information processing model, the concept of memory networks and the explanations of effects of eye movements are presented. The process of EMDR is also described.

  7. Treatment of dysfunctionally stored experiences with the method Eye Movement Desensitization and Reprocessing – EMDR

    Directory of Open Access Journals (Sweden)

    Robert Cvetek

    2002-09-01

    Full Text Available In this paper a new therapeutic method called EMDR (Eye Movement Desensitization and Reprocessing is described. The method was formed mainly for treatment of posttraumatic stress disorder, but there are also some reports about success with other mental disorders. The theoretical base of EMDR and especially the accelerated information processing model, the concept of memory networks and the explanations of effects of eye movements are presented. The process of EMDR is also described.

  8. Alternative fuel vehicles for the state fleets: Results of the 5-year planning process

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-01

    This report documents the first attempt by the Department of Energy (DOE) to work with states to prepare five-year Alternative Fuel Vehicle (AFV) acquisition plans to identify alternative fuels and vehicles that they are planning on or would like to acquire. The DOE Regional Support Offices (RSOs) met with representatives from the states in their regions and assisted in the preparation of the plans. These plans will be used in conjunction with previously gathered Federal five-year plans to encourage Original Equipment Manufacturers (OEMs) to expand the variety of AFVs produced, reduce the incremental cost of AFVs, and to encourage fuel suppliers to expand the alternative fuel infrastructure and alternative fuel availability. By identifying the needs and requirements of state fleets, DOE can begin to describe the specific nature of the future state fleets, and establish a defined market for OEMs and fuel suppliers. DOE initiated the development and collection of the state five-year plans before the signing of the Energy Policy Act, to raise the awareness of states that they will be required by law to acquire AFVs. As a result, several states that had no AFV acquisition plan when queried have developed or are in the process of developing plans. The DOE and its RSOs are still working with the states to develop and refine acquisition plans, and this report should be treated as documentation of work in progress.

  9. Catalysts and process for liquid hydrocarbon fuel production

    Science.gov (United States)

    White, Mark G; Liu, Shetian

    2014-12-09

    The present invention provides a novel process and system in which a mixture of carbon monoxide and hydrogen synthesis gas, or syngas, is converted into hydrocarbon mixtures composed of high quality gasoline components, aromatic compounds, and lower molecular weight gaseous olefins in one reactor or step. The invention utilizes a novel molybdenum-zeolite catalyst in high pressure hydrogen for conversion, as well as a novel rhenium-zeolite catalyst in place of the molybdenum-zeolite catalyst, and provides for use of the novel catalysts in the process and system of the invention.

  10. Fuel decomposition and boundary-layer combustion processes of hybrid rocket motors

    Science.gov (United States)

    Chiaverini, Martin J.; Harting, George C.; Lu, Yeu-Cherng; Kuo, Kenneth K.; Serin, Nadir; Johnson, David K.

    1995-01-01

    Using a high-pressure, two-dimensional hybrid motor, an experimental investigation was conducted on fundamental processes involved in hybrid rocket combustion. HTPB (Hydroxyl-terminated Polybutadiene) fuel cross-linked with diisocyanate was burned with GOX under various operating conditions. Large-amplitude pressure oscillations were encountered in earlier test runs. After identifying the source of instability and decoupling the GOX feed-line system and combustion chamber, the pressure oscillations were drastically reduced from +/-20% of the localized mean pressure to an acceptable range of +/-1.5% Embedded fine-wire thermocouples indicated that the surface temperature of the burning fuel was around 1000 K depending upon axial locations and operating conditions. Also, except near the leading-edge region, the subsurface thermal wave profiles in the upstream locations are thicker than those in the downstream locations since the solid-fuel regression rate, in general, increases with distance along the fuel slab. The recovered solid fuel slabs in the laminar portion of the boundary layer exhibited smooth surfaces, indicating the existence of a liquid melt layer on the burning fuel surface in the upstream region. After the transition section, which displayed distinct transverse striations, the surface roughness pattern became quite random and very pronounced in the downstream turbulent boundary-layer region. Both real-time X-ray radiography and ultrasonic pulse-echo techniques were used to determine the instantaneous web thickness burned and instantaneous solid-fuel regression rates over certain portions of the fuel slabs. Globally averaged and axially dependent but time-averaged regression rates were also obtained and presented.

  11. Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Lippek, H.E.; Schuller, C.R.

    1979-03-01

    A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light most of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.

  12. Microchemical Systems for Fuel Processing and Conversion to Electrical Power

    Science.gov (United States)

    2007-03-15

    6.8 Vacuum Packaging ........................................................................................................................................43...production may take place. Vacuum packaging and reflective shields, required to maximize the thermal isolation of the reaction zone, have been tested...microfabrication strategies. Substantial effort has been expended on the chip level vacuum packaging process, which is critical to the performance of the

  13. EVALUATION OF A PROCESS TO CONVERT BIOMASS TO METHANOL FUEL

    Science.gov (United States)

    The report gives results of a review of the design of a reactor capable of gasifying approximately 50 lb/hr of biomass for a pilot-scale facility to develop, demonstrate, and evaluate the Hynol Process, a high-temperature, high-pressure method for converting biomass into methanol...

  14. INNOVATIVE FRESH WATER PRODUCTION PROCESS FOR FOSSIL FUEL PLANTS

    Energy Technology Data Exchange (ETDEWEB)

    James F. Klausner; Renwei Mei; Yi Li; Jessica Knight

    2004-09-01

    An innovative Diffusion Driven Desalination (DDD) process was recently described where evaporation of mineralized water is driven by diffusion within a packed bed. The energy source to drive the process is derived from low pressure condensing steam within the main condenser of a steam power generating plant. Since waste heat is used to drive the process, the main cost of fresh water production is attributed to the energy cost of pumping air and water through the packed bed. This report describes the annual progress made in the development and analysis of a Diffusion Driven Desalination (DDD) system. A combined thermodynamic and dynamic analysis demonstrates that the DDD process can yield a fresh water production of 1.03 million gallon/day by utilizing waste heat from a 100 MW steam power plant based on a condensing steam pressure of only 3'' Hg. Throughout the past year, the main focus of the desalination process has been on the diffusion tower and direct contact condenser. Detailed heat and mass transfer analyses required to size and analyze these heat and mass transfer devices are described. An experimental DDD facility has been fabricated, and temperature and humidity data have been collected over a range of flow and thermal conditions. The analyses agree quite well with the current data and the information available in the literature. Direct contact condensers with and without packing have been investigated. It has been experimentally observed that the fresh water production rate is significantly enhanced when packing is added to the direct contact condensers.

  15. Reducing Actinide Production Using Inert Matrix Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Deinert, Mark [Colorado School of Mines, Golden, CO (United States)

    2017-08-23

    The environmental and geopolitical problems that surround nuclear power stem largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent nuclear fuel. New methods for transmuting these elements into more benign forms are needed. Current research efforts focus largely on the development of fast burner reactors, because it has been shown that they could dramatically reduce the accumulation of transuranics. However, despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical limitation to this, and other such strategies, is that they require a type of spent fuel reprocessing that can efficiently separate all of the transuranics from the fission products with which they are mixed. Unfortunately, the technology for doing this on an industrial scale is still in development. In this project, we explore a strategy for transmutation that can be deployed using existing, current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to recycle transuranics in a conventional pressurized water reactor could reduce overall production of these materials by an amount that is similar to what is achievable using proposed fast reactor cycles. Furthermore, we show that these transuranic reductions can be achieved even if the fission products are carried into the inert matrix fuel along with the transuranics, bypassing the critical separations hurdle described above. The implications of these findings are significant, because they imply that inert matrix fuel could be made directly from the material streams produced by the commercially available PUREX process. Zirconium dioxide would be an ideal choice of inert matrix in this context because it is known to form a stable solid solution with both fission products and transuranics.

  16. JAEA key facilities for global advanced fuel cycle R and D

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Shigeo; Yamamoto, Ryuichi [Nuclear Fuel Cycle Engineering Labos, JAEA, 4-33 Tokai-mura, Ibaraki, 319-1194 (Japan)

    2008-07-01

    Advanced fuel cycle will be realized with the mid and long term R and D during the long-term transition period from LWR cycle to advanced reactor fuel cycle. Most of JAEA facilities have been utilized to establish the current LWR and FBR (Fast Breeder Reactor) fuel cycle by implementing evolutionary R and D. An assessment of today's state experimental facilities concerning the following research issues: reprocessing, Mox fuel fabrication, irradiation and post-irradiation examination, waste management and nuclear data measurement, is made. The revolutionary R and D requests new issues to be studied: the TRU multi-recycling, minor actinide recycling, the assessment of proliferation resistance and the assessment of cost reduction. To implement the revolutionary R and D for advanced fuel cycle, however, these facilities should be refurbished to install new machines and process equipment to provide more flexible testing parameters.

  17. Innovative Fresh Water Production Process for Fossil Fuel Plants

    Energy Technology Data Exchange (ETDEWEB)

    James F. Klausner; Renwei Mei; Yi Li; Jessica Knight; Venugopal Jogi

    2005-09-01

    This project concerns a diffusion driven desalination (DDD) process where warm water is evaporated into a low humidity air stream, and the vapor is condensed out to produce distilled water. Although the process has a low fresh water to feed water conversion efficiency, it has been demonstrated that this process can potentially produce low cost distilled water when driven by low grade waste heat. This report describes the annual progress made in the development and analysis of a Diffusion Driven Desalination (DDD) system. A dynamic analysis of heat and mass transfer demonstrates that the DDD process can yield a fresh water production of 1.03 million gallon/day by utilizing waste heat from a 100 MW steam power plant based on a condensing steam pressure of only 3 Hg. The optimum operating condition for the DDD process with a high temperature of 50 C and sink temperature of 25 C has an air mass flux of 1.5 kg/m{sup 2}-s, air to feed water mass flow ratio of 1 in the diffusion tower, and a fresh water to air mass flow ratio of 2 in the condenser. Operating at these conditions yields a fresh water production efficiency (m{sub fW}/m{sub L}) of 0.031 and electric energy consumption rate of 0.0023 kW-hr/kg{sub fW}. Throughout the past year, the main focus of the desalination process has been on the direct contact condenser. Detailed heat and mass transfer analyses required to size and analyze these heat and mass transfer devices are described. The analyses agree quite well with the current data. Recently, it has been recognized that the fresh water production efficiency can be significantly enhanced with air heating. This type of configuration is well suited for power plants utilizing air-cooled condensers. The experimental DDD facility has been modified with an air heating section, and temperature and humidity data have been collected over a range of flow and thermal conditions. It has been experimentally observed that the fresh water production rate is enhanced when air

  18. Innovative Fresh Water Production Process for Fossil Fuel Plants

    Energy Technology Data Exchange (ETDEWEB)

    James F. Klausner; Renwei Mei; Yi Li; Jessica Knight

    2006-09-29

    This project concerns a diffusion driven desalination (DDD) process where warm water is evaporated into a low humidity air stream, and the vapor is condensed out to produce distilled water. Although the process has a low fresh water to feed water conversion efficiency, it has been demonstrated that this process can potentially produce low cost distilled water when driven by low grade waste heat. This report summarizes the progress made in the development and analysis of a Diffusion Driven Desalination (DDD) system. Detailed heat and mass transfer analyses required to size and analyze the diffusion tower using a heated water input are described. The analyses agree quite well with the current data and the information available in the literature. The direct contact condenser has also been thoroughly analyzed and the system performance at optimal operating conditions has been considered using a heated water/ambient air input to the diffusion tower. The diffusion tower has also been analyzed using a heated air input. The DDD laboratory facility has successfully been modified to include an air heating section. Experiments have been conducted over a range of parameters for two different cases: heated air/heated water and heated air/ambient water. A theoretical heat and mass transfer model has been examined for both of these cases and agreement between the experimental and theoretical data is good. A parametric study reveals that for every liquid mass flux there is an air mass flux value where the diffusion tower energy consumption is minimal and an air mass flux where the fresh water production flux is maximized. A study was also performed to compare the DDD process with different inlet operating conditions as well as different packing. It is shown that the heated air/heated water case is more capable of greater fresh water production with the same energy consumption than the ambient air/heated water process at high liquid mass flux. It is also shown that there can be

  19. Solvent degradation products in nuclear fuel processing solvents

    Energy Technology Data Exchange (ETDEWEB)

    Shook, H.E. Jr.

    1988-06-01

    The Savannah River Plant uses a modified Purex process to recover enriched uranium and separate fission products. This process uses 7.5% tri-n-butyl phosphate (TBP) dissolved in normal paraffin hydrocarbons for the solvent extraction of a nitric acid solution containing the materials to be separated. Periodic problems in product decontamination result from solvent degradation. A study to improve process efficiency has identified certain solvent degradation products and suggested mitigation measures. Undecanoic acid, lauric acid, and tridecanoic acid were tentatively identified as diluent degradation products in recycle solvent. These long-chain organic acids affect phase separation and lead to low decontamination factors. Solid phase extraction (SPE) was used to concentrate the organic acids in solvent prior to analysis by high performance liquid chromatography (HPLC). SPE and HPLC methods were optimized in this work for analysis of decanoic acid, undecanoic acid, and lauric acid in solvent. Accelerated solvent degradation studies with 7.5% TBP in normal paraffin hydrocarbons showed that long-chain organic acids and long-chain alkyl butyl phosphoric acids are formed by reactions with nitric acid. Degradation of both tributyl phosphate and hydrocarbon can be minimized with purified normal paraffin replacing the standard grade presently used. 12 refs., 1 fig., 3 tabs.

  20. INNOVATIVE FRESH WATER PRODUCTION PROCESS FOR FOSSIL FUEL PLANTS

    Energy Technology Data Exchange (ETDEWEB)

    James F. Klausner; Renwei Mei; Yi Li; Mohamed Darwish; Diego Acevedo; Jessica Knight

    2003-09-01

    This report describes the annual progress made in the development and analysis of a Diffusion Driven Desalination (DDD) system, which is powered by the waste heat from low pressure condensing steam in power plants. The desalination is driven by water vapor saturating dry air flowing through a diffusion tower. Liquid water is condensed out of the air/vapor mixture in a direct contact condenser. A thermodynamic analysis demonstrates that the DDD process can yield a fresh water production efficiency of 4.5% based on a feed water inlet temperature of only 50 C. An example is discussed in which the DDD process utilizes waste heat from a 100 MW steam power plant to produce 1.51 million gallons of fresh water per day. The main focus of the initial development of the desalination process has been on the diffusion tower. A detailed mathematical model for the diffusion tower has been described, and its numerical implementation has been used to characterize its performance and provide guidance for design. The analysis has been used to design a laboratory scale diffusion tower, which has been thoroughly instrumented to allow detailed measurements of heat and mass transfer coefficient, as well as fresh water production efficiency. The experimental facility has been described in detail.

  1. INNOVATIVE FRESH WATER PRODUCTION PROCESS FOR FOSSIL FUEL PLANTS

    Energy Technology Data Exchange (ETDEWEB)

    James F. Klausner; Renwei Mei; Yi Li; Mohamed Darwish; Diego Acevedo; Jessica Knight

    2003-09-01

    This report describes the annual progress made in the development and analysis of a Diffusion Driven Desalination (DDD) system, which is powered by the waste heat from low pressure condensing steam in power plants. The desalination is driven by water vapor saturating dry air flowing through a diffusion tower. Liquid water is condensed out of the air/vapor mixture in a direct contact condenser. A thermodynamic analysis demonstrates that the DDD process can yield a fresh water production efficiency of 4.5% based on a feed water inlet temperature of only 50 C. An example is discussed in which the DDD process utilizes waste heat from a 100 MW steam power plant to produce 1.51 million gallons of fresh water per day. The main focus of the initial development of the desalination process has been on the diffusion tower. A detailed mathematical model for the diffusion tower has been described, and its numerical implementation has been used to characterize its performance and provide guidance for design. The analysis has been used to design a laboratory scale diffusion tower, which has been thoroughly instrumented to allow detailed measurements of heat and mass transfer coefficient, as well as fresh water production efficiency. The experimental facility has been described in detail.

  2. Chemical Processing of Non-Crop Plants for Jet Fuel Blends Production

    Science.gov (United States)

    Kulis, M. J.; Hepp, A. F.; McDowell, M.; Ribita, D.

    2009-01-01

    The use of Biofuels has been gaining in popularity over the past few years due to their ability to reduce the dependence on fossil fuels. Biofuels as a renewable energy source can be a viable option for sustaining long-term energy needs if they are managed efficiently. We describe our initial efforts to exploit algae, halophytes and other non-crop plants to produce synthetics for fuel blends that can potentially be used as fuels for aviation and non-aerospace applications. Our efforts have been dedicated to crafting efficient extraction and refining processes in order to extract constituents from the plant materials with the ultimate goal of determining the feasibility of producing biomass-based jet fuel from the refined extract. Two extraction methods have been developed based on communition processes, and liquid-solid extraction techniques. Refining procedures such as chlorophyll removal and transesterification of triglycerides have been performed. Gas chromatography in tandem with mass spectroscopy is currently being utilized in order to qualitatively determine the individual components of the refined extract. We also briefly discuss and compare alternative methods to extract fuel-blending agents from alternative biofuels sources.

  3. /sup 238/Pu fuel form processes. Quarterly report, July-September 1981

    Energy Technology Data Exchange (ETDEWEB)

    1982-04-01

    This report is one of a series to summarize progress in the Savannah River Laboratory /sup 238/Pu Fuel Form Program. Goals of the Savannah River Laboratory (SRL) program include providing technical support for the production of /sup 238/PuO/sub 2/ fuel forms in the Savannah River Plant's (SRP) Plutonium Fuel Form (PuFF) Facility. This part of the program includes: demonstration of processes and techniques developed by the Los Alamos National Laboratory (LANL) for production at SRP. Information from the demonstration will provide the technical data for technical standards and operating procedures; technical support to assist plant startup and to ensure continuation of safe and efficient production of high-quality heat-source fuel; and technical assistance after startup to accommodate changes in product and product specifications, to assist user agencies in improving product performance, to assist SRP in making process improvements that increase efficiency and product reliability, and to adapt plant facilities for new products. This report summarizes progress in improving the quality of welds in /sup 238/PuO/sub 2/ fuel capsules.

  4. Designing and optimization of a micro CHP system based on Solid Oxide Fuel Cell with different fuel processing technologies

    DEFF Research Database (Denmark)

    Liso, Vincenzo; Nielsen, Mads Pagh; Kær, Søren Knudsen

    2009-01-01

    of the Micro Combined Heat and Power plant (mCHP) will be identified including fuel and air supply, fuel management anode re-circulation, exhaust gas heat management, power conditioning and control system. Using mass and energy balance, different types of fuel reforming including steam reforming...

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

    Science.gov (United States)

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

    2015-12-01

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

  6. Kiln process impact of alternative solid fuel combustion in the cement kiln main burner - Mathematical modelling and full-scale experiment

    OpenAIRE

    Ariyaratne, Hiromi Wijesinghe; Melaaen, Morten Christian; Tokheim, Lars André; Manjula, Edirisinghe V. P. J.

    2014-01-01

    Increased use of alternative fuels in cement kilns is a trend in the world. However, replacing fossil fuels like coal with different alternative fuels will give various impacts on the overall kiln process due to the fuel characteristics. Hence, it is important to know to what extent the fossil fuels can be replaced by different alternative fuels without severely changing process conditions, product quality or emissions. In the present study, a mass and energy balance for the combustion of dif...

  7. Fuel pellets from biomass - Processing, bonding, raw materials

    Energy Technology Data Exchange (ETDEWEB)

    Stelte, W.

    2011-12-15

    The present study investigates several important aspects of biomass pelletization. Seven individual studies have been conducted and linked together, in order to push forward the research frontier of biomass pelletization processes. The first study was to investigate influence of the different processing parameters on the pressure built up in the press channel of a pellet mill. It showed that the major factor was the press channel length as well as temperature, moisture content, particle size and extractive content. Furthermore, extractive migration to the pellet surface at an elevated temperature played an important role. The second study presented a method of how key processing parameters can be estimated, based on a pellet model and a small number of fast and simple laboratory trials using a single pellet press. The third study investigated the bonding mechanisms within a biomass pellet, which indicate that different mechanisms are involved depending on biomass type and pelletizing conditions. Interpenetration of polymer chains and close intermolecular distance resulting in better secondary bonding were assumed to be the key factors for high mechanical properties of the formed pellets. The outcome of this study resulted in study four and five investigating the role of lignin glass transition for biomass pelletization. It was demonstrated that the softening temperature of lignin was dependent on species and moisture content. In typical processing conditions and at 8% (wt) moisture content, transitions were identified to be at approximately 53-63 deg. C for wheat straw and about 91 deg. C for spruce lignin. Furthermore, the effects of wheat straw extractives on the pelletizing properties and pellet stability were investigated. The sixth and seventh study applied the developed methodology to test the pelletizing properties of thermally pre-treated (torrefied) biomass from spruce and wheat straw. The results indicated that high torrefaction temperatures above 275 deg

  8. TAPE CALENDERING MANUFACTURING PROCESS FOR MULTILAYER THIN-FILM SOLID OXIDE FUEL CELLS

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen Minh; Kurt Montgomery

    2004-10-01

    This report summarizes the work performed by Hybrid Power Generation Systems, LLC during the Phases I and II under Contract DE-AC26-00NT40705 for the U. S. Department of Energy, National Energy Technology Laboratory (DOE/NETL) entitled ''Tape Calendering Manufacturing Process For Multilayer Thin-Film Solid Oxide Fuel Cells''. The main objective of this project was to develop the manufacturing process based on tape calendering for multilayer solid oxide fuel cells (SOFC's) using the unitized cell design concept and to demonstrate cell performance under specified operating conditions. Summarized in this report is the development and improvements to multilayer SOFC cells and the unitized cell design. Improvements to the multilayer SOFC cell were made in electrochemical performance, in both the anode and cathode, with cells demonstrating power densities of nearly 0.9 W/cm{sup 2} for 650 C operation and other cell configurations showing greater than 1.0 W/cm{sup 2} at 75% fuel utilization and 800 C. The unitized cell design was matured through design, analysis and development testing to a point that cell operation at greater than 70% fuel utilization was demonstrated at 800 C. The manufacturing process for both the multilayer cell and unitized cell design were assessed and refined, process maps were developed, forming approaches explored, and nondestructive evaluation (NDE) techniques examined.

  9. Integrating fuel treatment into ecosystem management: A proposed project planning process

    Science.gov (United States)

    Keith D. Stockmann; Kevin D. Hyde; J. Greg Jones; Dan R. Loeffler; Robin P. Silverstein

    2010-01-01

    Concern over increased wildland fire threats on public lands throughout the western United States makes fuel reduction activities the primary driver of many management projects. This single-issue focus recalls a management planning process practiced frequently in recent decades - a least-harm approach where the primary objective is first addressed and then plans are...

  10. Hot Experiment on Fission Gas Release Behavior from Voloxidation Process using Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Geun Il; Park, J. J.; Jung, I. H.; Shin, J. M.; Cho, K. H.; Yang, M. S.; Song, K. C

    2007-08-15

    Quantitative analysis of the fission gas release characteristics during the voloxidation and OREOX processes of spent PWR fuel was carried out by spent PWR fuel in a hot-cell of the DFDF. The release characteris