Science.gov

Sample records for fuel reprocessing process

  1. Process for recovery of palladium from nuclear fuel reprocessing wastes

    DOEpatents

    Campbell, David O.; Buxton, Samuel R.

    1981-01-01

    Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M, (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound, (c) heating the solution at reflux temperature until precipitation is complete, and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

  2. Process for recovery of palladium from nuclear fuel reprocessing wastes

    DOEpatents

    Campbell, D.O.; Buxton, S.R.

    1980-06-16

    Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M; (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound; (c) heating the solution at reflux temperature until precipitation is complete; and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

  3. Actinide partitioning processes for fuel reprocessing and refabrication plant wastes

    SciTech Connect

    Finney, B.C.; Tedder, D.W.

    1980-01-01

    Chemical processing methods have been developed on a laboratory scale to partition the actinides from the liquid and solid fuel reprocessing plant (FRP) and refabrication plant (FFP) wastes. It was envisioned that these processes would be incorporated into separate waste treatment facilities (WTFs) that are adjacent to, but not integrated with, the fuel reprocessing and refabrication plants. Engineering equipment and material balance flowsheets have been developed for WTFs in support of a 2000-MTHM/year FRP and a 660-MTHM/year MOX-FFP. The processing subsystems incorporated in the FRP-WTF are: High-Level Solid Waste Treatment, High-Level Liquid Waste Treatment, Solid Alpha Waste Treatment, Cation Exchange Chromatography, Salt Waste Treatment, Actinide Recovery, Solvent Cleanup and recycle, Off-Gas Treatment, Actinide Product Concentration, and Acid and Water Recycle. The WTF supporting a fuel refabrication facility, although similar, does not contain subsystems (1) and (2). Based on the results of the laboratory and hot-cell experimental work, we believe that the processes and flowsheets offer the potential to reduce the total unrecovered actinides in FRP and FFP wastes to less than or equal to 0.25%. The actinide partitioning processes and the WTF concept represent advanced technology that would require substantial work before commercialization. It is estimated that an orderly development program would require 15 to 20 years to complete and would cost about 700 million 1979 dollars. It is estimated that the capital cost and annual operating cost, in mid-1979 dollars, for the FRP-WTF are $1035 million and $71.5 million/year, and for the FFP-WTF are $436 million and $25.6 million/year, respectively.

  4. Reprocessing RERTR silicide fuels

    SciTech Connect

    Rodrigues, G.C.; Gouge, A.P.

    1983-05-01

    The Reduced Enrichment Research and Test Reactor Program is one element of the United States Government's nonproliferation effort. High-density, low-enrichment, aluminum-clad uranium silicide fuels may be substituted for the highly enriched aluminum-clad alloy fuels now in use. Savannah River Laboratory has performed studies which demonstrate reprocessability of spent RERTR silicide fuels at Savannah River Plant. Results of dissolution and feed preparation tests and solvent extraction processing demonstrations with both unirradiated and irradiated uranium silicide fuels are presented.

  5. Nuclear Fuel Reprocessing

    SciTech Connect

    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

  6. Demonstrations of safeguards process monitoring sensitivities. Consolidated Fuel Reprocessing Program

    SciTech Connect

    Ehinger, M.H.; Wachter, J.W.

    1986-09-01

    Can process-monitoring information be incorporated into safeguards tests. What level of sensitivity to removals of materials can be achieved with process monitoring tests. These questions are being answered by a series of tests in US facilities. These tests involve full-scale facilities that simulate operating reprocessing plant conditions with natural or depleted uranium solutions as surrogate feed materials. Safeguards systems are in place to detect loss or unauthorized removals of solutions. As part of the tests, actual removals of material from the operating facilities are made. Removals have ranged from several kilograms down to a few hundred grams of uranium. For purposes of the tests, uranium is considered to be plutonium and is the focus of safeguards concerns.

  7. Nuclear Fuel Reprocessing

    SciTech Connect

    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.

  8. Head-end process for the reprocessing of HTGR spent fuel

    SciTech Connect

    Chen, J.; Wen, M.

    2013-07-01

    The reprocessing of HTGR spent fuels is in favor of the sustainable development of nuclear energy to realize the maximal use of nuclear resource and the minimum disposal of nuclear waste. The head-end of HTGR spent fuels reprocessing is different from that of the LWR spent fuels reprocessing because of the difference of spent fuel structure. The dismantling of the graphite spent fuel element and the highly effective dissolution of fuel kernel is the most difficult process in the head end of the reprocessing. Recently, some work on the head-end has been done in China. First, the electrochemical method with nitrate salt as electrolyte was studied to disintegrate the graphite matrix from HTGR fuel elements and release the coated fuel particles, to provide an option for the head-end technology of reprocessing. The results show that the graphite matrix can be effectively separated from the coated particle without any damage to the SiC layer. Secondly, the microwave-assisted heating was applied to dissolve the UO{sub 2} kernel from the crashed coated fuel particles. The ceramic UO{sub 2} as the solute has a good ability to absorb the microwave energy. The results of UO{sub 2} kernel dissolution from crushed coated particles by microwave heating show that the total dissolution percentage of UO{sub 2} is more than 99.99% after 3 times cross-flow dissolution with the following parameters: 8 mol/L HNO{sub 3}, temperature 100 Celsius degrees, initial ratio of solid to liquid 1.2 g/ml. (authors)

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

    SciTech Connect

    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.

  10. Reprocessing RERTR fuels

    SciTech Connect

    Rodrigues, G.C.

    1983-01-01

    The Reduced Enrichment Research and Test Reactor Program is one element of the United States Government's nonproliferation effort. High density, low enrichment aluminum-clad dispersed uranium compound fuels may be substituted for the highly enriched aluminum-clad aluminum-uranium alloy fuels now in use. Savannah River Laboratory has performed studies which demonstrate reprocessability of spent RERTR fuels at Savannah River Plant. Results of dissolution and feed preparation tests with both unirradiated and irradiated (up to approximately 90% burnup) fuels are presented. 13 references, 2 figures, 4 tables.

  11. Nuclear fuel reprocessing deactivation plan for the Idaho Chemical Processing Plant, Revision 1

    SciTech Connect

    Patterson, M.W.

    1994-10-01

    The decision was announced on April 28, 1992 to cease all United States Department of Energy (DOE) reprocessing of nuclear fuels. This decision leads to the deactivation of all fuels dissolution, solvent extraction, krypton gas recovery operations, and product denitration at the Idaho Chemical Processing Plant (ICPP). The reprocessing facilities will be converted to a safe and stable shutdown condition awaiting future alternate uses or decontamination and decommissioning (D&D). This ICPP Deactivation Plan includes the scope of work, schedule, costs, and associated staffing levels necessary to achieve a safe and orderly deactivation of reprocessing activities and the Waste Calcining Facility (WCF). Deactivation activities primarily involve shutdown of operating systems and buildings, fissile and hazardous material removal, and related activities. A minimum required level of continued surveillance and maintenance is planned for each facility/process system to ensure necessary environmental, health, and safety margins are maintained and to support ongoing operations for ICPP facilities that are not being deactivated. Management of the ICPP was transferred from Westinghouse Idaho Nuclear Company, Inc. (WINCO) to Lockheed Idaho Technologies Company (LITCO) on October 1, 1994 as part of the INEL consolidated contract. This revision of the deactivation plan (formerly the Nuclear Fuel Reprocessing Phaseout Plan for the ICPP) is being published during the consolidation of the INEL site-wide contract and the information presented here is current as of October 31, 1994. LITCO has adopted the existing plans for the deactivation of ICPP reprocessing facilities and the plans developed under WINCO are still being actively pursued, although the change in management may result in changes which have not yet been identified. Accordingly, the contents of this plan are subject to revision.

  12. Molten tin reprocessing of spent nuclear fuel elements. [Patent application; continuous process

    DOEpatents

    Heckman, R.A.

    1980-12-19

    A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support te liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

  13. Spent nuclear fuel reprocessing modeling

    SciTech Connect

    Tretyakova, S.; Shmidt, O.; Podymova, T.; Shadrin, A.; Tkachenko, V.; Makeyeva, I.; Tkachenko, V.; Verbitskaya, O.; Schultz, O.; Peshkichev, I.

    2013-07-01

    The long-term wide development of nuclear power requires new approaches towards the realization of nuclear fuel cycle, namely, closed nuclear fuel cycle (CNFC) with respect to fission materials. Plant nuclear fuel cycle (PNFC), which is in fact the reprocessing of spent nuclear fuel unloaded from the reactor and the production of new nuclear fuel (NF) at the same place together with reactor plant, can be one variant of CNFC. Developing and projecting of PNFC is a complicated high-technology innovative process that requires modern information support. One of the components of this information support is developed by the authors. This component is the programme conducting calculations for various variants of process flow sheets for reprocessing SNF and production of NF. Central in this programme is the blocks library, where the blocks contain mathematical description of separate processes and operations. The calculating programme itself has such a structure that one can configure the complex of blocks and correlations between blocks, appropriate for any given flow sheet. For the ready sequence of operations balance calculations are made of all flows, i.e. expenses, element and substance makeup, heat emission and radiation rate are determined. The programme is open and the block library can be updated. This means that more complicated and detailed models of technological processes will be added to the library basing on the results of testing processes using real equipment, in test operating mode. The development of the model for the realization of technical-economic analysis of various variants of technologic PNFC schemes and the organization of 'operator's advisor' is expected. (authors)

  14. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    SciTech Connect

    Jubin, R. T.; Strachan, D. M.; Ilas, G.; Spencer, B. B.; Soelberg, N. R.

    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

  15. Spectroscopic On-Line Monitoring for Process Control and Safeguarding of Radiochemical Streams in Nuclear Fuel Reprocessing Facilities

    SciTech Connect

    Bryan, Samuel A.; Levitskaia, Tatiana G.; Casella, Amanda J.; Peterson, James M.; Johnsen, Amanda M.; Lines, Amanda M.; Thomas, Elizabeth M.

    2011-03-01

    The current book chapter presents preliminary work toward the use of spectroscopic on-line monitoring for process control and safeguarding of radiochemical streams. Raman spectroscopy was demonstrated as a method for determining U(VI), nitrate, and nitric acid, while visible-near-infrared (vis-NIR) spectroscopy was demonstrated as a method for determining Pu(IV), Np(V), U(VI) and Nd(III). This method has been established using fuel reprocessing solution stimulants under dynamic flow conditions and on commercial spent nuclear fuel samples. Partial least squares (PLS) models for each analyte were prepared, and the fits of the data presented. A brief review of literature relevant to the use of vibrational spectroscopy and physicochemical measurements for process monitoring of nuclear fuel solutions is reported.

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

    NASA Astrophysics Data System (ADS)

    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.

  17. Equipment specifications for an electrochemical fuel reprocessing plant

    SciTech Connect

    Hemphill, Kevin P

    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.

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

    SciTech Connect

    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.

  19. Methodology of Qualification of CCIM Vitrification Process Applied to the High- Level Liquid Waste from Reprocessed Oxide Fuels - 12438

    SciTech Connect

    Lemonnier, S.; Labe, V.; Ledoux, A.; Nonnet, H.; Godon, N.

    2012-07-01

    The vitrification of high-level liquid waste from reprocessed oxide fuels (UOX fuels) by Cold Crucible Induction Melter is planed by AREVA in 2013 in a production line of the R7 facility at La Hague plant. Therefore, the switch of the vitrification technology from the Joule Heated Metal Melter required a complete process qualification study. It involves three specialties, namely the matrix formulation, the glass long-term behavior and the vitrification process development on full-scale pilot. A new glass frit has been elaborated in order to adapt the redox properties and the thermal conductivity of the glass suitable for being vitrified with the Cold Crucible Induction Melter. The role of cobalt oxide on the long term behavior of the glass has been described in the range of the tested concentrations. Concerning the process qualification, the nominal tests, the sensitivity tests and the study of the transient modes allowed to define the nominal operating conditions. Degraded operating conditions tests allowed to identify means of detecting incidents leading to these conditions and allowed to define the procedures to preserve the process equipments protection and the material quality. Finally, the endurance test validated the nominal operating conditions over an extended time period. This global study allowed to draft the package qualification file. The qualification file of the UOX package is currently under approval by the French Nuclear Safety Authority. (authors)

  20. Spectroscopic On-Line Monitoring for Process Control and Safeguarding of Radiochemical Streams in Nuclear Fuel Reprocessing Facilities

    SciTech Connect

    Bryan, Samuel A.; Levitskaia, Tatiana G.; Casella, Amanda J.; Peterson, James M.; Johnsen, Amanda M.; Lines, Amanda M.; Thomas, Elizabeth M.

    2011-02-24

    There is a renewed interest worldwide to promote the use of nuclear power. The long term successful use of nuclear power is critically dependent upon adequate and safe processing and disposition of the spent nuclear fuel. Liquid-liquid extraction is a separation technique commonly employed for the processing of the dissolved spent nuclear fuel. Our approach is based on the prerequisite that real time monitoring of the solvent extraction flowsheets provides unique capability to quickly detect unwanted manipulations with fissile isotopes present in the radiochemical streams during reprocessing activities. Our ability to identify material intentionally diverted from a liquid-liquid solvent extraction contactor system was successfully tested using on-line process monitoring. A chemical diversion and detection from a solvent extraction scheme was demonstrated using a centrifugal contactor system operating with the simulant PUREX extraction system of Nd(NO3)3/nitric acid aqueous phase and TBP/dodecane organic phase. A portion of the feed from a counter-current extraction system was diverted while a continuous extraction experiment was underway; the spectroscopic on-line process monitoring system was simultaneously measuring the feed, raffinate and organic products streams. The amount observed to be diverted by on-line spectroscopic process monitoring was in excellent agreement with values based from known mass of sample directly taken (diverted) from system feed solution. We conclude that real-time spectroscopic process monitoring would be a useful tool for the IAEA to detect the diversion of nuclear material in a timely manner. This poster summarizes our methodology of on-line process monitoring and shows recent results of specific examples.

  1. Reprocessing of research reactor fuel the Dounreay option

    SciTech Connect

    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.

  2. Classic Nuclear Fuel Reprocessing Flowsheet

    SciTech Connect

    Fallgren, Andrew James

    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.

  3. CORAL: a stepping stone for establishing the Indian fast reactor fuel reprocessing technology

    SciTech Connect

    Venkataraman, M.; Natarajan, R.; Raj, Baldev

    2007-07-01

    The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR) spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)

  4. Summary of nuclear fuel reprocessing activities around the world

    SciTech Connect

    Mellinger, P.J.; Harmon, K.M.; Lakey, L.T.

    1984-11-01

    This review of international practices for nuclear fuel reprocessing was prepared to provide a nontechnical summary of the current status of nuclear fuel reprocessing activities around the world. The sources of information are widely varied.

  5. Integrated international safeguards concepts for fuel reprocessing

    SciTech Connect

    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.

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

    SciTech Connect

    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.

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

    SciTech Connect

    Soelberg, Nicolas R.; Garn, Troy; Greenhalgh, Mitchell; Law, Jack; Jubin, Robert T.; Strachan, Denis M.; Thallapally, Praveen K.

    2013-07-22

    Nuclear fission results in the production of fission products and activation products, some of which tend to be volatile during used fuel reprocessing. These can evolve in volatile species in the reprocessing facility off-gas streams, depending on the separations and reprocessing technologies that are used. Radionuclides that have been identified as “volatile radionuclides” are noble gases (most notably isotopes of Kr and Xe); 3H; 14C; and 129I. Radionuclides that tend to form volatile species that evolve into reprocessing facility off-gas systems are more challenging to efficiently control compared to radionuclides that tend to stay in solid or liquid phases. Future used fuel reprocessing facilities in the United States can require efficient capture of some volatile radionuclides in their off-gas streams to meet regulatory emission requirements. In aqueous reprocessing, these radionuclides are most commonly expected to evolve into off-gas streams in tritiated water [3H2O (T2O) and 3HHO (THO)], radioactive CO2, noble gases, and gaseous HI, I2, or volatile organic iodides. The fate and speciation of these radionuclides from a non-aqueous fuel reprocessing facility is less well known at this time, but active investigations are in progress. An Off-Gas Sigma Team was formed in late FY 2009 to integrate and coordinate the Fuel Cycle Research and Development (FCR&D) activities directed towards the capture and sequestration of the these volatile radionuclides (Jubin 2012a). The Sigma Team concept was envisioned to bring together multidisciplinary teams from across the DOE complex that would work collaboratively to solve the technical challenges and to develop the scientific basis for the capture and immobilization technologies such that the sum of the efforts was greater than the individual parts. The Laboratories currently participating in this effort are Argonne National Laboratory (ANL), Idaho National Laboratory (INL), Oak Ridge National Laboratory (ORNL), Pacific

  8. Concept of advanced spent fuel reprocessing based on ion exchange

    SciTech Connect

    Suzuki, Tatsuya; Takahashi, Kazuyuki; Nogami, Masanobu; Nomura, Masao; Fujii, Yasuhiko; Ozawa, Masaki |; Koyama, Shinichi; Mimura, Hitosi; Fujita, Reiko

    2007-07-01

    . Furthermore, the ion exchange is appropriate for multi-element mutual separation rather than single element extraction. In the future, ion exchange reprocessing would be expected to be the comprehensive separation process for spent fuels to recover precious and usable elements and to reduce the amount of wastes. (authors)

  9. MONITORING SPENT NUCLEAR FUEL REPROCESSING CONDITIONS NON-DESTRUCTIVELY AND IN NEAR-REAL-TIME USING THE MULTI-ISOTOPE PROCESS (MIP) MONITOR

    SciTech Connect

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

    2010-05-07

    Researchers from Pacific Northwest National Laboratory and The Ohio State University are working to develop a system for monitoring spent nuclear fuel reprocessing facilities on-line, nondestructively, and in near-real-time. This method, known as the Multi-Isotope Process (MIP) Monitor, is based upon the measurement of distribution patterns of a suite of indicator (radioactive) isotopes present within product and waste streams of a nuclear reprocessing facility. Signatures from these indicator isotopes are monitored on-line by gamma spectrometry and compared, in near-real-time, to patterns representing "normal" process conditions using multivariate pattern recognition software. By targeting gamma-emitting indicator isotopes, the MIP Monitor approach is compatible with the use of small, portable, high-resolution gamma detectors that may be easily deployed throughout an existing facility. In addition, utilization of a suite of radio-elements, including ones with multiple oxidation states, increases the likelihood that attempts to divert material via process manipulation would be detected. Proof-of-principle modeling exercises simulating changes in acid strength have been completed and the results are promising. Laboratory testing is currently under way and significant results are available. Recent experimental results, along with an overview of the method are presented.

  10. Survey of Dynamic Simulation Programs for Nuclear Fuel Reprocessing

    SciTech Connect

    Troy J. Tranter; Daryl R. Haefner

    2008-06-01

    The absence of any industrial scale nuclear fuel reprocessing in the U.S. has precluded the necessary driver for developing the advanced simulation capability now prevalent in so many other industries. Modeling programs to simulate the dynamic behavior of nuclear fuel separations and processing were originally developed to support the US government’s mission of weapons production and defense fuel recovery. Consequently there has been little effort is the US devoted towards improving this specific process simulation capability during the last two or three decades. More recent work has been focused on elucidating chemical thermodynamics and developing better models of predicting equilibrium in actinide solvent extraction systems. These equilibrium models have been used to augment flowsheet development and testing primarily at laboratory scales. The development of more robust and complete process models has not kept pace with the vast improvements in computational power and user interface and is significantly behind simulation capability in other chemical processing and separation fields.

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

    SciTech Connect

    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.

  12. Consolidated fuel reprocessing program. Progress report, July 1-September 30, 1981

    SciTech Connect

    1981-12-01

    Technical progress is reported in overview fashion in the following areas: process development, laboratory R and D, engineering research, engineering systems, integrated equipment test facility (IET) operations, and HTGR fuel reprocessing. (DLC)

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

    SciTech Connect

    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.

  14. Decontamination and decommissioning of a fuel reprocessing pilot plant

    SciTech Connect

    Heine, W.F.; Speer, D.R.

    1988-01-01

    SYNOPSIS The strontium Semiworks Pilot Fuel Reprocessing Plant at the Hanford Site in Washington State was decommissioned by a combination of dismantlement and entombment. The facility contained 9600 Ci of Sr-90 and 10 Ci of plutonium. Process cells were entombed in place. The above-grade portion of one cell with 1.5-m- (5-ft-) thick walls and ceilings was demolished by means of expanding grout. A contaminated stack was remotely sandblasted and felled by explosives. The entombed structures were covered with a 4.6-m- (15-ft-) thick engineered earthen barrier. 5 figs., 2 tabs.

  15. Experience in the reprocessing of mixed-oxide fuels at PNC (Power Reactor and Nuclear Fuel Development Corporation)

    SciTech Connect

    Komatsu, Hisato; Onishi, Moichi; Araya, Sadao; Fukushima, Misao

    1989-01-01

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) in Japan has experience in reprocessing mixed-oxide (MOX) fuels for the advanced thermal reactor (ATR) Fugen at the Tokai Reprocessing Plant (TRP) and for fast breeder reactors (FBRs) at the Chemical Processing Facility (CPF). The TRP was originally designed and constructed as the first reprocessing plant for light water reactor fuels in Japan. It has processed {approximately}400 t of spent fuels since 1977. To utilize recovered plutonium, PNC has developed the prototype ATR Fugen. This reactor has been operated using MOX fuel since 1978. In parallel, utilities are promoting a plutonium thermal project. Several MOX assemblies have already been loaded in a boiling water and a pressurized water reactor. To facilitate the operation of Fugen and promote research and development for the reprocessing of MOX fuels in Japan, PNC obtained a license for reprocessing fuels for Fugen at TRP in 1985. PNC has designed and constructed the CPF at Tokai Works to conduct basic research on the reprocessing of FBR fuels. The Recycle Equipment Test Facility, an engineering scale hot facility, is now being designed for further R and D in this field. It will start hot operation in the mid-1990s.

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

    SciTech Connect

    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/.

  17. Molten tin reprocessing of spent nuclear fuel elements

    DOEpatents

    Heckman, Richard A.

    1983-01-01

    A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

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

    SciTech Connect

    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.

  19. Alternate extractants to tributyl phosphate for reactor fuel reprocessing

    SciTech Connect

    Crouse, D.J.; Arnold, W.D.; Hurst, F.J.

    1983-01-01

    Both tri(n-hexyl) phosphate (THP) and tri(2-ethylhexyl) phosphate (TEHP) have some important potential process advantages over TBP for reactor fuel reprocessing. These include negligible aqueous phase solubility and less tendency toward third phase and crud formation. The alkyl chain branching of TEHP makes it much more stable to chemical degradation than TBP and probably also accounts for its much weaker ruthenium extraction. The higher uranium and plutonium extraction power of THP and TEHP allows higher solvent loadings in extraction but makes them somewhat more difficult to strip. The phase separation properties of 1.09 M solutions of THP and TEHP are inferior to those of 1.09 M TBP (30 vol %) but are favorable at lower concentrations. Use of more dilute THP and TEHP solutions is recommended for this reason and to obtain a better balance of extraction power in the extraction versus stripping steps.

  20. New method of uranium and plutonium extraction in reprocessing of the spent nuclear fuel

    SciTech Connect

    Volk, V.; Dvoeglazov, K.; Veslov, S.; Rubisov, V.; Alekseenko, V.; Krivitsky, Y.; Alekseenko, S.; Bondin, V.

    2013-07-01

    It is shown that a two-stage process of uranium and plutonium extraction during the reprocessing of spent nuclear fuel solves the problem of obtaining a high-concentrated extract without increasing the loss risk with raffinate and avoids the accumulation of plutonium in the unit. A possible further optimization of the process would be the creation of steps inside the stages.

  1. Removal of actinides from nuclear fuel reprocessing wastes using an organophosphorous extractant. [DHDECMP

    SciTech Connect

    Chamberlain, D.B.; Maxey, H.R.; McIsaac, L.D.; McManus, G.J.

    1980-01-01

    By removing actinides from nuclear fuel reprocessing wastes, long term waste storage hazards are reduced. A solvent extraction process to remove actinides has been demonstrated in miniature mixer-settlers and in simulated columns using actinide feeds. Nonradioactive pilot plant results have established the feasibility of using pulse columns for the process.

  2. Electrolysis cell for reprocessing plutonium reactor fuel

    DOEpatents

    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.

  3. Electrolysis cell for reprocessing plutonium reactor fuel

    DOEpatents

    Miller, W.E.; Steindler, M.J.; Burris, L.

    1985-01-04

    An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals is claimed. The cell includes 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. The anode basket is extendable into the lower pool to dissolve at least some metallic contaminants; the anode basket contains the spent fuel acting as a second anode when in the electrolyte.

  4. Organic derivatives of hydrazine and hydroxylamine in future technology of spent nuclear fuel reprocessing

    SciTech Connect

    Koltunov, V.S.; Baranov, S.M.

    1993-12-31

    An important issue in nuclear fuel reprocessing is the reduction of salts. It is seen that this can be accomplished utilizing organic derivatives of hydrazine and hydroxylamine as reductants of Np(VI) and Pu(IV). The chemistry of this process is described.

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

    SciTech Connect

    Mcwilliams, A. J.

    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.

  6. Status of radioiodine control for nuclear fuel reprocessing plants

    SciTech Connect

    Burger, L.L.; Scheele, R.D.

    1983-07-01

    This report summarizes the status of radioiodine control in a nuclear fuel reprocessing plant with respect to capture, fixation, and disposal. Where possible, we refer the reader to a number of survey documents which have been published in the last four years. We provide updates where necessary. Also discussed are factors which must be considered in developing criteria for iodine control. For capture from gas streams, silver mordenite and a silver nitrate impregnated silica (AC-6120) are considered state-of-the-art and are recommended. Three aqueous scrubbing processes have been demonstrated: Caustic scrubbing is simple but probably will not give an adequate iodine retention by itself. Mercurex (mercuric nitrate-nitric acid scrubbing) has a number of disadvantages including the use of toxic mercury. Iodox (hyperazeotropic nitric acid scrubbing) is effective but employs a very corrosive and hazardous material. Other technologies have been tested but require extensive development. The waste forms recommended for long-term storage or disposal are silver iodide, the iodates of barium, strontium, or calcium, and silver loaded sorbents, all fixed in cement. Copper iodide in bitumen (asphalt) is a possibility but requires testing. The selection of a specific form will be influenced by the capture process used.

  7. Method for reprocessing and separating spent nuclear fuels

    DOEpatents

    Krikorian, Oscar H.; Grens, John Z.; Parrish, Sr., William H.

    1983-01-01

    Spent nuclear fuels, including actinide fuels, volatile and non-volatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

  8. Method for reprocessing and separating spent nuclear fuels. [Patent application

    DOEpatents

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

    1982-01-19

    Spent nuclear fuels, including actinide fuels, volatile and nonvolatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

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

    SciTech Connect

    N. R. Soelberg; J. D. Law; T. G. Garn; M. Greenhalgh; R. T. Jubin; P. Thallapally; D. M. Strachan

    2013-08-01

    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.

  10. Improved measurement of aluminum in irradiated fuel reprocessed at the Savannah River Site

    SciTech Connect

    Maxwell, S.L. III

    1991-12-31

    At the Savannah River Site (SRS), irradiated fuel from research reactor operators or their contract fuel service companies is reprocessed in the H-Canyon Separations Facility. Final processing costs are based on analytical measurements of the amount of total metal dissolved. Shipper estimates for uranium and uranium-235 and measured values at SRS have historically agreed very well. There have occasionally been significant differences between shipper estimates for aluminum and the aluminum content determined at SRS. To minimize analytical error that might contribute to poor shipper-receiver agreement for the reprocessing of off-site fuel, a new analytical method to measure aluminum was developed by SRS Analytical Laboratories at the Central Laboratory Facilities. An EDTA (ethylenediaminetetraacetic acid) titration method, subject to dissolver matrix interferences, was previously used at SRS to measure aluminum in H-Canyon dissolver during the reprocessing of offsite fuel. The new method combines rapid ion exchange technology with direct current argon plasma spectrometry to enhance the reliability of aluminum measurements for off-site fuel. The technique rapidly removes spectral interferences such as uranium and significantly lowers gamma levels due to fission products. Aluminium is separated quantitatively by using an anion exchange technique that employs oxalate complexing, small particle size resin and rapid flow rates. The new method, which has eliminated matrix interference problems with these analyses and improved the quality of aluminum measurements, has improved the overall agreement between shipper-receiver values for offsite fuel processed SRS.

  11. Improved measurement of aluminum in irradiated fuel reprocessed at the Savannah River Site

    SciTech Connect

    Maxwell, S.L. III.

    1991-01-01

    At the Savannah River Site (SRS), irradiated fuel from research reactor operators or their contract fuel service companies is reprocessed in the H-Canyon Separations Facility. Final processing costs are based on analytical measurements of the amount of total metal dissolved. Shipper estimates for uranium and uranium-235 and measured values at SRS have historically agreed very well. There have occasionally been significant differences between shipper estimates for aluminum and the aluminum content determined at SRS. To minimize analytical error that might contribute to poor shipper-receiver agreement for the reprocessing of off-site fuel, a new analytical method to measure aluminum was developed by SRS Analytical Laboratories at the Central Laboratory Facilities. An EDTA (ethylenediaminetetraacetic acid) titration method, subject to dissolver matrix interferences, was previously used at SRS to measure aluminum in H-Canyon dissolver during the reprocessing of offsite fuel. The new method combines rapid ion exchange technology with direct current argon plasma spectrometry to enhance the reliability of aluminum measurements for off-site fuel. The technique rapidly removes spectral interferences such as uranium and significantly lowers gamma levels due to fission products. Aluminium is separated quantitatively by using an anion exchange technique that employs oxalate complexing, small particle size resin and rapid flow rates. The new method, which has eliminated matrix interference problems with these analyses and improved the quality of aluminum measurements, has improved the overall agreement between shipper-receiver values for offsite fuel processed SRS.

  12. Feasibility study of a plant for LWR used fuel reprocessing by pyrochemical methods

    SciTech Connect

    Bychkov, A.V.; Kormilitsyn, M.V.; Savotchkin, Yu.P.; Sokolovsky, Yu.S.; Baganz, Catherine; Lopoukhine, Serge; Maurin, Guy; Medzadourian, Michel

    2007-07-01

    In 2005, experts from AREVA and RIAR performed a joint research work on the feasibility study of a plant reprocessing 1000 t/y of LWR spent nuclear fuel by the gas-fluoride and pyro-electrochemical techniques developed at RIAR. This work was based on the RIAR experience in development of pyrochemical processes and AREVA experience in designing UNF reprocessing plants. UNF reprocessing pyrochemical processes have been developed at RIAR at laboratory scale and technology for granulated MOX fuel fabrication and manufacturing of vibro-packed fuel rods is developed at pilot scale. The research work resulted in a preliminary feasibility assessment of the reprocessing plant according to the norms and standards applied in France. The study results interpretation must integrate the fact that the different technology steps are at very different stage of development. It appears clearly however that in its present state of development, pyro-electrochemical technology is not adapted to the treatment of an important material flow issuing from thermal reactors. There is probably an economic optimum to be studied for the choice of hydrometallurgical or pyro-electrochemical technology, depending on the area of application. This work is an example of successful and fruitful collaboration between French and Russian specialists. (authors)

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

    SciTech Connect

    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

  14. Development of fast breeder reactor fuel reprocessing technology at the Power Reactor and Nuclear Fuel Development Corporation

    SciTech Connect

    Kawata, T.; Takeda, H.; Togashi, A.; Hayashi, S. . Tokai Works); Stradley, J.G. )

    1991-01-01

    For the past two decades, a broad range of research development (R D) programs to establish fast breeder reactor (FBR) system and its associated fuel cycle technology have been pursued by the Power Reactor and Nuclear Fuel Development Corporation (PNC). Developmental activities for FBR fuel reprocessing technology have been primarily conducted at PNC Tokai Works where many important R D facilities for nuclear fuel cycle are located. These include cold and uranium tests for process equipment development in the Engineering Demonstration Facilities (EDF)-I and II, and laboratory-scale hot tests in the Chemical Processing Facility (CPF) where fuel dissolution and solvent extraction characteristics are being investigated with irradiated FBR fuel pins whose burn-up ranges up to 100,000 MWd/t. An extensive effort has also been made at EDF-III to develop advanced remote technology which enables to increase plant availability and to decrease radiation exposures to the workers in future reprocessing plants. The PNC and the United States Department of Energy (USDOE) entered into the joint collaboration in which the US shares the R Ds to support FBR fuel reprocessing program at the PNC. Several important R Ds on advanced process equipment such as a rotary dissolver and a centrifugal contactor system are in progress in a joint effort with the Oak Ridge National Laboratory (ORNL) Consolidated Fuel Reprocessing Program (CFRP). In order to facilitate hot testing on advanced processes and equipment, the design of a new engineering-scale hot test facility is now in progress aiming at the start of hot operation in late 90's. 31 refs., 2 tabs.

  15. Issues for Conceptual Design of AFCF and CFTC LWR Spent Fuel Separations Influencing Next-Generation Aqueous Fuel Reprocessing

    SciTech Connect

    D. Hebditch; R. Henry; M. Goff; K. Pasamehmetoglu; D. Ostby

    2007-09-01

    In 2007, the U.S. Department of Energy (DOE) published the Global Nuclear Energy Partnership (GNEP) strategic plan, which aims to meet US and international energy, safeguards, fuel supply and environmental needs by harnessing national laboratory R&D, deployment by industry and use of international partnerships. Initially, two industry-led commercial scale facilities, an advanced burner reactor (ABR) and a consolidated fuel treatment center (CFTC), and one developmental facility, an advanced fuel cycle facility (AFCF) are proposed. The national laboratories will lead the AFCF to provide an internationally recognized R&D center of excellence for developing transmutation fuels and targets and advancing fuel cycle reprocessing technology using aqueous and pyrochemical methods. The design drivers for AFCF and the CFTC LWR spent fuel separations are expected to impact on and partly reflect those for industry, which is engaging with DOE in studies for CFTC and ABR through the recent GNEP funding opportunity announcement (FOA). The paper summarizes the state-of-the-art of aqueous reprocessing, gives an assessment of engineering drivers for U.S. aqueous processing facilities, examines historic plant capital costs and provides conclusions with a view to influencing design of next-generation fuel reprocessing plants.

  16. Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant

    SciTech Connect

    Perkins, W.C.; Durant, W.S.; Dexter, A.H.

    1980-12-01

    The occurrence of certain potential events in nuclear fuel reprocessing plants could lead to significant consequences involving risk to operating personnel or to the general public. This document is a compilation of such potential initiating events in nuclear fuel reprocessing plants. Possible general incidents and incidents specific to key operations in fuel reprocessing are considered, including possible causes, consequences, and safety features designed to prevent, detect, or mitigate such incidents.

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

    SciTech Connect

    Not Available

    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.

  18. Workshop on instrumentation and analyses for a nuclear fuel reprocessing hot pilot plant

    SciTech Connect

    Babcock, S.M.; Feldman, M.J.; Wymer, R.G.; Hoffman, D.

    1980-05-01

    In order to assist in the study of instrumentation and analytical needs for reprocessing plants, a workshop addressing these needs was held at Oak Ridge National Laboratory from May 5 to 7, 1980. The purpose of the workshop was to incorporate the knowledge of chemistry and of advanced measurement techniques held by the nuclear and radiochemical community into ideas for improved and new plant designs for both process control and inventory and safeguards measurements. The workshop was athended by experts in nuclear and radiochemistry, in fuel recycle plant design, and in instrumentation and analysis. ORNL was a particularly appropriate place to hold the workshop since the Consolidated Fuel Reprocessing Program (CFRP) is centered there. Requirements for safeguarding the special nuclear materials involved in reprocessing, and for their timely measurement within the process, within the reprocessing facility, and at the facility boundaries are being studied. Because these requirements are becoming more numerous and stringent, attention is also being paid to the analytical requirements for these special nuclear materials and to methods for measuring the physical parameters of the systems containing them. In order to provide a focus for the consideration of the workshop participants, the Hot Experimental Facility (HEF) being designed conceptually by the CFRP was used as a basis for consideration and discussions.

  19. Processes for the control of /sup 14/CO/sub 2/ during reprocessing

    SciTech Connect

    Notz, K.J.; Holladay, D.W.; Forsberg, C.W.; Haag, G.L.

    1980-01-01

    The fixation of /sup 14/CO/sub 2/ may be required at some future time because of the significant fractional contribution of /sup 14/C, via the ingestion pathway, to the total population dose from the nuclear fuel cycle, even though the actual quantity of this dose is very small when compared to natural background. The work described here was done in support of fuel reprocessing development, of both graphite fuel (HTGRs) and metal-clad fuel (LWRs and LMFBRs), and was directed to the control of /sup 14/CO/sub 2/ released during reprocessing operations. However, portions of this work are also applicable to the control of /sup 14/CO/sub 2/ released during reactor operation. The work described falls in three major areas: (1) The application of liquid-slurry fixation with Ca(OH)/sub 2/, which converts the CO/sub 2/ to CaCO/sub 3/, carried out after treatment of the CO/sub 2/-containing stream to remove other gaseous radioactive components, mainly /sup 85/Kr. This approach is primarily for application to HTGR fuel reprocessing. (2) The above process for CO/sub 2/ fixation, but used ahead of Kr removal, and followed by a molecular sieve process to take out the /sup 85/Kr. This approach was developed for use with HTGR reprocessing, but certain aspects also have application to metal-clad fuel reprocessing and to reactor operation. (3) The use of solid Ba(OH)/sub 2/ hydrate reacting directly with the gaseous phase. This process is generally applicable to both reprocessing and to reactor operation.

  20. Container for reprocessing and permanent storage of spent nuclear fuel assemblies

    DOEpatents

    Forsberg, Charles W.

    1992-01-01

    A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

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

    SciTech Connect

    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.

  2. Development Of Electronic Tongue System For Quantification Of Rare Earth Metals In Spent Nuclear Fuel Reprocessing

    NASA Astrophysics Data System (ADS)

    Kirsanov, Dmitry; Legin, Andrey; Tkachenko, Mila; Surzhina, Irina; Khaidukova, Maria; Babain, Vasily

    2011-09-01

    The present study deals with development of an electronic tongue multisensor system which is capable of simultaneous quantification of several RE in a complex mixtures containing uranium and thorium in the acidic media simulating typical composition of spent nuclear fuel reprocessing solutions. Combination of specially designed cross-sensitive potentiometric sensors and multivariate data processing allows for fast and simple analysis of such mixtures.r.

  3. An Assessment of Spent Fuel Reprocessing for Actinide Destruction and Resource Sustainability.

    SciTech Connect

    Cipiti, Benjamin B.; Smith, James D.

    2008-09-01

    The reprocessing and recycling of spent nuclear fuel can benefit the nuclear fuel cycle by destroying actinides or extending fissionable resources if uranium supplies become limited. The purpose of this study was to assess reprocessing and recycling in both fast and thermal reactors to determine the effectiveness for actinide destruction and resource utilization. Fast reactor recycling will reduce both the mass and heat load of actinides by a factor of 2, but only after 3 recycles and many decades. Thermal reactor recycling is similarly effective for reducing actinide mass, but the heat load will increase by a factor of 2. Economically recoverable reserves of uranium are estimated to sustain the current global fleet for the next 100 years, and undiscovered reserves and lower quality ores are estimated to contain twice the amount of economically recoverable reserves--which delays the concern of resource utilization for many decades. Economic analysis reveals that reprocessed plutonium will become competitive only when uranium prices rise to about %24360 per kg. Alternative uranium sources are estimated to be competitive well below that price. Decisions regarding the development of a near term commercial-scale reprocessing fuel cycle must partially take into account the effectiveness of reactors for actnides destruction and the time scale for when uranium supplies may become limited. Long-term research and development is recommended in order to make more dramatic improvements in actinide destruction and cost reductions for advanced fuel cycle technologies.The original scope of this work was to optimize an advanced fuel cycle using a tool that couples a reprocessing plant simulation model with a depletion analysis code. Due to funding and time constraints of the late start LDRD process and a lack of support for follow-on work, the project focused instead on a comparison of different reprocessing and recycling options. This optimization study led to new insight into

  4. Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

    SciTech Connect

    Bond, W.D.; Mailen, J.C.; Michaels, G.E.

    1992-10-01

    The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO[sub 2] fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed.

  5. Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

    SciTech Connect

    Bond, W.D.; Mailen, J.C.; Michaels, G.E.

    1992-10-01

    The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO{sub 2} fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed.

  6. Materials management in an internationally safeguarded fuels reprocessing plant

    SciTech Connect

    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)

  7. Application of a Plasma Mass Separator to Advanced LWR Spent Fuel Reprocessing

    SciTech Connect

    Freeman, Richard; Miller, Robert; Papay, Larry; Wagoner, John; Ahlfeld, Charles; Czerwinski, Ken

    2006-07-01

    The US Department of Energy (DOE) is investigating spent fuel reprocessing for the purposes of increasing the effective capacity of a deep geological repository, reducing the radiotoxicity of waste placed in the repository and conserving nuclear fuel resources. DOE is considering hydro-chemical processing of the spent fuel after cutting the fuel cladding and fuel dissolution in nitric acid. The front end process, known as UREX, is largely based on the PUREX process and extracts U, Tc as well as fission product gases. A number of additional processing steps have become known as UREX+. One of the steps includes a further chemical treatment of remove Cs and Sr to reduce repository heat load. Other steps include successive extraction of the actinides from residual fission products, including the lanthanides. The additional UREX+ processing renders the actinides suitable for burning as reactor fuel in an advanced reactor to convert actinides to shorter-lived fission products and to produce power. New methods for separating groups of elements by their atomic mass have been developed and can be exploited to enhance spent fuel reprocessing. These physical processes dry the waste streams so that they can be vaporized and singly ionized in plasma that is contained in longitudinal magnetic and perpendicular electric fields. Proper configuration of the fields causes the plasma to rapidly rotate and expel heavier mass ions at the center of the machine. Lower mass ions form closed orbits within the cylindrical plasma column and are transported to either end of the machine. This plasma mass separator was originally developed to reduce the mass of material that must be immobilized in borosilicate glass from DOE defense waste at former weapons production facilities. The plasma mass separator appears to be well-suited for processing the UREX raffinate and solids streams by exploiting the large atomic mass gap that exists between lanthanides (< {approx}180 amu) and actinides

  8. ON-LINE MONITORING FOR CONTROL AND SAFEGUARDING OF RADIOCHEMICAL STREAMS AT SPENT FUEL REPROCESSING PLANT

    SciTech Connect

    Bryan, Samuel A.; Levitskaia, Tatiana G.; Lines, Amanda M.; Billing, Justin M.; Casella, Amanda J.; Johnsen, Amanda M.; Peterson, James M.; Thomas, Elizabeth M.

    2009-11-10

    Advanced techniques that enhance safeguarding of spent fuel reprocessing plants are urgently needed. Our approach is based on the prerequisite that real-time monitoring of solvent extraction flowsheets at a spent fuel reprocessing plant provides the unique capability to quickly detect unwanted manipulations with fissile isotopes present in the radiochemical streams during reprocessing activities. The methods used to monitor these processes must be robust and capable of withstanding harsh radiation and chemical environments. A new on-line monitoring system satisfying these requirements and featuring Raman spectroscopy combined with a Coriolis and conductivity probes recently has been developed by our research team for tank waste retrieval. It provides immediate chemical data and flow parameters of high-level radioactive waste streams with high brine content generated during retrieval activities from nuclear waste storage tanks at the Hanford Site. The nature of the radiochemical streams at the spent fuel reprocessing plant calls for additional spectroscopic information that can be gained by using Vis-NIR capabilities augmenting Raman spectroscopy. A fiber optic Raman probe allows monitoring of high concentration species encountered in both aqueous and organic phases within the UREX suite of flowsheets, including metal oxide ions, such as uranyl, components of the organic solvent, inorganic oxo-anions, and water. Actinides and lanthanides are monitored remotely by Vis-NIR spectroscopy in aqueous and organic phases. In this report, we present our results on spectroscopic measurements of simulant flowsheet solutions and commercial fuels designed to demonstrate the applicability of Raman and Vis-NIR spectroscopic analysis for actual dissolver feed solutions.

  9. Surveillance system using the CCTV at the fuel transfer pond in the Tokai reprocessing plant

    SciTech Connect

    Hayakawa, T.; Fukuhara, J.; Ochiai, K.; Ohnishi, T.; Ogata, Y.; Okamoto, H. )

    1991-01-01

    The Fuel Transfer Pond (FTP) in the Tokai Reprocessing Plant (TRP) is a strategic point for safeguards. Spent fuels, therefore, in the FTP have been surveyed by the surveillance system using the underwater CCTV. This system was developed through the improvement of devices composed of cameras and VCRs and the provision of tamper resistance function as one of the JASPAS (Japan Support Program for Agency Safeguards) program. The purpose of this program is to realize the continuous surveillance of the slanted tunnel through which the spent fuel on the conveyor is moved from the FTP to the Mechanical Processing Cell (MPC). This paper reports that, when this surveillance system is applied to an inspection device, the following requirements are needed: To have the ability of continuous and unattended surveillance of the spent fuel on the conveyor path from the FTP to the MPC; To have the tamper resistance function for continuous and unattended surveillance of the spent fuel.

  10. Export control guide: Spent nuclear fuel reprocessing and preparation of plutonium metal

    SciTech Connect

    1993-10-01

    The international Treaty on the Non-Proliferation of Nuclear Weapons, also referred to as the Non-Proliferation Treaty (NPT), states in Article III, paragraph 2(b) that {open_quotes}Each State Party to the Treaty undertakes not to provide . . . equipment or material especially designed or prepared for the processing, use or production of special fissionable material to any non-nuclear-weapon State for peaceful purposes, unless the source or special fissionable material shall be subject to the safeguards required by this Article.{close_quotes} This guide was prepared to assist export control officials in the interpretation, understanding, and implementation of export laws and controls relating to the international Trigger List for irradiated nuclear fuel reprocessing equipment, components, and materials. The guide also contains information related to the production of plutonium metal. Reprocessing and its place in the nuclear fuel cycle are described briefly; the standard procedure to prepare metallic plutonium is discussed; steps used to prepare Trigger List controls are cited; descriptions of controlled items are given; and special materials of construction are noted. This is followed by a comprehensive description of especially designed or prepared equipment, materials, and components of reprocessing and plutonium metal processes and includes photographs and/or pictorial representations. The nomenclature of the Trigger List has been retained in the numbered sections of this document for clarity.

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

    SciTech Connect

    Not Available

    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.

  12. On-Line Monitoring for Control and Safeguarding of Radiochemical Streams at Spent Fuel Reprocessing Plant

    SciTech Connect

    Bryan, Samuel A.; Levitskaia, Tatiana G.; Billing, Justin M.; Casella, Amanda J.; Johnsen, Amanda M.; Peterson, James M.

    2009-10-06

    Advanced techniques enabling enhanced safeguarding of the spent fuel reprocessing plants are urgently needed. Our approach is based on prerequisite that real time monitoring of the solvent extraction flowsheets provides unique capability to quickly detect unwanted manipulations with fissile isotopes present in the radiochemical streams during reprocessing activities. The methods used to monitor these processes must be robust and must be able to withstand harsh radiation and chemical environments. A new on-line monitoring system satisfying these requirements and featuring Raman spectroscopy combined with a Coriolis and conductivity probes, has been recently developed by our research team. It provides immediate chemical data and flow parameters of high-level radioactive waste streams with high brine content generated during retrieval activities from Hanford nuclear waste storage tanks. The nature of the radiochemical streams at the spent fuel reprocessing plant calls for additional spectroscopic information, which can be gained by the utilization of UV-vis-NIR capabilities. Raman and UV-vis-NIR spectroscopies are analytical techniques that have extensively been extensively applied for measuring the various organic and inorganic compounds including actinides. The corresponding spectrometers used under the laboratory conditions are easily convertible to the process-friendly configurations allowing remote measurements under the flow conditions. A fiber optic Raman probe allows monitoring of the high concentration species encountered in both aqueous and organic phases within the UREX suite of flowsheets, including metal oxide ions, such as uranyl, components of the organic solvent, inorganic oxo-anions, and water. The actinides and lanthanides are monitored remotely by UV-vis-NIR spectroscopy in aqueous and organic phases. In this report, we will present our recent results on spectroscopic measurements of simulant flowsheet solutions and commercial fuels available at

  13. Dynamic considerations in the development of centrifugal separators used for reprocessing nuclear fuel

    SciTech Connect

    Strunk, W.D.; Singh, S.P.; Tuft, R.M.

    1988-01-01

    The development of centrifugal separators has been a key ingredient in improving the process used for reprocessing of spent nuclear fuel. The separators are used to segregate uranium and plutonium from the fission products produced by a controlled nuclear reaction. The separators are small variable speed centrifuges, designed to operate in a harsh environment. Dynamic problems were detected by vibration analysis and resolved using modal analysis and trending. Problems with critical speeds, resonances in the base, balancing, weak components, precision manufacturing, and short life have been solved.

  14. An evaluation of retention and disposal options for tritium in fuel reprocessing

    SciTech Connect

    Benjamin, R.W.; Hampson, D.C.

    1987-12-31

    This report assesses the possible options for retention of tritium and its ultimate disposal during future reprocessing of irradiated oxide fuels discharged from light water reactors (LWRs) and liquid metal fast breeder reactors (LMFBRs). The assessment includes an appraisal of the state of the retention and disposal options, an estimate of the dose commitments to the general public, an estimation of the incremental costs of the several retention and disposal options, and the potential reduction of the dose commitments resulting from retention and disposal of the tritium. The assessment is based upon an extensive study of tritium retention in reprocessing completed in 1982 by Grimes et al. Two plants were assumed, one to process LWR oxide fuel and the other to process LMFBR fuel. In each base case plant the tritium was vaporized to the atmosphere. Each of the hypothetical plants was assumed to be constructed during the 1990`s and to operate for a 20-year lifetime beginning in the year 2000 at a rate of 1,500 metric tons of heavy metal (MTHM) per 300-d year. In addition to the base case (Case 1), six other cases which included tritium retention options were examined. Although many of the features of the base-case plants remain unchanged in the tritium retention options, each case requires some additions, deletions, and modifications of portions of the plants. The retained tritium must also be managed and disposed of in a manner that is environmentally acceptable.

  15. Container for reprocessing and permanent storage of spent nuclear fuel assemblies

    DOEpatents

    Forsberg, C.W.

    1992-03-24

    A single canister process container is described for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining its integrity at a temperature necessary to oxidize the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container. 10 figs.

  16. Method for cleaning solution used in nuclear fuel reprocessing

    DOEpatents

    Tallent, Othar K.; Crouse, David J.; Mailen, James C.

    1982-01-01

    Nuclear fuel processing solution consisting of tri-n-butyl phosphate and dodecane, with a complex of uranium, plutonium, or zirconium and with a solvent degradation product such as di-n-butyl phosphate therein, is contacted with an aqueous solution of a salt formed from hydrazine and either a dicarboxylic acid or a hydroxycarboxylic acid, thereby removing the aforesaid complex from the processing solution.

  17. Method for cleaning solution used in nuclear fuel reprocessing

    DOEpatents

    Tallent, O.K.; Crouse, D.J.; Mailen, J.C.

    1980-12-17

    Nuclear fuel processing solution consisting of tri-n-butyl phosphate and dodecane, with a complex of uranium, plutonium, or zirconium and with a solvent degradation product such as di-n-butyl phosphate therein, is contacted with an aqueous solution of a salt formed from hydrazine and either a dicarboxylic acid or a hydroxycarboxylic acid, thereby removing the aforesaid complex from the processing solution.

  18. Reprocessing of nuclear fuels at the Savannah River Plant

    SciTech Connect

    Gray, L.W.

    1986-10-04

    For more than 30 years, the Savannah River Plant (SRP) has been a major supplier of nuclear materials such as plutonium-239 and tritium-3 for nuclear and thermonuclear weapons, plutonium-238 for space exploration, and isotopes of americium, curium, and californium for use in the nuclear research community. SRP is a complete nuclear park, providing most of the processes in the nuclear fuel cycle. Key processes involve fabrication and cladding of the nuclear fuel, target, and control assemblies; rework of heavy water for use as reactor moderator; reactor loading, operation, and unloading; chemical recovery of the reactor transmutation products and spent fuels; and management of the gaseous, liquid, and solid nuclear and chemical wastes; plus a host of support operations. The site's history and the key processes from fabrication of reactor fuels and targets to finishing of virgin plutonium for use in the nuclear weapons complex are reviewed. Emphasis has been given to the chemistry of the recovery and purification of weapons grade plutonium from irradiated reactor targets.

  19. Consolidated fuel reprocessing program: Criticality experiments with fast test reactor fuel pins in an organic moderator

    SciTech Connect

    Bierman, S.R.

    1986-12-01

    The results obtained in a series of criticality experiments performed as part of a joint program on criticality data development between the United States Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan are presented in this report along with a complete description of the experiments. The experiments involved lattices of Fast Test Reactor (FTR) fuel pins in an organic moderator mixture similar to that used in the solvent extraction stage of fuel reprocessing. The experiments are designed to provide data for direct comparison with previously performed experimental measurements with water moderated lattices of FTR fuel pins. The same lattice arrangements and FTR fuel pin types are used in these organic moderated experimental assemblies as were used in the water moderated experiments. The organic moderator is a mixture of 38 wt % tributylphosphate in a normal paraffin hydrocarbon mixture of C{sub 11}H{sub 24} to C{sub 15}H{sub 32} molecules. Critical sizes of 1054.8, 599.2, 301.8, 199.5 and 165.3 fuel pins were obtained respectively for organic moderated lattices having 0.761 cm, 0.968 cm, 1.242 cm, 1.537 cm and 1.935 cm square lattice pitches as compared to 1046.9, 571.9, 293.9, 199.7 and 165.1 fuel pins for the same lattices water moderated.

  20. On-Line Monitoring and Control of Radiochemical Streams at Spent Fuel Reprocessing Plant

    SciTech Connect

    Levitskaia, Tatiana G.; Bryan, Samuel A.

    2008-05-23

    Techniques are needed to provide on-line monitoring and control of the radiochemical processes that are being developed and demonstrated under the Global Nuclear Energy Partnership (GNEP) initiative. The instrumentation used to monitor these processes must be robust and must be able to withstand harsh radiation and chemical environments. A new on-line monitoring system satisfying these requirements featuring Raman spectroscopy combined with a Coriolis and conductivity probes, has been recently developed by our research team. It provides immediate chemical data and flow parameters of high-level radioactive waste streams with high brine/high alkalinity generated during retrieval from Hanford nuclear waste storage tanks. We are currently applying similar methodology for monitoring the radiochemical streams generated at the spent fuel reprocessing plant. The nature of these strems calls for additional spectroscopic information, which can be gained by the utilization of UV-vis-NIR capabilities.

  1. The use of artificial intelligence for safeguarding fuel reprocessing plants

    SciTech Connect

    Wachter, J.W.; Forgy, C.L.

    1987-01-01

    Recorded process data from the ''Minirun'' campaigns conducted at the Barnwell Nuclear Fuel Plant (BNFP) in Barnwell, South Carolina during 1980 to 1981 have been utilized to study the suitability of computer-based Artificial Intelligence (AI) methods for process monitoring for safeguards purposes. The techniques of knowledge engineering were used to formulate the decision-making software which operates on the process data customarily used for process operations. The OPS5 AI language was used to construct an Expert System for this purpose. Such systems are able to form reasoned conclusions from incomplete, inaccurate or otherwise ''fuzzy'' data, and to explain the reasoning that led to them. The programs were tested using minirun data taken during simulated diversions ranging in size from 1 to 20 L of solution that had been monitored previously using conventional procedural techniques. 13 refs., 3 figs.

  2. Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

    SciTech Connect

    Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.; Baryshnikov, M.V.; Kryukov, O.V.; Khaperskaya, A.V.

    2013-07-01

    REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

  3. Method for cleaning solution used in nuclear fuel reprocessing

    DOEpatents

    Tallent, Othar K.; Dodson, Karen E.; Mailen, James C.

    1983-01-01

    A nuclear fuel processing solution containing (1) hydrocarbon diluent, (2) tri-n-butyl phosphate or tri-2-ethylhexyl phosphate, and (3) monobutyl phosphate, dibutyl phosphate, mono-2-ethylhexyl phosphate, di-2-ethylhexyl phosphate, or a complex formed by plutonium, uranium, or a fission product thereof with monobutyl phosphate, dibutyl phosphate, mono-2-ethylhexyl phosphate, or di-2-ethylhexyl phosphate is contacted with silica gel having alkali ions absorbed thereon to remove any one of the degradation products named in section (3) above from said solution.

  4. Chemical Forms and Distribution of Platinum Group Metals and Technetium During Spent Fuel Reprocessing

    SciTech Connect

    Pokhitonov, Y.

    2007-07-01

    Amongst the fission products present in spent nuclear fuel of Nuclear Power Plants there are considerable quantities of platinum group metals (PGMs): ruthenium, rhodium and palladium. At the same time there are considerable amounts of technetium in the spent fuel, the problem of its removal at radiochemical plants being in operation encountering serious difficulties. Increased interest in this radionuclides is due not only to its rather large yield, but to higher mobility in the environment as well. However, the peculiarities of technetium chemistry in nitric acid solutions create certain problems when trying to separate it as a single product in the course of NPP's spent fuel reprocessing. The object of this work was to conduct a comprehensive analysis of platinum group metals and technetium behavior at various stages of spent fuel reprocessing and to seek the decisions which could make it possible to separate its as a single product. The paper will report data on platinum metals (PGM) and technetium distribution in spent fuel reprocessing products. The description of various techniques for palladium recovery from differing in composition radioactive solutions arising from reprocessing is given. (authors)

  5. A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

    SciTech Connect

    Parker, Frank L.

    2012-07-01

    Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storage sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions), long

  6. Component failure-rate data with potential applicability to a nuclear fuel reprocessing plant

    SciTech Connect

    Dexter, A.H.; Perkins, W.C.

    1982-07-01

    Approximately 1223 pieces of component failure-rate data, under 136 subject categories, have been compiled from published literature and computer searches of a number of data bases. Component selections were based on potential applicability to facilities for reprocessing spent nuclear fuels. The data will be useful in quantifying fault trees for probabilistic safety analyses and risk assessments.

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

    SciTech Connect

    Paviet-Hartmann, P.; Riddle, C.; Campbell, K.; Mausolf, E.

    2013-07-01

    The most widely used reductant to partition plutonium from uranium in the Purex process was ferrous sulfamate, other alternates were proposed such as hydrazine-stabilized ferrous nitrate or uranous nitrate, platinum catalyzed hydrogen, and hydrazine, hydroxylamine salts. New candidates to replace hydrazine or hydroxylamine nitrate (HAN) are pursued worldwide. They may improve the performance of the industrial Purex process towards different operations such as de-extraction of plutonium and reduction of the amount of hydrazine which will limit the formation of hydrazoic acid. When looking at future recycling technologies using hydroxamic ligands, neither acetohydroxamic acid (AHA) nor formohydroxamic acid (FHA) seem promising because they hydrolyze to give hydroxylamine and the parent carboxylic acid. Hydroxyethylhydrazine, HOC{sub 2}H{sub 4}N{sub 2}H{sub 3} (HEH) is a promising non-salt-forming reductant of Np and Pu ions because it is selective to neptunium and plutonium ions at room temperature and at relatively low acidity, it could serve as a replacement of HAN or AHA for the development of a novel used nuclear fuel recycling process.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

  10. Assessing the effectiveness of safeguards at a medium-sized spent-fuel reprocessing facility

    SciTech Connect

    Higinbotham, W.; Fishbone, L.G.; Suda, S.

    1983-01-01

    In order to evaluate carefully and systematically the effectiveness of safeguards at nuclear-fuel-cycle facilities, the International Atomic Energy Agency has adopted a safeguards effectiveness assessment methodology. The methodology has been applied to a well-characterized, medium-sized, spent-fuel reprocessing plant to understand how explicit safeguards inspection procedures would serve to expose conceivable nuclear materials diversion schemes, should such diversion occur.

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

    SciTech Connect

    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.

  12. Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant

    SciTech Connect

    Durant, W.S.; Perkins, W.C.; Lee, R.; Stoddard, D.H.

    1982-05-20

    The Safety Technology Group is developing methodology that can be used to assess the risk of operating a plant to reprocess spent nuclear fuel. As an early step in the methodology, a preliminary hazards analysis identifies safety-related incidents. In the absence of appropriate safety features, these incidents could lead to significant consequences and risk to onsite personnel or to the public. This report is a compilation of potential safety-related incidents that have been identified in studies at SRL and in safety analyses of various commercially designed reprocessing plants. It is an expanded revision of the version originally published as DP-1558, Published December 1980.

  13. Spent Fuel Reprocessing: More Value for Money Spent in a Geological Repository?

    SciTech Connect

    Kaplan, P.; Vinoche, R.; Devezeaux, J-G.; Bailly, F.

    2003-02-25

    Today, each utility or country operating nuclear power plants can select between two long-term spent fuel management policies: either, spent fuel is considered as waste to dispose of through direct disposal or, spent fuel is considered a resource of valuable material through reprocessing-recycling. Reading and listening to what is said in the nuclear community, we understand that most people consider that the choice of policy is, actually, a choice among two technical paths to handle spent fuel: direct disposal versus reprocessing. This very simple situation has been recently challenged by analysis coming from countries where both policies are on survey. For example, ONDRAF of Belgium published an interesting study showing that, economically speaking for final disposal, it is worth treating spent fuel rather than dispose of it as a whole, even if there is no possibility to recycle the valuable part of it. So, the question is raised: is there such a one-to-one link between long term spent fuel management political option and industrial option? The purpose of the presentation is to discuss the potential advantages and drawbacks of spent fuel treatment as an implementation of the policy that considers spent fuel as waste to dispose of. Based on technical considerations and industrial experience, we will study qualitatively, and quantitatively when possible, the different answers proposed by treatment to the main concerns of spent-fuel-as-a-whole geological disposal.

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

    PubMed

    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

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

    DOEpatents

    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).

  16. Method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions

    DOEpatents

    Horwitz, E. Philip; Delphin, Walter H.

    1979-07-24

    A method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions containing these and other values by contacting the waste solution with an extractant of tricaprylmethylammonium nitrate in an inert hydrocarbon diluent which extracts the palladium and technetium values from the waste solution. The palladium and technetium values are recovered from the extractant and from any other coextracted values with a strong nitric acid strip solution.

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

    SciTech Connect

    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.

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

    SciTech Connect

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

    2003-02-25

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

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

    SciTech Connect

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

    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

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

    SciTech Connect

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

    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.

  1. Effect of changes in DOE pricing policies for enrichment and reprocessing on research reactor fuel cycle costs

    SciTech Connect

    Matos, J.E.; Freese, K.E.

    1986-11-03

    Fuel cycle costs with HEU and LEU fuels for the IAEA generic 10 MW reactor are updated to reflect the change in DOE pricing policy for enrichment services as of October 1985 and the published charges for LEU reprocessing services as of February 1986. The net effects are essentially no change in HEU fuel cycle costs and a reduction of about 8 to 10% in the fuel cycle costs for LEU silicide fuel.

  2. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  3. Sensitivity Analysis of Reprocessing Cooling Times on Light Water Reactor and Sodium Fast Reactor Fuel Cycles

    SciTech Connect

    R. M. Ferrer; S. Bays; M. Pope

    2008-04-01

    The purpose of this study is to quantify the effects of variations of the Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) and fast reactor reprocessing cooling time on a Sodium Fast Reactor (SFR) assuming a single-tier fuel cycle scenario. The results from this study show the effects of different cooling times on the SFR’s transuranic (TRU) conversion ratio (CR) and transuranic fuel enrichment. Also, the decay heat, gamma heat and neutron emission of the SFR’s fresh fuel charge were evaluated. A 1000 MWth commercial-scale SFR design was selected as the baseline in this study. Both metal and oxide CR=0.50 SFR designs are investigated.

  4. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    DOEpatents

    Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

    2013-11-05

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  5. 76 FR 24494 - Draft Guidance for Industry and FDA Staff: Processing/Reprocessing Medical Devices in Health Care...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-02

    .../ Reprocessing Medical Devices in Health Care Settings: Validation Methods and Labeling; Availability AGENCY... Staff: Processing/Reprocessing Medical Devices in Health Care Settings: Validation Methods and Labeling... ``Draft Guidance for Industry and FDA Staff: Processing/Reprocessing Medical Devices in Health...

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

  8. Novel sorbent development and evaluation for the capture of krypton and xenon from nuclear fuel reprocessing off-gas stream

    SciTech Connect

    Garn, T.G.; Greenhalgh, M.R.; Law, J.D.

    2013-07-01

    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, Idaho National Laboratory 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. (authors)

  9. Process centrifuge operating problems and equipment failures in canyon reprocessing facilities at the Savannah River Site

    SciTech Connect

    Durant, W.S.; Baughman, D.F.

    1990-03-01

    The Savannah River Laboratory (SRL) maintains a compilation of operating problems and equipment failures that have occurred in the fuel reprocessing areas of the Savannah River Site (SRS). At present, the data bank contains more than 230,000 entries ranging from minor equipment malfunctions to incidents with the potential for injury or contamination of personnel, or for economic loss. The data bank has been used extensively for a wide variety of purposes, such as failure analyses, trend analyses, and preparation of safety analyses. Typical of the data are problems associated with the canyon process centrifuges. This report contains a compilation of the centrifuge operating problems and equipment failures primarily as an aid to organizations with related equipment. Publication of these data was prompted by a number of requests for this information by other Department of Energy (DOE) sites. 11 refs., 2 figs., 4 tabs.

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

    NASA Astrophysics Data System (ADS)

    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.

  11. Principles of qualification of the PAMELA process for the vitrification of HLLW of the Karlsruhe Reprocessing Plant (WAK)

    SciTech Connect

    Ewest, E.; Kunz, W.; Demonie, M.; Martens, B.R.; Goeyse, M. de

    1993-12-31

    After having reprocessed about 211 t of Uranium, the WAK Karlsruhe Pilot Reprocessing Plant was shut down in 1991. While all the other radioactive waste arising from reprocessing were conditioned parallel to the plant operation, some 60 m{sup 3} of High Level Liquid Waste (HLLW) having a specific {beta}, {gamma}-activity of about 2 E13 Bq/l is not yet processed. The waste is stored in two tanks, having a different activity level and chemical composition. In order to obtain a uniform product both solutions will be blended in a suitable way. It is intended to ship this waste to the PAMELA Vitrification Plant located on the Belgoprocess (BP) site in Dessel, Belgium. The vitrified product shall be returned to Germany. As from October 1986 until September 1991, the facility was operated by a mixed Belgian-German crew under the responsibility of BP for the vitrification of 800 m{sup 3} of HEWC (concentrated high-level waste from the reprocessing of high-enriched uranium fuels). Between October 1, 1985 and September 1, 1991, the total amount of 907 m{sup 3} of EUROCHEMIC HLLW has been successfully vitrified and conditioned in about 2,200 canisters. The typical composition of the different types of glass products are compared with the design data of the WAK glass product.

  12. A novel waste form for disposal of spent-nuclear-fuel reprocessing waste: A vitrifiable cement

    SciTech Connect

    Gougar, M.L.D.; Scheetz, B.E.; Siemer, D.D.

    1999-01-01

    A cement capable of being hot isostatically pressed into a glass ceramic has been proposed as the waste form for spent-nuclear-fuel reprocessing wastes at the Idaho National Engineering and Environmental Laboratory (INEEL). This intermediate cement, with a composition based on that of common glasses, has been designed and tested. The cement formulations included mixed INEEL wastes, blast furnace slag, reactive silica, and INEEL soil or vermiculite, which were activated with potassium or sodium hydroxide. Following autoclave processing, the cements were characterized. X-ray diffraction analysis revealed three notable crystalline phases: quartz, calcite, and fluorite. Results of compressive strength testing ranged from 1452 and 4163 psi, exceeding the US Nuclear Regulatory Commission (NRC)-suggested standard of >500 psi. From American National Standards Institute/American Nuclear Society 16.1-1986 leach testing, effective diffusivities for Cs were determined to be on the order of 10{sup {minus}11} to 10{sup {minus}10} cm{sup 2}/s and for Sr were 10{sup {minus}12} cm{sup 2}/s, which are four orders of magnitude less than diffusivities in some other radwaste materials. Average leach indices (LI) were 9.6 and 11.9 for Cs and Sr, respectively, meeting the NRC Standard of LI > 6. The 28-day Materials Characterization Center-1 leach testing resulted in normalized elemental mass losses between 0.63 and 28 g/(m{sup 2}{center_dot}day) for Cs and between 0.34 and 0.70 g/(m{sup 2}{center_dot}day) industry-accepted standard while Cs losses indicate a process sensitive parameter.

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

    DOEpatents

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

    1982-01-19

    A method and apparatus for separating and reprocessing spent nuclear fuels includes a separation vessel housing a molten metal solvent in a reaction region, a reflux region positioned above and adjacent to the reaction region, and a porous filter member defining the bottom of the separation vessel in a supporting relationship with the metal solvent. Spent fuels are added to the metal solvent. A nonoxidizing nitrogen-containing gas is introduced into the separation vessel, forming solid actinide nitrides in the metal solvent from actinide fuels, while leaving other fission products in solution. A pressure of about 1.1 to 1.2 atm is applied in the reflux region, forcing the molten metal solvent and soluble fission products out of the vessel, while leaving the solid actinide nitrides in the separation vessel.

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

    NASA Astrophysics Data System (ADS)

    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.

  15. Potential radiological impact of tornadoes on the safety of Nuclear Fuel Services' West Valley Fuel Reprocessing Plant. 2. Reentrainment and discharge of radioactive materials

    SciTech Connect

    Davis, W Jr

    1981-07-01

    This report describes results of a parametric study of quantities of radioactive materials that might be discharged by a tornado-generated depressurization on contaminated process cells within the presently inoperative Nuclear Fuel Services' (NFS) fuel reprocessing facility near West Valley, New York. The study involved the following tasks: determining approximate quantities of radioactive materials in the cells and characterizing particle-size distribution; estimating the degree of mass reentrainment from particle-size distribution and from air speed data presented in Part 1; and estimating the quantities of radioactive material (source term) released from the cells to the atmosphere. The study has shown that improperly sealed manipulator ports in the Process Mechanical Cell (PMC) present the most likely pathway for release of substantial quantities of radioactive material in the atmosphere under tornado accident conditions at the facility.

  16. 10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... property. 2. A fuel reprocessing plant's inventory of high-level liquid radioactive wastes will be limited... requirements of 10 CFR part 71. The dry solid shall be chemically, thermally, and radiolytically stable to the... 10 Energy 1 2011-01-01 2011-01-01 false Policy Relating to the Siting of Fuel Reprocessing...

  17. 10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... property. 2. A fuel reprocessing plant's inventory of high-level liquid radioactive wastes will be limited... requirements of 10 CFR part 71. The dry solid shall be chemically, thermally, and radiolytically stable to the... 10 Energy 1 2013-01-01 2013-01-01 false Policy Relating to the Siting of Fuel Reprocessing...

  18. 76 FR 34007 - Draft Regulatory Basis for a Potential Rulemaking on Spent Nuclear Fuel Reprocessing Facilities

    Federal Register 2010, 2011, 2012, 2013, 2014

    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... regulatory basis for licensing commercial reprocessing facilities: (1) Regulatory framework, (2)...

  19. Advantages of co-located spent fuel reprocessing, repository and underground reactor facilities

    SciTech Connect

    Mahar, James M.; Kunze, Jay F.; Wes Myers, Carl; Loveland, Ryan

    2007-07-01

    The purpose of this work is to extend the discussion of potential advantages of the underground nuclear park (UNP) concept by making specific concept design and cost estimate comparisons for both present Generation III types of reactors and for some of the modular Gen IV or the GNEP modular concept. For the present Gen III types, we propose co-locating reprocessing and (re)fabrication facilities along with disposal facilities in the underground park. The goal is to determine the site costs and facility construction costs of such a complex which incorporates the advantages of a closed fuel cycle, nuclear waste repository, and ultimate decommissioning activities all within the UNP. Modular power generation units are also well-suited for placement underground and have the added advantage of construction using current and future tunnel boring machine technology. (authors)

  20. High corrosion resistant Ti-5%Ta-1.8%Nb alloy for fuel reprocessing application

    NASA Astrophysics Data System (ADS)

    Kapoor, K.; Kain, Vivekanand; Gopalkrishna, T.; Sanyal, T.; De, P. K.

    2003-10-01

    The conventional low carbon austenitic stainless steels display good corrosion resistance behaviour in nitric acid media. However, they are sensitive to intergranular corrosion in boiling nitric acid media in the presence of oxidizing ions like hexavalent chromium, tetravalent iron and hexavalent plutonium. The Ti-5%Ta-1.8%Nb alloy was evaluated as a candidate material for such applications of nuclear fuel reprocessing. Extensive tests were carried out to establish the superior corrosion properties in comparison to the conventional stainless steel or nitric acid grade stainless steel. The fabricability of this new alloy to various shapes like rod, sheet, wire and its weldability, which is required for making vessels, was found to be good.

  1. Use of the Waste-Incidental-to-Reprocessing Citation Process at the West Valley Demonstration Project - 12250

    SciTech Connect

    Sullivan, Dan; Suttora, Linda; Goldston, Sonny; Petras, Robert; Rowell, Laurene; McNeil, Jim

    2012-07-01

    The West Valley Demonstration Project recently achieved a breakthrough in management of radioactive waste from reprocessing of spent nuclear fuel by taking advantage of lessons learned at other Department of Energy (DOE) sites in implementation of the waste-incidental-to-reprocessing citation process of DOE Manual 435.1-1, Radioactive Waste Management. This breakthrough involved a revision to the site procedure on waste-incidental to reprocessing. This procedure revision served as the basis for a determination by the DOE West Valley field office using the citation process that three secondary waste streams consisting of equipment that had once been contaminated by association with HLW are not HLW following decontamination and may be disposed of as low-level waste (LLW) or transuranic waste. These waste streams, which comprised much of the approximately 380 cubic meters of West Valley waste contaminated by association with HLW, included several vessels and certain tank farm equipment. By making use of lessons learned in use of the citation process by other DOE sites and information developed to support use of the citation process at the Hanford site and the Savannah River Site, the team developed a technical basis for showing that use of the citation process of DOE Manual 435.1-1 for the three new waste stream was appropriate and technically justified. The Waste Management Working Group of the EFCOG assisted in transferring lessons learned by drawing on experience from around the DOE complex. This process shared knowledge about effective implementation of the citation process in a manner that proved to be beneficial to the West Valley Demonstration Project and resulted in a technical basis document that could be used to determine that the three new waste streams were not HLW. (authors)

  2. Study on Gaseous Effluent Treatment for Dissolution Step of Spent Nuclear Fuel Reprocessing

    SciTech Connect

    Mineo, H.; Iizuka, M.; Fujisaki, S.; Hotoku, S.; Asakura, T.; Uchiyama, G.

    2002-02-27

    Behavior of radioiodine and carbon-14 during spent fuel dissolution was studied in a bench-scale reprocessing test rig where 29 and 44 GWdt-1 spent fuels were respectively dissolved. Decontamination factor of AGS (silica-gel impregnated with silver nitrate) column for iodine-129 removal was measured to be more than 36,000. The measurement of iodine-129 profile in the adsorption column showed that the nuclide was effectively trapped by the adsorbent. Measurement of iodine-129 in the dissolver solution after the iodine-stripping operation using NO2 gas at 363 K, revealed that less than 0.57% of total iodine-129 generated, which was estimated by ORIGEN II calculation, was remained in the dissolver solution. Also, measurement of iodine-129 by an iodine-stripping operation from the dissolver solution using potassium iodate showed that another 2.72% of total iodine-129 precipitated as iodide. In addition, about 70 % of total iodine generated was measured in the AGS columns. Rest of iodine-129 was supposed to adsorb to a HEPA filter and the inner surface of dissolver off-gas lines. Those results on iodine-129 distribution were found to be almost identical to the results obtained in the study using iodine-131 as tracer and the results reported by other works. It was demonstrated that the two-steps iodine-stripping method using potassium iodate could expel additional iodine from the solution, more effectively than iodine-stripping operation using NO2 gas. Iodine-131 was also detected on the AGS columns at the spent fuel dissolution. Increasing burnup showed larger amount of iodine-131 since amount of curium-244 contained in the spent fuel increased with the burnup. Release of carbon-14 as carbon dioxide during dissolution was found to occur when the release of krypton-85. From the 14CO2 measurement, initial nitrogen-14 concentration in the fuel was estimated to be about several ppm, which was within the range reported.

  3. Breeder Reprocessing Engineering Test

    SciTech Connect

    Burgess, C.A.; Meacham, S.A.

    1984-01-01

    The Breeder Reprocessing Engineering Test (BRET) is a developmental activity of the US Department of Energy to demonstrate breeder fuel reprocessing technology while closing the fuel cycle for the Fast Flux Test Facility (FFTF). It will be installed in the existing Fuels and Materials Examination Facility (FMEF) at the Hanford Site near Richland, Washington, The major objectives of BRET are: (1) close the US breeder fuel cycle; (2) develop and demonstrate reprocessing technology and systems for breeder fuel; (3) provide an integrated test of breeder reactor fuel cycle technology - rprocessing, safeguards, and waste management. BRET is a joint effort between the Westinghouse Hanford Company and Oak Ridge National Laboratory. 3 references, 2 figures.

  4. Peculiarities of highly burned-up NPP SNF reprocessing and new approach to simulation of solvent extraction processes

    SciTech Connect

    Fedorov, Y.S.; Zilberman, B.Y.; Goletskiy, N.D.; Puzikov, E.A.; Ryabkov, D.V.; Rodionov, S.A.; Beznosyuk, V.I.; Petrov, Y.Y.; Saprykin, V.F.; Murzin, A.A.; Bibichev, B.A.; Aloy, A.S.; Kudinov, A.S.; Blazheva, I.V.; Kurenkov, N.V.

    2013-07-01

    Substantiation, general description and performance characteristics of a reprocessing flowsheet for WWER-1000 spent fuel with burn-up >60 GW*day/t U is given. Pu and U losses were <0.1%, separation factor > 10{sup 4}; their decontamination factor from γ-emitting fission products was 4*10{sup 4} and 3*10{sup 7}, respectively. Zr, Tc, Np removal was >98% at U and Pu losses <0.05%. A new approach to simulation of extraction equilibrium has been developed. It is based on a set of simultaneous chemical reactions characterized by apparent concentration constants. A software package was created for simulation of spent fuel component distribution in multistage countercurrent extraction processes in the presence of salting out agents. (authors)

  5. Process monitoring for reprocessing plant safeguards: a summary review

    SciTech Connect

    Kerr, H.T.; Ehinger, M.H.; Wachter, J.W.; Hebble, T.L.

    1986-10-01

    Process monitoring is a term typically associated with a detailed look at plant operating data to determine plant status. Process monitoring has been generally associated with operational control of plant processes. Recently, process monitoring has been given new attention for a possible role in international safeguards. International Safeguards Project Office (ISPO) Task C.59 has the goal to identify specific roles for process monitoring in international safeguards. As the preliminary effort associated with this task, a review of previous efforts in process monitoring for safeguards was conducted. Previous efforts mentioned concepts and a few specific applications. None were comprehensive in addressing all aspects of a process monitoring application for safeguards. This report summarizes the basic elements that must be developed in a comprehensive process monitoring application for safeguards. It then summarizes the significant efforts that have been documented in the literature with respect to the basic elements that were addressed.

  6. Light water reactor fuel reprocessing: dissolution studies of voloxidized and nonvoloxidized fuel

    SciTech Connect

    Johnson, D.R.; Stone, J.A.

    1980-04-01

    Small-scale tests with irradiated Zircaloy-clad fuels from Robinson, Oconee, Saxton, and Point Beach reactors with burnups from about 200 to 28,000 MWD/MTHM have been made to determine the dissolution behavior of both voloxidized (U{sub 3}O{sub 8}) and nonvoloxidized (UO{sub 2}) fuel. No significant technical problems were encountered in batch-dissolving of either form. Dissolution rates were well-controlled in all tests. Significant characteristics of U{sub 3}O{sub 8} dissolution that differed from UO{sub 2} dissolution included: (1) reduced tritium and ruthenium ({sup 106}Ru) concentrations in product solutions, (2) increased insoluble noble metal fission product residue (about 2.2X greater), and (3) increased insoluble plutonium in the fission product residue. The insoluble plutonium is easily leached from the residue by 10M HNO{sub 3}. The weight of the fission product residue collected from both U{sub 3}O{sub 8} and UO{sub 2} fuels increased aproximately linearly with fuel burnup. A major fraction (>83%) of the {sup 85}Kr was evolved from U{sub 3}O{sub 8} fuel during dissolution rather than voloxidation. The {sup 85}Kr evolution rate was an appropriate monitor of fuel dissolution rate. Virtually all of the {sup 129}I was evolved by air sparging of the dissolver solution during dissolution. 30 tables, 18 figures.

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

    SciTech Connect

    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.

  8. Fuel gas conditioning process

    DOEpatents

    Lokhandwala, Kaaeid A.

    2000-01-01

    A process for conditioning natural gas containing C.sub.3+ hydrocarbons and/or acid gas, so that it can be used as combustion fuel to run gas-powered equipment, including compressors, in the gas field or the gas processing plant. Compared with prior art processes, the invention creates lesser quantities of low-pressure gas per unit volume of fuel gas produced. Optionally, the process can also produce an NGL product.

  9. DEVELOPMENT OF GLASS COMPOSITIONS TO IMMOBILIZE ALKALI, ALKALINE EARTH, LANTHANIDE AND TRANSITION METAL FISSION PRODUCTS FROM NUCLEAR FUEL REPROCESSING

    SciTech Connect

    Marra, J.; Billings, A.

    2009-06-24

    The Advanced Fuel Cycle Initiative (AFCI) waste management strategy revolves around specific treatment of individual or groups of separated waste streams. A goal for the separations processes is to efficiently manage the waste to be dispositioned as high level radioactive waste. The Advanced Fuel Cycle Initiative (AFCI) baseline technology for immobilization of the lanthanide (Ln) and transition metal fission product (TM) wastes is vitrification into a borosilicate glass. A current interest is to evaluate the feasibility of vitrifying combined waste streams to most cost effectively immobilize the wastes resulting from aqueous fuel reprocessing. Studies showed that high waste loadings are achievable for the Ln only (Option 1) stream. Waste loadings in excess of 60 wt % (on a calcined oxide basis) were demonstrated via a lanthanide borosilicate (LaBS) glass. The resulting glasses had excellent relative durability as determined by the Product Consistency Test (PCT). For a combined Ln and TM waste stream glass (Option 2), noble metal solubility was found to limit waste loading. However, the measured PCT normalized elemental releases for this glass were at least an order of magnitude below that of Environmental Assessment (EA) glass. Current efforts to evaluate the feasibility of vitrifying combined Ln, TM, alkali (Cs is the primary radionuclide of concern) and alkaline earth (Sr is the primary radionuclide of concern) wastes (Option 3) have shown that these approaches are feasible. However, waste loading limitations with respect to heat load (Cs/Sr loading), molybdenum solubility and/or noble metal solubility will likely be realized and must be considered in determining the cost effectiveness of these approaches.

  10. Potential radiological impact of tornadoes on the safety of Nuclear Fuel Services' West Valley Fuel Reprocessing Plant. Volume I. Tornado effects on head-end cell airflow

    SciTech Connect

    Holloway, L.J.; Andrae, R.W.

    1981-09-01

    This report describes results of a parametric study of the impacts of a tornado-generated depressurization on airflow in the contaminated process cells within the presently inoperative Nuclear Fuel Services fuel reprocessing facility near West Valley, NY. The study involved the following tasks: (1) mathematical modeling of installed ventilation and abnormal exhaust pathways from the cells and prediction of tornado-induced airflows in these pathways; (2) mathematical modeling of individual cell flow characteristics and prediction of in-cell velocities induced by flows from step 1; and (3) evaluation of the results of steps 1 and 2 to determine whether any of the pathways investigated have the potential for releasing quantities of radioactively contaminated air from the main process cells. The study has concluded that in the event of a tornado strike, certain pathways from the cells have the potential to release radioactive materials of the atmosphere. Determination of the quantities of radioactive material released from the cells through pathways identified in step 3 is presented in Part II of this report.

  11. [Safe reprocessing of medical devices with a view of the entire process chain. Recommendations of the VDI 5700 guidelines].

    PubMed

    Kraft, M; Wille, F; Attenberger, J; Müller, U

    2014-12-01

    The reprocessing of medical devices for low pathogen or sterile use is in itself potentially risky even though the aim of reprocessing is the avoidance of hygienic or technically functional risks. The methodological principles of risk management for medical devices are described in the standard DIN EN ISO 14971. The recommendations of the Commission for Hospital Hygiene and Infectious Disease Prevention (Kommission für Krankenhaushygiene und Infektionsprävention KRINKO) of the Robert Koch Institute (RKI) and the Federal Institute for Drugs and Medical Devices (Bundesinstituts für Arzneimittel und Medizinprodukte BfArM) "hygiene requirements for the reprocessing of medical devices" clarify numerous reprocessing-specific risks and are structured with reference to the different steps of reprocessing. The aim was a practical combination of the normative risk management methodology with the process-oriented KRINKO/BfArM recommendations, which has provided an interdisciplinary group of experts moderated by the Association of German Engineers (VDI). The main contents of the VDI 5700 guidelines on "hazards associated with the reprocessing--risk management in the reprocessing of medical devices--measures for risk control" and the process of the development of these guidelines is described. PMID:25348217

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

    SciTech Connect

    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.

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

    SciTech Connect

    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

  14. Glutarimidedioxime: A Complexing and Reducing Reagent for Plutonium Recovery from Spent Nuclear Fuel Reprocessing.

    PubMed

    Xian, Liang; Tian, Guoxin; Beavers, Christine M; Teat, Simon J; Shuh, David K

    2016-04-01

    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 (H2 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 HNO3 with H2 A. The complexation-reduction mechanism is illustrated with the combination of UV/Vis absorption spectra and the crystal structure of a Pu(IV) complex with the reagent. The fast stripping rate and the high efficiency for stripping Pu(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. PMID:26970221

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

    PubMed

    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. PMID:27411177

  16. 10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities F Appendix F to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. F Appendix F to Part 50—Policy Relating to the Siting of Fuel...

  17. 10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities F Appendix F to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. F Appendix F to Part 50—Policy Relating to the Siting of Fuel...

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

    SciTech Connect

    Demori, R.; Mauler, R. S.; Ashton, E.; Weschenfelder, V. F.; Cândido, L. H. A.; Kindlein, W.

    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.

  19. Corrosion property of 9Cr-ODS steel in nitric acid solution for spent nuclear fuel reprocessing

    SciTech Connect

    Takeuchi, M.; Koizumi, T.; Inoue, M.; Koyama, S.I.

    2013-07-01

    Corrosion tests of oxide dispersion strengthened with 9% Cr (9Cr-ODS) steel, which is one of the desirable materials for cladding tube of sodium-cooled fast reactors, in pure nitric acid solution, spent FBR fuel solution, and its simulated solution were performed to understand the corrosion behavior in a spent nuclear fuel reprocessing. In this study, the 9Cr-ODS steel with lower effective chromium content was evaluated to understand the corrosion behavior conservatively. As results, the tube-type specimens of the 9Cr-ODS steels suffered severe weight loss owing to active dissolution at the beginning of the immersion test in pure nitric acid solution in the range from 1 to 3.5 M. In contrast, the weight loss was decreased and they showed a stable corrosion in the higher nitric acid concentration, the dissolved FBR fuel solution, and its simulated solution by passivation. The corrosion rates of the 9Cr-ODS steel in the dissolved FBR fuel solution and its simulated solution were 1-2 mm/y and showed good agreement with each other. The passivation was caused by the shift of corrosion potential to noble side owing to increase in nitric acid concentration or oxidative ions in the dissolved FBR fuel solution and the simulated spent fuel solution. (authors)

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

    PubMed

    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. PMID:20022548

  1. Consolidated fuel-reprocessing program. Quarterly progress report for the period ending November 30, 1980

    SciTech Connect

    Carney, H. C.; Pierce, V. H.; Rickman, W. S.; Strand, J. B.; Holder, N. D.; Brimhall, W. L.; Fields, D. E.; Callahan, S. F.; Higuchi, K.; Benedict, G. E.; Abraham, L.; Hirsch, P. M.; Engler, D. R.; Stock, W. C.; Wilbourn, R. G.; Dunlap, S. A.; Hostbjor, G. J.; Olquin, L. J.; Swieda, D. E.; Field, R. E.; Newman, P. W.; Rode, J. S.; Brown, L. C.; Haldy, B. B.

    1980-12-01

    Large-scale crushing and burning tests of fuel received from the Federal Republic of Germany (FRG) have identified differences in processing characteristics. The NO/sub x/ converter experimental tests continued this quarter in the GA off-gas system. The dependence of nitric oxide conversion efficiency on ammonia and oxygen concentrations has been determined. An optimum ratio of ammonia to nitric oxide was identified in terms of the conversion factor for nitrix oxide. No residual ammonia was detected downstream of the NO/sub x/ converter. Design of the radon source subsystem and fabrication of the radon source assembly and shielded radon source containment were completed. Second-thorium-cycle column tests confirm the feasibility of producing thorium product having acceptably low uranium contents. Two axial mixing tests were completed on the thorium extraction section. The Mott inertial filter tests continued. The dissolution rate of HTGR fuel spheres is speeded by a factor of three by using an airlift acid recirculator to improve ThO/sub 2/-dissolvent contact. The FRG HTGR fuel spheres dissolve a factor of three more slowly than the sol-gel derived ThO/sub 2/ spheres. The task group completed a thorium transfer kinetics comparison of alternative solvents. Hydrolysis studies were started on alternative solvents.

  2. Safety research of multi-functional reprocessing process considering nonproliferation based on an ion-exchange method

    SciTech Connect

    Koyama, Shin-ichi; Ozawa, Masaki |; Okada, Ken; Kurosawa, Kiyoko; Suzuki, Tatsuya; Fujii, Yasuhiko

    2007-07-01

    A simplified separation process was proposed based on an ion-exchange technique. A tertiary pyridine-type ion-exchange resin was used in this process to treat the mixed oxide fuel highly irradiated in the experimental fast reactor 'JOYO'. It was demonstrated that the process is a realistic candidate for future reprocessing using hydrochloric acid and a mixed eluent solution of nitric acid and methanol. In order to develop an engineering scale concept, it is indispensable to establish the conditions for safe operation, so two types of experiments were done to obtain fundamental aspects. The corrosion experiment for structural materials in hydrochloric acid at room temperature was done using tantalum, zirconium, niobium, hastelloy and SUS316L. Results showed that tantalum, zirconium, niobium, and hastelloy had good corrosion resistance to hydrochloric acid. The second experiment looked at the thermal hazards of pyridine-type ion-exchange resin and the methanol, or nitric acid eluent system from the viewpoints of fire and explosion safety. No hazardous reactions occurred between the resin and the eluent system. Above 150 deg. C, attention should be paid to the exothermic reactions for the dried resin. (authors)

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

    SciTech Connect

    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.

  4. Plasma techniques for reprocessing nuclear wastes

    SciTech Connect

    Siciliano, E.R.; Lucoff, D.M.; Omberg, R.P.; Walter, A.E.

    1993-06-01

    A newly emerging plasma-based system, currently under development for material dissociation and mass separation applications in the area of high-level radioactive waste treatment, may have possible applications as a central processing unit for spent nuclear fuel reprocessing. Because this system has no moving parts and obtains separations by electromagnetic techniques, it offers a distinct advantage over chemically based separation techniques, in that the total waste volume does not increase. The basic concepts underlying the operation of this plasma-based system are discussed, along with the demonstrated and expected capabilities of this system. Possible fuel reprocessing configurations using this plasma-based technology are also mentioned.

  5. The distribution of 129I around West Valley, an inactive nuclear fuel reprocessing facility in Western New York

    NASA Astrophysics Data System (ADS)

    Rao, Usha; Fehn, Udo

    1997-03-01

    A study of 129I levels in surface waters around an inactive nuclear fuel reprocessing facility at West Valley, Cattaraugus County, New York shows a strong presence of this long-lived radoiisotope (T{1}/{2} = 15.7 Ma) of iodine around the facility. The signal is strong in creeks which drain the facility as well as those in the general vicinity over two decades after reprocessing activities at the site ceased in 1972. Highest 129I levels (1.36 × 1011 atoms/L) are observed at the site boundary in Buttermilk Creek which drains the site, and the resulting plume can be tracked into Lake Erie via Cattaraugus Creek. Other creeks in the West Valley area which do not receive drainage from the site have 129I concentrations on the order of 109-1010 atoms/L, indicating that atmospheric transport of the radionuclide is significant. 129I levels in surface waters around West Valley are 10-1000 times higher than background lelels in western New York, including 129I levels around active nuclear power plants (reported in Rao and Fehn, in preparation), and 100-10000 times higher than levels of 129I in areas outside western New York. However, {36Cl}/{Cl} and 3H measurements in Buttermilk Creek at the site boundary are consistent with present day rainwater values for the region.

  6. The search for advanced remote technology in fast reactor reprocessing

    SciTech Connect

    Burch, W.D.; Herndon, J.N.; Stradley, J.G.

    1990-01-01

    Research and development in fast reactor reprocessing has been under way about 20 years in several countries throughout the world. During the past decade in France and the United Kingdom, active development programs have been carried out in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the EBR-II facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. Germany and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in all of these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper will focus principally on the search for improved facility concepts and better maintenance systems in the CFRP and, in turn, on how developments at ORNL have influenced the technology elsewhere.

  7. DEVELOPMENT OF A HYDROGEN MORDENITE SORBENT FOR THE CAPTURE OF KRYPTON FROM USED NUCLEAR FUEL REPROCESSING OFF-GAS STREAMS

    SciTech Connect

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

    2014-04-01

    A novel new sorbent for the separation of krypton from off-gas streams resulting from the reprocessing of used nuclear fuel has been developed and evaluated. A hydrogen mordenite powder was successfully incorporated into a macroporous polymer binder and formed into spherical beads. The engineered form sorbent retained the characteristic surface area and microporosity indicative of mordenite powder. The sorbent was evaluated for krypton adsorption capacities utilizing thermal swing operations achieving capacities of 100 mmol of krypton per kilogram of sorbent at a temperature of 191 K. A krypton adsorption isotherm was also obtained at 191 K with varying krypton feed gas concentrations. Adsorption/desorption cycling effects were also evaluated with results indicating that the sorbent experienced no decrease in krypton capacity throughout testing.

  8. 10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... requirements of 10 CFR part 71. The dry solid shall be chemically, thermally, and radiolytically stable to the... DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. F Appendix F to Part 50—Policy... property. 2. A fuel reprocessing plant's inventory of high-level liquid radioactive wastes will be...

  9. Progress on Cleaning Up the Only Commercial Nuclear Fuel Reprocessing Facility to Operate in the United States

    SciTech Connect

    Jackson, T. J.; MacVean, S. A.; Szlis, K. A.

    2002-02-26

    This paper describes the progress on cleanup of the West Valley Demonstration Project (WVDP), an environmental management project located south of Buffalo, NY. The WVDP was the site of the only commercial nuclear fuel reprocessing facility to have operated in the United States (1966 to 1972). Former fuel reprocessing operations generated approximately 600,000 gallons of liquid high-level radioactive waste stored in underground tanks. The U.S. Congress passed the WVDP Act in 1980 (WVDP Act) to authorize cleanup of the 220-acre facility. The facility is unique in that it sits on the 3,345-acre Western New York Nuclear Service Center (WNYNSC), which is owned by New York State through the New York State Energy Research and Development Authority (NYSERDA). The U.S. Department of Energy (DOE) has overall responsibility for the cleanup that is authorized by the WVDP Act, paying 90 percent of the WVDP costs; NYSERDA pays 10 percent. West Valley Nuclear Services Company (WVNSCO) is the management contractor at the WVDP. This paper will provide a description of the many accomplishments at the WVDP, including the pretreatment and near completion of vitrification of all the site's liquid high-level radioactive waste, a demonstration of technologies to characterize the remaining material in the high-level waste tanks, the commencement of decontamination and decommissioning (D&D) activities to place the site in a safe configuration for long-term site management options, and achievement of several technological firsts. It will also include a discussion of the complexities involved in completing the WVDP due to the various agency interests that require integration for future cleanup decisions.

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

    PubMed

    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. PMID:25948824

  11. Spent graphite fuel element processing

    SciTech Connect

    Holder, N.D.; Olsen, C.W.

    1981-07-01

    The Department of Energy currently sponsors two programs to demonstrate the processing of spent graphite fuel elements. General Atomic in San Diego operates a cold pilot plant to demonstrate the processing of both US and German high-temperature reactor fuel. Exxon Nuclear Idaho Company is demonstrating the processing of spent graphite fuel elements from Rover reactors operated for the Nuclear Rocket Propulsion Program. This work is done at Idaho National Engineering Laboratory, where a hot facility is being constructed to complete processing of the Rover fuel. This paper focuses on the graphite combustion process common to both programs.

  12. On-line molecular iodine isotopologue detection in gaseous media during spent nuclear fuel reprocessing using a laser-induced fluorescence method

    NASA Astrophysics Data System (ADS)

    Kireev, S. V.; Shnyrev, S. L.

    2015-06-01

    The paper reports on on-line measurement of the {}129{{\\text{I}}2}, 127I129I, and {}127{{\\text{I}}2} concentrations during spent nuclear fuel (SNF) reprocessing using a laser-induced fluorescence method. A He-Ne laser (632.8 nm) was used as a fluorescence excitation source. The detection limits obtained for molecular iodine isotopologue concentrations demonstrate the possibility of using this method for iodine control both in gaseous technological media generated during SNF reprocessing and after passing through the gas purification system (in atmosphere emission).

  13. Effect of high burn-up and MOX fuel on reprocessing, vitrification and disposal of PWR and BWR spent fuels based on accurate burn-up calculation

    SciTech Connect

    Yoshikawa, T.; Iwasaki, T.; Wada, K.; Suyama, K.

    2006-07-01

    To examine the procedures of the reprocessing, the vitrification and the geologic disposal, precise burn-up calculation for high burn-up and MOX fuels has been performed for not only PWR but also BWR by using SWAT and SWAT2 codes which are the integrated bum-up calculation code systems combined with the bum-up calculation code, ORIGEN2, and the transport calculation code, SRAC (the collision probability method) or MVP (the continuous energy Monte Carlo method), respectively. The calculation results shows that all of the evaluated items (heat generation and concentrations of Mo and Pt) largely increase and those significantly effect to the current procedures of the vitrification and the geologic disposal. The calculation result by SWAT2 confirms that the bundle calculation is required for BWR to be discussed about those effects in details, especially for the MOX fuel. (authors)

  14. Fuel processing device

    DOEpatents

    Ahluwalia, Rajesh K.; Ahmed, Shabbir; Lee, Sheldon H. D.

    2011-08-02

    An improved fuel processor for fuel cells is provided whereby the startup time of the processor is less than sixty seconds and can be as low as 30 seconds, if not less. A rapid startup time is achieved by either igniting or allowing a small mixture of air and fuel to react over and warm up the catalyst of an autothermal reformer (ATR). The ATR then produces combustible gases to be subsequently oxidized on and simultaneously warm up water-gas shift zone catalysts. After normal operating temperature has been achieved, the proportion of air included with the fuel is greatly diminished.

  15. Signal transmission techniques for large-scale nuclear fuel reprocessing applications

    SciTech Connect

    Herndon, J.N.; Bible, D.W.

    1985-01-01

    The RCE is currently developing a prototypic microwave-based signal transmission system for reprocessing cell applications. This system, being developed for use in the Advanced Integrated Maintenance System (AIMS), will operate in the 10-GHz frequency range. Provisions are being made for five real-time video channels, three bidirectional data channels at one megabaud data rate each, and two audio channels. The basic utility of the concept has been proven in a laboratory demonstration using gallium arsenide gunn diode transmitter/receivers with horn antennas. Unidirectional transmission of one real-time video channel over a distance of 200 ft was demonstrated. No evidence of multipath interference was detected even when the transmission path was surrounded by metallic reflectors. The microwave signal transmission system for the AIMS application is in final design. Fabrication in the ORNL instrument shops will begin in October 1985, and the system should be operational in the Maintenance Systems Test Area (MSTA) at ORNL in the latter half of 1986.

  16. Monolithic Fuel Fabrication Process Development

    SciTech Connect

    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.

  17. Used nuclear fuel separations process simulation and testing

    SciTech Connect

    Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D.

    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)

  18. Disposal of defense spent fuel and HLW from the Idaho Chemical Processing Plant

    SciTech Connect

    Ermold, L.F.; Loo, H.H.; Klingler, R.D.; Herzog, J.D.; Knecht, D.A.

    1992-12-01

    Acid high-level radioactive waste (HLW) resulting from fuel reprocessing at the Idaho Chemical Processing Plant (ICPP) for the US Department of Energy (DOE) has been solidified to a calcine since 1963 and stored in stainless steel bins enclosed by concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also in storage ate the ICPP. In April, 1992, DOE announced that spent fuel would no longer be reprocessed to recover enriched uranium and called for a shutdown of the reprocessing facilities at the ICPP. A new Spent Fuel and HLW Technology Development program was subsequently initiated to develop technologies for immobilizing ICPP spent fuels and HLW for disposal, in accordance with the Nuclear Waste Policy Act. The Program elements include Systems Analysis, Graphite Fuel Disposal, Other Spent Fuel Disposal, Sodium-Bearing Liquid Waste Processing, Calcine Immobilization, and Metal Recycle/Waste Minimization. This paper presents an overview of the ICPP radioactive wastes and current spent fuels, with an emphasis on the description of HLW and spent fuels requiring repository disposal.

  19. Advanced LMFBR fuel cladding susceptability to stress corrosion due to reprocessing impurities

    SciTech Connect

    Henslee, S.P.

    1987-03-01

    The potential degradation of LMFBR fuel cladding alloys by chlorides, when used in metallic fuel systems, was evaluated. The alloys tested were D-9 and HT-9 stainless steels, austenitic and ferritic alloys respectively. These two alloys were tested in parallel with and their performance compared to the austenitic stainless steel Type 316. All alloys were tested for 7400 hours in a stress rupture environment with chloride exposure at either 550/degree/C 650/degree/C. None of the alloys tested were found to exhibit any degradation in time-to-rupture by the presence of chlorides under the conditions imposed during testing. 8 refs., 4 figs., 2 tabs.

  20. Economic Study of Spent Nuclear Fuel Storage and Reprocessing Practices in Russia

    SciTech Connect

    C. E. Singer; G. H. Miley

    1997-10-01

    This report describes a study of nuclear power economics in Russia. It addresses political and institutional background factors which constrain Russia's energy choices in the short and intermediate run. In the approach developed here, political and institutional factors might dominate short-term decisions, but the comparative costs of Russia's fuel-cycle options are likely to constrain her long-term energy strategy. To this end, the authors have also formulated a set of policy questions which should be addressed using a quantitative decision modeling which analyzes economic costs for all major components of different fuel cycle options, including the evolution of uranium prices.

  1. FY 2007 LDRD Director's R&D Progress SummaryProposal Title: Developing a Science Base for Fuel Reprocessing Separations in the Global Nuclear Energy Program

    SciTech Connect

    de Almeida, Valmor F; Tsouris, Costas; Birdwell Jr, Joseph F; D'Azevedo, Ed F; Jubin, Robert Thomas; DePaoli, David W; Moyer, Bruce A

    2011-01-01

    This work is aimed at developing an experimentally validated computational capability for understanding the complex processes governing the performance of solvent extraction devices used for separations in nuclear fuel reprocessing. These applications pose a grand challenge due to the combination of complicating factors in a three-dimensional, turbulent, reactive, multicomponent, multiphase/interface fluid flow system. The currently limited process simulation and scale-up capabilities provides uncertainty in the ability to select and design the separations technology for the demonstration plan of the Global Nuclear Energy Partnership (GNEP) program. We anticipate the development of science-based models for technology development and design. This project will position ORNL to address the emerging opportunity by creating an expandable process model validated experimentally. This project has three major thrusts, namely, a prototype experimental station, a continuum modeling and simulation effort, and molecular modeling and kinetics support. Excellent progress has been made in corresponding activities in this first year in: (1) defining, assembling, and operating a relevant prototype system for model validation; (2) establishing a mathematical model for fluid flow and transport; (3) deploying sub-scale molecular modeling.

  2. Comparison studies of head-end reprocessing using three LWR fuels

    SciTech Connect

    Goode, J.H.; Stacy, R.G.; Vaughen, V.C.A.

    1980-06-01

    The removal of {sup 3}H by voloxidation and the dissolution behavior of two PWR and one BWR fuels were compared in hot-cell studies. The experiments showed that >99% of the {sup 3}H contained in the irradiated UO{sub 2} was volatilized by oxidation in air at 753{sup 0}K (480{sup 0}C). The oxidation did not affect the dissolution of the uranium and plutonium in 7 M HNO{sub 3} (0.02 to 0.03% insoluble plutonium) but did create a fission-product residue that was two to three times more insoluble. From 40 to 69% of the ternary fission-product {sup 3}H was found in the Zircaloy cladding of the fuel rods. Voloxidation had little effect on the {sup 3}H held in the Zircaloy cladding; oxidation for 6 h at 753{sup 0}K released only 0.05% of the {sup 3}H.

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

    PubMed Central

    Narimani, Mohamad; Gamari-give, Hossien; Abolgasemi, Abas; Molavi, Parviz

    2011-01-01

    Objective 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. Method 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. Results 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. Conclusion 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. PMID:22952539

  4. Corrosion study of a highly durable electrolyzer based on cold crucible technique for pyrochemical reprocessing of spent nuclear oxide fuel

    NASA Astrophysics Data System (ADS)

    Takeuchi, M.; Arai, Y.; Kase, T.; Nakajima, Y.

    2013-01-01

    The application of the cold crucible technique to a pyrochemical electrolyzer used in the oxide-electrowinning method, which is a method for the pyrochemical reprocessing of spent nuclear oxide fuel, is proposed as a means for improving corrosion resistance. The electrolyzer suffers from a severe corrosion environment consisting of molten salt and corrosive gas. In this study, corrosion tests for several metals in molten 2CsCl-NaCl at 923 K with purging chlorine gas were conducted under controlled material temperature conditions. The results revealed that the corrosion rates of several materials were significantly decreased by the material cooling effect. In particular, Hastelloy C-22 showed excellent corrosion resistance with a corrosion rate of just under 0.01 mm/y in both molten salt and vapor phases by controlling the material surface at 473 K. Finally, an engineering-scale crucible composed of Hastelloy C-22 was manufactured to demonstrate the basic function of the cold crucible. The cold crucible induction melting system with the new concept Hastelloy crucible showed good compatibility with respect to its heating and cooling performances.

  5. Metal-Organic Frameworks for Removal of Xe and Kr from Nuclear Fuel Reprocessing Plants

    SciTech Connect

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

    2012-08-07

    Removal of Xenon (Xe) and Krypton (Kr) from in parts per million (ppm) levels were demonstrated for the first time using two well known metal-organic frameworks (MOFs), HKUST-1 and Ni/DOBDC. Results of an activated carbon were also included for comparison. Ni/DOBDC has higher Xe/Kr selectivities than those of the activated carbon. Moreover, results show that the Ni/DOBDC and HKUST-1 can selectively adsorb Xe and Kr from air even at 1000 ppm concentration. This shows a promising future for MOFs in a radioactive nuclides separation from spent fuel.

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

    NASA Astrophysics Data System (ADS)

    Ochkin, A.; Gladilov, D.; Stefanovsky, S.

    2012-10-01

    Radiation hazard of radionuclide has been calculated as a product of Aɛ where A is an activity of radionuclide and ɛ is a dose coefficient through ingestion. The values Aɛ 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.

  7. On the possibility of reprocessing spent nuclear fuel and radioactive waste by plasma methods

    SciTech Connect

    Vorona, N. A.; Gavrikov, A. V. Samokhin, A. A.; Smirnov, V. P.; Khomyakov, Yu. S.

    2015-12-15

    The concept of plasma separation of spent nuclear fuel and radioactive waste is presented. An approach that is based on using an accelerating potential to overcome the energy and angular spread of plasma ions at the separation region inlet and utilizing a potential well to separate spatially the ions of different masses is proposed. It is demonstrated that such separation may be performed at distances of about 1 m with electrical potentials of about 1 kV and a magnetic field of about 1 kG. The estimates of energy consumption and performance of the plasma separation method are presented. These estimates illustrate its potential for technological application. The results of development and construction of an experimental setup for testing the method of plasma separation are presented.

  8. On the possibility of reprocessing spent nuclear fuel and radioactive waste by plasma methods

    NASA Astrophysics Data System (ADS)

    Vorona, N. A.; Gavrikov, A. V.; Samokhin, A. A.; Smirnov, V. P.; Khomyakov, Yu. S.

    2015-12-01

    The concept of plasma separation of spent nuclear fuel and radioactive waste is presented. An approach that is based on using an accelerating potential to overcome the energy and angular spread of plasma ions at the separation region inlet and utilizing a potential well to separate spatially the ions of different masses is proposed. It is demonstrated that such separation may be performed at distances of about 1 m with electrical potentials of about 1 kV and a magnetic field of about 1 kG. The estimates of energy consumption and performance of the plasma separation method are presented. These estimates illustrate its potential for technological application. The results of development and construction of an experimental setup for testing the method of plasma separation are presented.

  9. Head-end reprocessing studies of H.B. Robinson-2 fuel: II. Parametric voloxidation studies

    SciTech Connect

    Goode, J.H.; Stacy, R.G.; Vaughen, V.C.A.

    1980-05-01

    A series of hot-cell tests was conducted with UO{sub 2} that had been irradiated to an average of 28,000 MWd/t in the H.B. Robinson-2 reactor of the Carolina Power and Light Company. The tests examined the effects of temperature and of the rate of oxygen supply on the release of gaseous and semivolatile fission products, while the fuel fragments were tumbled at 12 rpm during voloxidation - the high-temperature oxidation of UO{sub 2} to U{sub 3}O{sub 8}. The experiments showed that >99.9% of the tritium in the irradiated UO{sub 2} was released to the off-gas stream at temperatures of 480 and 550{sup 0}C and at oxygen feed rates ranging from 0.1 to 1.2 mol/h. The release of {sup 85}Kr varied from 2 to 7% of the fuel inventory. The U{sub 3}O{sub 8} product ({similar_to}99% smaller than 44 {mu}m) was easily dissolved in 7 M HNO{sub 3}. One 2-h leach in 7 M HNO{sub 3} dissolved {similar_to}99.5% of the heavy metals; a second 2-h leach in 7 M HNO{sub 3} brought the total to >99.98%. Voloxidation did not affect the final solubility of the uranium and plutonium but did increase the weight of the insoluble fission product residue from 0.18% of the irradiated UO{sub 2} to {similar_to}0.62%.

  10. Dismantling of Highly Contaminated Process Installations of the German Reprocessing Facility (WAK) - Status of New Remote Handling Technology - 13287

    SciTech Connect

    Dux, Joachim; Friedrich, Daniel; Lutz, Werner; Ripholz, Martina

    2013-07-01

    Decommissioning and dismantling of the former German Pilot Reprocessing Plant Karlsruhe (WAK) including the Vitrification Facility (VEK) is being executed in different Project steps related to the reprocessing, HLLW storage and vitrification complexes /1/. While inside the reprocessing building the total inventory of process equipment has already been dismantled and disposed of, the HLLW storage and vitrification complex has been placed out of operation since vitrification and tank rinsing procedures where finalized in year 2010. This paper describes the progress made in dismantling of the shielded boxes of the highly contaminated laboratory as a precondition to get access to the hot cells of the HLLW storage. The major challenges of the dismantling of this laboratory were the high dose rates up to 700 mSv/h and the locking technology for the removal of the hot cell installations. In parallel extensive prototype testing of different carrier systems and power manipulators to be applied to dismantle the HLLW-tanks and other hot cell equipment is ongoing. First experiences with the new manipulator carrier system and a new master slave manipulator with force reflection will be reported. (authors)

  11. Noble gas atmospheric monitoring at reprocessing facilities

    SciTech Connect

    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.

  12. 75 FR 45167 - Notice of Public Workshop on a Potential Rulemaking for Spent Nuclear Fuel Reprocessing Facilities

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-02

    ... Register on August 31, 1984 (49 FR 34658) discusses waste from reprocessing facilities in the first and... Transuranic Special Nuclear Material (SNM) Classification Certain fissile elements such as americium...

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

    PubMed

    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. PMID:21703737

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

    SciTech Connect

    Enokida, Youichi; Kitano, Motoki; Sawada, Kayo

    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)

  15. Shaping process makes fuels

    SciTech Connect

    Tabak, S.A.; Krambeck, F.J.

    1985-09-01

    The Mobil Olefin to Gasoline and Distillate (MOGD) process is described in which light olefinic compunds can be converted to high quality gasoline and distillate. This process, now ready for commercialization is based on a unique synthetic zeolite catalyst, the shape of which selectively oligomerizes light olefins to higher molecular weight iso-olefins. The highly flexible process can be designed to produce distillate/gasoline ratios of 0/100 to 90/10 for a commercial plant, depending on market requirements. MOGD is applicable to a wide range of feed streams ranging from ethylene to 400 degrees F end point olefinic naphtha. The process has been tested using commercially produced catalyst in refinery-scale equipment.

  16. Re-processing and interpretation of 2D seismic data from the Kristineberg mining area, northern Sweden

    NASA Astrophysics Data System (ADS)

    Ehsan, Siddique Akhtar; Malehmir, Alireza; Dehghannejad, Mahdieh

    2012-05-01

    The Kristineberg mining area in the western part of the Skellefte ore district, northern Sweden, contains the largest massive sulphide deposit in the district. In 2003, two parallel seismic lines, Profiles 1 and 5, each about 25 km long and about 8 km apart were acquired in the Kristineberg area. The initial processing results were successful in imaging the large-scale structures of the area down to 12 km of the crust, but resulted in relatively poor seismic image near the mine. In this paper, we re-processed the seismic data along Profile 1 that crosses the mine. The main objective was to improve the seismic section near the mine for further correlation with new seismic data recently acquired in the area. The crooked-line acquisition geometry, very low fold coverage of less than 17, complex geology and sparse outcrops in the area made the data re-processing and interpretation challenging. Despite these challenges, significant improvement is observed in the seismic data, in terms of event continuity and resolution. Refraction static corrections allowed high frequencies to be retained, which improved the seismic section. The refraction static solution was manually checked and adjusted at every iteration to avoid unstable solutions. 3D visualization of the re-processed data with other seismic profiles recently acquired in the area allowed the seismic reflections to be correlated. The majority of the reflections are interpreted to originate from either fault zones or lithological contacts. A very shallow reflection correlates well with the location of the Kristineberg mineralized horizon.

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

    SciTech Connect

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

    SciTech Connect

    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

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

    NASA Astrophysics Data System (ADS)

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

    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.

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

    PubMed

    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. PMID:26931895

  1. Principles of Product Quality Control of German Radioactive Waste Forms from the Reprocessing of Spent Fuel: Vitrification, Compaction and Numerical Simulation - 12529

    SciTech Connect

    Tietze-Jaensch, Holger; Schneider, Stephan; Aksyutina, Yuliya; Bosbach, Dirk; Gauthier, Rene; Eissler, Alexander

    2012-07-01

    The German product quality control is inter alia responsible for control of two radioactive waste forms of heat generating waste: a) homogeneous vitrified HLW and b) heterogeneous compacted hulls, end-pieces and technological metallic waste. In either case, significantly different metrology is employed at the site of the conditioning plant for the obligatory nuclide inventory declaration. To facilitate an independent evaluation and checking of the accompanying documentation numerical simulations are carried out. The physical and chemical properties of radioactive waste residues are used to assess the data consistency and uncertainty margins, as well as to predict the long-term behavior of the radioactive waste. This is relevant for repository acceptance and safety considerations. Our new numerical approach follows a bottom-up simulation starting from the burn-up behavior of the fuel elements in the reactor core. The output of these burn-up calculations is then coupled with a program that simulates the material separation in the subsequent dissolution and extraction processes normalized to the mass balance. Follow-up simulations of the separated reprocessing lines of a) the vitrification of highly-active liquid and b) the compaction of residual intermediate-active metallic hulls remaining after fuel pellets dissolution, end-pieces and technological waste, allows calculating expectation values for the various repository relevant properties of either waste stream. The principles of the German product quality control of radioactive waste residues from the spent fuel reprocessing have been introduced and explained. Namely, heat generating homogeneous vitrified HLW and heterogeneous compacted metallic MLW have been discussed. The advantages of a complementary numerical property simulation have been made clear and examples of benefits are presented. We have compiled a new program suite to calculate the physical and radio-chemical properties of common nuclear waste

  2. Cost/performance comparison between pulse columns and centrifugal contactors designed to process Clinch River Breeder Reactor fuel

    SciTech Connect

    Ciucci, J.A. Jr.

    1983-12-01

    A comparison between pulse columns and centrifugal contactors was made to determine which type of equipment was more advantageous for use in the primary decontamination cycle of a remotely operated fuel reprocessing plant. Clinch River Breeder Reactor (CRBR) fuel was chosen as the fuel to be processed in the proposed 1 metric tonne/day reprocessing facility. The pulse columns and centrifugal contactors were compared on a performance and total cost basis. From this comparison, either the pulse columns or the centrifugal contactors will be recommended for use in a fuel reprocessing plant built to reprocess CRBR fuel. The reliability, solvent exposure to radiation, required time to reach steady state, and the total costs were the primary areas of concern for the comparison. The pulse column units were determined to be more reliable than the centrifugal contactors. When a centrifugal contactor motor fails, it can be remotely changed in less than one eight hour shift. Pulse columns expose the solvent to approximately five times as much radiation dose as the centrifugal contactor units; however, the proposed solvent recovery system adequately cleans the solvent for either case. The time required for pulse columns to reach steady state is many times longer than the time required for centrifugal contactors to reach steady state. The cost comparison between the two types of contacting equipment resulted in centrifugal contactors costing 85% of the total cost of pulse columns when the contactors were stacked on three levels in the module. If the centrifugal contactors were all positioned on the top level of a module with the unoccupied volume in the module occupied by other equipment, the centrifugal contactors cost is 66% of the total cost of pulse columns. Based on these results, centrifugal contactors are recommended for use in a remotely operated reprocessing plant built to reprocess CRBR fuel.

  3. Mesoscale to plant-scale models of nuclear waste reprocessing.

    SciTech Connect

    Noble, David Frederick; O'Hern, Timothy John; Moffat, Harry K.; Nemer, Martin B.; Domino, Stefan Paul; Rao, Rekha Ranjana; Cipiti, Benjamin B.; Brotherton, Christopher M.; Jove-Colon, Carlos F.; Pawlowski, Roger Patrick

    2010-09-01

    Imported oil exacerabates our trade deficit and funds anti-American regimes. Nuclear Energy (NE) is a demonstrated technology with high efficiency. NE's two biggest political detriments are possible accidents and nuclear waste disposal. For NE policy, proliferation is the biggest obstacle. Nuclear waste can be reduced through reprocessing, where fuel rods are separated into various streams, some of which can be reused in reactors. Current process developed in the 1950s is dirty and expensive, U/Pu separation is the most critical. Fuel rods are sheared and dissolved in acid to extract fissile material in a centrifugal contactor. Plants have many contacts in series with other separations. We have taken a science and simulation-based approach to develop a modern reprocessing plant. Models of reprocessing plants are needed to support nuclear materials accountancy, nonproliferation, plant design, and plant scale-up.

  4. Processing sunflower oil for fuel

    SciTech Connect

    Backer, L.F.; Jacobsen, L.; Olson, C.

    1982-05-01

    Research on processing of sunflower seed for oil was initiated to evaluate the equipment that might adapt best to on-farm or small factory production facilities. The first devices identified for evaluation were auger press expeller units, primary oil cleaning equipment, and final filters. A series of standard finishing filtration tests were carried out on sunflower oil and sunflower oil - diesel fuel blends using sunflower oil from four different sources.

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

    SciTech Connect

    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.

  6. Synthetic fuels handbook: properties, process and performance

    SciTech Connect

    Speight, J.

    2008-07-01

    The handbook is a comprehensive guide to the benefits and trade-offs of numerous alternative fuels, presenting expert analyses of the different properties, processes, and performance characteristics of each fuel. It discusses the concept systems and technology involved in the production of fuels on both industrial and individual scales. Chapters 5 and 7 are of special interest to the coal industry. Contents: Chapter 1. Fuel Sources - Conventional and Non-conventional; Chapter 2. Natural Gas; Chapter 3. Fuels From Petroleum and Heavy Oil; Chapter 4. Fuels From Tar Sand Bitumen; Chapter 5. Fuels From Coal; Chapter 6. Fuels From Oil Shale; Chapter 7. Fuels From Synthesis Gas; Chapter 8. Fuels From Biomass; Chapter 9. Fuels From Crops; Chapter 10. Fuels From Wood; Chapter 11. Fuels From Domestic and Industrial Waste; Chapter 12. Landfill Gas. 3 apps.

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

    PubMed

    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. PMID:24138408

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

    SciTech Connect

    Charlton, William S

    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.

  9. Microbial fuel cell treatment of fuel process wastewater

    DOEpatents

    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.

  10. Dry process dependency of dupic fuel cycle

    SciTech Connect

    Park, Kwangheon; Whang, Juho; Kim, Yun-goo; Kim, Heemoon

    1996-12-31

    During the Dry Process, volatile and semi-volatile elements are released from the fuel. The effects of these released radioactive nuclides on DUPIC fuel cycle are analyzed from the view-point of radiation hazard, decay beat, and hazard index. Radiation hazard of fresh and spent DUPIC fuel is sensitive to the method of Dry Process. Decay beat of the fuel is also affected. Hazard index turned out not to be dependent on Dry Process.

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

    SciTech Connect

    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 the critical reactors

  12. Corrosion resistance of ceramic materials in pyrochemical reprocessing condition by using molten salt for spent nuclear oxide fuel

    NASA Astrophysics Data System (ADS)

    Takeuchi, M.; Kato, T.; Hanada, K.; Koizumi, T.; Aose, S.

    2005-02-01

    The corrosion resistance of ceramic materials in pyrochemical reprocessing using molten salts was discussed through the thermodynamic calculation and corrosion test. The corrosion test was basically carried out in alkali molten salt under chlorine gas. In addition, the effects of oxygen, carbon and main fission product's chlorides on ceramics corrosion were evaluated in that condition. Most of ceramic oxides showed good chemical stability on chlorine, oxygen and uranyl chloride from thermodynamic calculation results. On the other hand, from corrosion test result, silicon nitride, mullite (Al6Si2O13) and cordierite (Mg2Al3(AlSi5O18)) have a good corrosion resistance which is corresponding to 0.1 mm/y or less. No cracks on the materials were observed and flexural strength did not drop remarkably after 480 h corrosion testing in molten salt under Cl2 O2 atmosphere.

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

    PubMed Central

    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. PMID:26594183

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

    PubMed

    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. PMID:26594183

  15. Evolution of isotopic composition of reprocessed uranium during the multiple recycling in light water reactors with natural uranium feed

    SciTech Connect

    Smirnov, A. Yu. Sulaberidze, G. A.; Alekseev, P. N.; Dudnikov, A. A.; Nevinitsa, V. A. Proselkov, V. N.; Chibinyaev, A. V.

    2012-12-15

    A complex approach based on the consistent modeling of neutron-physics processes and processes of cascade separation of isotopes is applied for analyzing physical problems of the multiple usage of reprocessed uranium in the fuel cycle of light water reactors. A number of scenarios of multiple recycling of reprocessed uranium in light water reactors are considered. In the process, an excess absorption of neutrons by the {sup 236}U isotope is compensated by re-enrichment in the {sup 235}U isotope. Specific consumptions of natural uranium for re-enrichment of the reprocessed uranium depending on the content of the {sup 232}U isotope are obtained.

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

    SciTech Connect

    1993-09-01

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

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

    SciTech Connect

    Not Available

    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.

  18. Fuel cycles for the 80's

    SciTech Connect

    Not Available

    1980-01-01

    Papers presented at the American Nuclear Society's topical meeting on the fuel cycle are summarized. Present progress and goals in the areas of fuel fabrication, fuel reprocessing, spent fuel storage, accountability, and safeguards are reported. Present governmental policies which affect the fuel cycle are also discussed. Individual presentations are processed for inclusion in the Energy Data Base.(DMC)

  19. CONSTRUCTION PROGRESS PHOTO SHOWING WEST STORAGE BASIN AT FUEL STORAGE ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    CONSTRUCTION PROGRESS PHOTO SHOWING WEST STORAGE BASIN AT FUEL STORAGE BUILDING (CPP-603). INL PHOTO NUMBER NRTS-51-689. Unknown Photographer, 1950 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  20. Spent fuel storage and management in the United Kingdom

    SciTech Connect

    Sills, R.J.

    1989-04-01

    During the past 33 years, fuel of various types have been stored, transported and reprocessed in the United Kingdom. This paper provides an overview of those programs starting from the Magnox stations, through the AGR program and the move to LWR fuel. Throughout this time BNFL has provided services for fuel storage, reprocessing, transportation and the enrichment and fabrication of new fuel. The development of new plants and processes to handle the changing fuel types and the associated waste management schemes will be addressed. A description of future plans for fuel storage and reprocessing is included.

  1. Simulation of ground-water flow near the nuclear-fuel reprocessing facility at the Western New York Nuclear Service Center, Cattaraugus County, New York

    USGS Publications Warehouse

    Yager, R.M.

    1987-01-01

    A two-dimensional finite-difference model was developed to simulate groundwater flow in a surficial sand and gravel deposit underlying the nuclear fuel reprocessing facility at Western New York Nuclear Service Center near West Valley, N.Y. The sand and gravel deposit overlies a till plateau that abuts an upland area of siltstone and shale on its west side, and is bounded on the other three sides by deeply incised stream channels that drain to Buttermilk Creek, a tributary to Cattaraugus Creek. Radioactive materials are stored within the reprocessing plant and are also buried within a till deposit at the facility. Tritiated water is stored in a lagoon system near the plant and released under permit to Franks Creek, a tributary to Buttermilk Creek. Groundwater levels predicted by steady-state simulations closely matched those measured in 23 observation wells, with an average error of 0.5 meter. Simulated groundwater discharges to two stream channels and a subsurface drain were within 5% of recorded values. Steady-state simulations used an average annual recharge rate of 46 cm/yr; predicted evapotranspiration loss from the ground was 20 cm/yr. The lateral range in hydraulic conductivity obtained through model calibration was 0.6 to 10 m/day. Model simulations indicated that 33% of the groundwater discharged from the sand and gravel unit (2.6 L/sec) is lost by evapotranspiration, 3% (3.0 L/sec) flows to seepage faces at the periphery of the plateau, 20% (1.6 L/sec) discharges to stream channels that drain a large wetland area near the center of the plateau, and the remaining 8% (0.6 L/sec) discharges to a subsurface french drain and to a wastewater treatment system. Groundwater levels computed by a transient-state simulation of an annual climatic cycle, including seasonal variation in recharge and evapotranspiration, closely matched water levels measured in eight observation wells. The model predicted that the subsurface drain and the stream channel that drains the

  2. Clinical Practice Guidelines for Endoscope Reprocessing

    PubMed Central

    Oh, Hyun Jin

    2015-01-01

    Gastrointestinal endoscopy is effective and safe for the screening, diagnosis, and treatment of gastrointestinal disease. However, issues regarding endoscope-transmitted infections are emerging. Many countries have established and continuously revise guidelines for endoscope reprocessing in order to prevent infections. While there are common processes used in endoscope reprocessing, differences exist among these guidelines. It is important that the reprocessing of gastrointestinal endoscopes be carried out in accordance with the recommendations for each step of the process. PMID:26473117

  3. Plasma coal reprocessing

    NASA Astrophysics Data System (ADS)

    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.

  4. Fuel quality processing study, volume 1

    NASA Astrophysics Data System (ADS)

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

    1981-04-01

    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.

  5. Fuel quality processing study, volume 1

    NASA Technical Reports Server (NTRS)

    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.

  6. Nuclear fuel, refueling, fuel handling, and licensing and regulation. Volume eleven

    SciTech Connect

    Not Available

    1986-01-01

    Volume eleven covers nuclear fuel (what is nuclear fuel, the nuclear fuel cycle, uranium mining, milling, and refining, uranium enrichment, nuclear fuel fabrication, fuel reprocessing), refueling and fuel handling (fuel assembly identification, fuel handling equipment, the fueling and refueling process, PWR refueling, BWR refueling), and licensing and regulation requirements (development of nuclear energy, federal licensing and regulatory organization, schedule for nuclear power plants, contents of reports to the Federal regulatory agency, nuclear power plant operator qualification).

  7. Dry Processing of Used Nuclear Fuel

    SciTech Connect

    K. M. Goff; M. F. Simpson

    2009-09-01

    Dry (non-aqueous) separations technologies have been used for treatment of used nuclear fuel since the 1960s, and they are still being developed and demonstrated in many countries. Dry technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. Within the Department of Energy’s Advanced Fuel Cycle Initiative, an electrochemical process employing molten salts is being developed for recycle of fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. Much of the development of this technology is based on treatment of used Experimental Breeder Reactor II (EBR-II) fuel, which is metallic. Electrochemical treatment of the EBR-II fuel has been ongoing in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory since 1996. More than 3.8 metric tons of heavy metal of metallic fast reactor fuel have been treated using this technology. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including high-level waste work. A historic perspective on the background of dry processing will also be provided.

  8. 21 CFR 111.90 - What requirements apply to treatments, in-process adjustments, and reprocessing when there is a...

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... established in accordance with § 111.70 is not met? 111.90 Section 111.90 Food and Drugs FOOD AND DRUG... when a specification established in accordance with § 111.70 is not met? (a) You must not reprocess...

  9. 21 CFR 111.90 - What requirements apply to treatments, in-process adjustments, and reprocessing when there is a...

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... established in accordance with § 111.70 is not met? 111.90 Section 111.90 Food and Drugs FOOD AND DRUG... when a specification established in accordance with § 111.70 is not met? (a) You must not reprocess...

  10. 21 CFR 111.90 - What requirements apply to treatments, in-process adjustments, and reprocessing when there is a...

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... established in accordance with § 111.70 is not met? 111.90 Section 111.90 Food and Drugs FOOD AND DRUG... when a specification established in accordance with § 111.70 is not met? (a) You must not reprocess...

  11. 21 CFR 111.90 - What requirements apply to treatments, in-process adjustments, and reprocessing when there is a...

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... established in accordance with § 111.70 is not met? 111.90 Section 111.90 Food and Drugs FOOD AND DRUG... when a specification established in accordance with § 111.70 is not met? (a) You must not reprocess...

  12. 21 CFR 111.90 - What requirements apply to treatments, in-process adjustments, and reprocessing when there is a...

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... established in accordance with § 111.70 is not met? 111.90 Section 111.90 Food and Drugs FOOD AND DRUG... when a specification established in accordance with § 111.70 is not met? (a) You must not reprocess...

  13. Advanced Safeguards Approaches for New Reprocessing Facilities

    SciTech Connect

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Richard; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-06-24

    U.S. efforts to promote the international expansion of nuclear energy through the Global Nuclear Energy Partnership (GNEP) will result in a dramatic expansion of nuclear fuel cycle facilities in the United States. New demonstration facilities, such as the Advanced Fuel Cycle Facility (AFCF), the Advanced Burner Reactor (ABR), and the Consolidated Fuel Treatment Center (CFTC) will use advanced nuclear and chemical process technologies that must incorporate increased proliferation resistance to enhance nuclear safeguards. The ASA-100 Project, “Advanced Safeguards Approaches for New Nuclear Fuel Cycle Facilities,” commissioned by the NA-243 Office of NNSA, has been tasked with reviewing and developing advanced safeguards approaches for these demonstration facilities. Because one goal of GNEP is developing and sharing proliferation-resistant nuclear technology and services with partner nations, the safeguards approaches considered are consistent with international safeguards as currently implemented by the International Atomic Energy Agency (IAEA). This first report reviews possible safeguards approaches for the new fuel reprocessing processes to be deployed at the AFCF and CFTC facilities. Similar analyses addressing the ABR and transuranic (TRU) fuel fabrication lines at AFCF and CFTC will be presented in subsequent reports.

  14. Powder handling for automated fuel processing

    SciTech Connect

    Frederickson, J.R.; Eschenbaum, R.C.; Goldmann, L.H.

    1989-04-09

    Installation of the Secure Automated Fabrication (SAF) line has been completed. It is located in the Fuel Cycle Plant (FCP) at the Department of Energy's (DOE) Hanford site near Richland, Washington. The SAF line was designed to fabricate advanced reactor fuel pellets and assemble fuel pins by automated, remote operation. This paper describes powder handling equipment and techniques utilized for automated powder processing and powder conditioning systems in this line. 9 figs.

  15. Mathematical modeling of biomass fuels formation process

    SciTech Connect

    Gaska, Krzysztof Wandrasz, Andrzej J.

    2008-07-01

    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.

  16. Fuel quality/processing study. Volume 3: Fuel upgrading studies

    NASA Technical Reports Server (NTRS)

    Jones, G. E., Jr.; Bruggink, P.; Sinnett, C.

    1981-01-01

    The methods used to calculate the refinery selling prices for the turbine fuels of low quality are described. Detailed descriptions and economics of the upgrading schemes are included. These descriptions include flow diagrams showing the interconnection between processes and the stream flows involved. Each scheme is in a complete, integrated, stand alone facility. Except for the purchase of electricity and water, each scheme provides its own fuel and manufactures, when appropriate, its own hydrogen.

  17. The effects of oxygen, carbon dioxide and water vapor on reprocessing silicon carbide inert matrix fuels by corrosion in molten potassium carbonate

    NASA Astrophysics Data System (ADS)

    Cheng, Ting; Baney, Ronald H.; Tulenko, James

    2011-04-01

    The molten salt reaction/dissolution method for reprocessing silicon carbide based inert matrix fuels (IMF) is further developed in this paper through comparison of the corrosion rate in multiple gases and gas mixtures. Water vapor was firstly introduced in the SiC/K 2CO 3 corrosion system. The SiC corrosion rate in the H 2O atmosphere was dramatically enhanced 3-4-fold compared to the rate under an O 2 atmosphere. The corrosion rates in different atmospheres of O 2, CO 2, O 2/CO 2, H 2O, O 2/H 2O and CO 2/H 2O with various partial pressures were compared in order to determine the optimal reaction atmosphere and to better understand the reaction mechanism. The SiC pellets with 5 wt.% of CeO 2, a surrogate for PuO 2 were fabricated. CeO 2 was successfully separated from the SiC matrix by using the molten salt reaction/dissolution strategy.

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

    SciTech Connect

    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.

  19. SOUTH ELEVATION OF IRRADIATED FUEL STORAGE FACILITY LOCATED IN FUEL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    SOUTH ELEVATION OF IRRADIATED FUEL STORAGE FACILITY LOCATED IN FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTH. INL PHOTO NUMBER HD-54-15-2. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  20. NORTH ELEVATION OF IRRADIATED FUEL STORAGE FACILITY LOCATED IN FUEL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    NORTH ELEVATION OF IRRADIATED FUEL STORAGE FACILITY LOCATED IN FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTH. INL PHOTO NUMBER HD-54-16-1. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  1. Process for vaporizing a liquid hydrocarbon fuel

    DOEpatents

    Szydlowski, Donald F.; Kuzminskas, Vaidotas; Bittner, Joseph E.

    1981-01-01

    The object of the invention is to provide a process for vaporizing liquid hydrocarbon fuels efficiently and without the formation of carbon residue on the apparatus used. The process includes simultaneously passing the liquid fuel and an inert hot gas downwardly through a plurality of vertically spaed apart regions of high surface area packing material. The liquid thinly coats the packing surface, and the sensible heat of the hot gas vaporizes this coating of liquid. Unvaporized liquid passing through one region of packing is uniformly redistributed over the top surface of the next region until all fuel has been vaporized using only the sensible heat of the hot gas stream.

  2. Decommissioning the Fuel Process Building, a Shift in Paradigm for Terminating Safeguards on Process Holdup

    SciTech Connect

    Ivan R. Thomas

    2010-07-01

    INMM Abstract 51st Annual Meeting Decommissioning the Fuel Process Building, a Shift in Paradigm for Terminating Safeguards on Process Holdup The Fuel Process Building at the Idaho Nuclear Technology and Engineering Center (INTEC) is being decommissioned after nearly four decades of recovering high enriched uranium from various government owned spent nuclear fuels. The separations process began with fuel dissolution in one of multiple head-ends, followed by three cycles of uranium solvent extraction, and ending with denitration of uranyl nitrate product. The entire process was very complex, and the associated equipment formed an extensive maze of vessels, pumps, piping, and instrumentation within several layers of operating corridors and process cells. Despite formal flushing and cleanout procedures, an accurate accounting for the residual uranium held up in process equipment over extended years of operation, presented a daunting safeguards challenge. Upon cessation of domestic reprocessing, the holdup remained inaccessible and was exempt from measurement during ensuing physical inventories. In decommissioning the Fuel Process Building, the Idaho Cleanup Project, which operates the INTEC, deviated from the established requirements that all nuclear material holdup be measured and credited to the accountability books and that all nuclear materials, except attractiveness level E residual holdup, be transferred to another facility. Instead, the decommissioning involved grouting the process equipment in place, rather than measuring and removing the contained holdup for subsequent transfer. The grouting made the potentially attractiveness level C and D holdup even more inaccessible, thereby effectually converting the holdup to attractiveness level E and allowing for termination of safeguards controls. Prior to grouting the facility, the residual holdup was estimated by limited sampling and destructive analysis of solutions in process lines and by acceptable knowledge

  3. CONSTRUCTION PROGRESS PHOTO SHOWING FUEL STORAGE BUILDING (CPP603) LOOKING NORTHWEST. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    CONSTRUCTION PROGRESS PHOTO SHOWING FUEL STORAGE BUILDING (CPP-603) LOOKING NORTHWEST. INL PHOTO NUMBER NRTS-50-895. Unknown Photographer, 10/30/1950 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  4. CONSTRUCTION VIEW FUEL STORAGE BUILDING (CPP603) LOOKING EAST SHOWING ASBESTOS ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    CONSTRUCTION VIEW FUEL STORAGE BUILDING (CPP-603) LOOKING EAST SHOWING ASBESTOS SIDING. INL PHOTO NUMBER NRTS-51-1543. Unknown Photographer, 2/28/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  5. Impacts of (14)C discharges from a nuclear fuel reprocessing plant on surrounding vegetation: Comparison between grass field measurements and TOCATTA-χ and SSPAM(14)C model computations.

    PubMed

    Limer, Laura M C; Le Dizès-Maurel, Séverine; Klos, Ryk; Maro, Denis; Nordén, Maria

    2015-09-01

    This article compares and discusses the ability of two different models to reproduce the observed temporal variability in grass (14)C activity in the vicinity of AREVA-NC La Hague nuclear fuel reprocessing plant in France. These two models are the TOCATTA-χ model, which is specifically designed for modelling transfer of (14)C (and tritium) in the terrestrial environment over short to medium timescales (days to years), and SSPAM(14)C, which has been developed to model the transfer of (14)C in the soil-plant-atmosphere with consideration over both short and long timescales (days to thousands of years). The main goal of this article is to discuss the strengths and weaknesses of the models studied, and to investigate if modelling could be improved through consideration of a much higher level of detail of plant physiology and/or higher number of plant compartments. These models 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. The two models have different objectives in terms of the timescales they are intended to be applied over, and thus incorporate biological processes, such as photosynthesis and plant growth, at different levels of complexity. It was found that the inclusion of seasonal dynamics in the models improved predictions of the specific activity in grass for such a source term of atmospheric (14)C. PMID:26063400

  6. Geohydrologic conditions at the nuclear-fuels reprocessing plant and waste-management facilities at the Western New York Nuclear Service Center, Cattaraugus County, New York

    USGS Publications Warehouse

    Bergeron, M.P.; Kappel, W.M.; Yager, R.M.

    1987-01-01

    A nuclear-fuel reprocessing plant, a high-level radioactive liquid-waste tank complex, and related waste facilities occupy 100 hectares (ha) within the Western New York Nuclear Service Center near West Valley, N.Y. The facilities are underlain by glacial and postglacial deposits that fill an ancestrial bedrock valley. The main plant facilities are on an elevated plateau referred to as the north plateau. Groundwater on the north plateau moves laterally within a surficial sand and gravel from the main plant building to areas northeast, east, and southeast of the facilities. The sand and gravel ranges from 1 to 10 m thick and has a hydraulic conductivity ranging from 0.1 to 7.9 m/day. Two separate burial grounds, a 4-ha area for low-level radioactive waste disposal and a 2.9-ha area for disposal of higher-level waste are excavated into a clay-rich till that ranges from 22 to 28 m thick. Migration of an organic solvent from the area of higher level waste at shallow depth in the till suggests that a shallow, fractured, oxidized, and weathered till is a significant pathway for lateral movement of groundwater. Below this zone, groundwater moves vertically downward through the till to recharge a lacustrine silt and fine sand. Within the saturated parts of the lacustrine unit, groundwater moves laterally to the northeast toward Buttermilk Creek. Hydraulic conductivity of the till, based on field and laboratory analyses , ranges from 0.000018 to 0.000086 m/day. (USGS)

  7. Geohydrologic conditions at the Nuclear Fuel Reprocessing Plant and Waste-Management Facilities at the western New York Nuclear Service Center, Cattaraugus County, New York

    SciTech Connect

    Bergeron, M.P.; Kappel, W.M.; Yager, R.M.

    1987-01-01

    A nuclear-fuel reprocessing plant, a high-level radioactive liquid-waste tank complex, and related waste facilities occupy 100 hectares (ha) within the Western New York Nuclear Service Center near West Valley, NY. The facilities are underlain by glacial and postglacial deposits that fill an ancestral bedrock valley. The main plant facilities are on an elevated plateau referred to as the north plateau. Groundwater on the north plateau moves laterally within a surficial sand and gravel from the main plant building to areas northeast, east, and southeast of the facilities. The sand and gravel ranges from 1 to 10 m thick and has a hydraulic conductivity ranging from 0.1 to 7.9 m/day. Two separate burial grounds, a 4-ha area for low-level radioactive waste disposal and a 2.9-ha area for disposal of higher-level waste are excavated into a clay-rich till that ranges from 22 to 28 m thick. Migration of an organic solvent from the area of higher level waste at shallow depth in the till suggests that a shallow, fractured, oxidized, and weathered till is a significant pathway for lateral movement of groundwater. Below this zone, groundwater moves vertically downward through the till to recharge a lacustrine silt and fine sand. Within the saturated parts of the lacustrine unit, groundwater moves laterally to the northeast toward Buttermilk Creek. Hydraulic conductivity of the till, based on field and laboratory analyses, ranges from 0.000018 to 0.000086 m/day.

  8. Process for preparing a liquid fuel composition

    DOEpatents

    Singerman, Gary M.

    1982-03-16

    A process for preparing a liquid fuel composition which comprises liquefying coal, separating a mixture of phenols from said liquefied coal, converting said phenols to the corresponding mixture of anisoles, subjecting at least a portion of the remainder of said liquefied coal to hydrotreatment, subjecting at least a portion of said hydrotreated liquefied coal to reforming to obtain reformate and then combining at least a portion of said anisoles and at least a portion of said reformate to obtain said liquid fuel composition.

  9. Data validation and security for reprocessing.

    SciTech Connect

    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.

  10. Biomass conversion processes for energy and fuels

    NASA Astrophysics Data System (ADS)

    Sofer, S. S.; Zaborsky, O. R.

    The book treats biomass sources, promising processes for the conversion of biomass into energy and fuels, and the technical and economic considerations in biomass conversion. Sources of biomass examined include crop residues and municipal, animal and industrial wastes, agricultural and forestry residues, aquatic biomass, marine biomass and silvicultural energy farms. Processes for biomass energy and fuel conversion by direct combustion (the Andco-Torrax system), thermochemical conversion (flash pyrolysis, carboxylolysis, pyrolysis, Purox process, gasification and syngas recycling) and biochemical conversion (anaerobic digestion, methanogenesis and ethanol fermentation) are discussed, and mass and energy balances are presented for each system.

  11. SOFC system with integrated catalytic fuel processing

    NASA Astrophysics Data System (ADS)

    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.

  12. Safeguarding spent fuel storage and final disposal activities

    SciTech Connect

    Weh, R.; Wogatzki, E. )

    1991-01-01

    In Germany, the Atomic Energy Act provides for the spent fuel generated by nuclear power reactors to be reprocessed, if this is technically safe and economically viable. Thus the major share of used fuel from the German reactors is brought to reprocessing. The fuel recovered in this process is intended to be recycled into suitable reactors. The actual reprocessing is carried out abroad, in preference to a domestic solution, and the residues returned to Germany. This paper describes safeguarding measures for spent fuel storage and final disposal activities that are employed in Germany.

  13. Characterization of Used Nuclear Fuel with Multivariate Analysis for Process Monitoring

    SciTech Connect

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

    2014-01-01

    The Multi-Isotope Process (MIP) Monitor combines gamma spectroscopy and multivariate analysis to detect anomalies in various process streams in a nuclear fuel reprocessing system. Measured spectra are compared to models of nominal behavior at each measurement location to detect unexpected changes in system behavior. In order to improve the accuracy and specificity of process monitoring, fuel characterization may be used to more accurately train subsequent models in a full analysis scheme. This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict 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

  14. Corrosion-Resistant Ti- xNb- xZr Alloys for Nitric Acid Applications in Spent Nuclear Fuel Reprocessing Plants

    NASA Astrophysics Data System (ADS)

    Manivasagam, Geetha; Anbarasan, V.; Kamachi Mudali, U.; Raj, Baldev

    2011-09-01

    This article reports the development, microstructure, and corrosion behavior of two new alloys such as Ti-4Nb-4Zr and Ti-2Nb-2Zr in boiling nitric acid environment. The corrosion test was carried out in the liquid, vapor, and condensate phases of 11.5 M nitric acid, and the potentiodynamic anodic polarization studies were performed at room temperature for both alloys. The samples subjected to three-phase corrosion testing were characterized using scanning electron microscopy (SEM) and energy-dispersive X-ray microanalysis (EDAX). As Ti-2Nb-2Zr alloy exhibited inferior corrosion behavior in comparison to Ti-4Nb-4Zr in all three phases, weldability and heat treatment studies were carried out only on Ti-4Nb-4Zr alloy. The weldability of the new alloy was evaluated using tungsten inert gas (TIG) welding processes, and the welded specimen was thereafter tested for its corrosion behavior in all three phases. The results of the present investigation revealed that the newly developed near alpha Ti-4Nb-4Zr alloy possessed superior corrosion resistance in all three phases and excellent weldability compared to conventional alloys used for nitric acid application in spent nuclear reprocessing plants. Further, the corrosion resistance of the beta heat-treated Ti-4Nb-4Zr alloy was superior when compared to the sample heat treated in the alpha + beta phase.

  15. Process automation

    SciTech Connect

    Moser, D.R.

    1986-01-01

    Process automation technology has been pursued in the chemical processing industries and to a very limited extent in nuclear fuel reprocessing. Its effective use has been restricted in the past by the lack of diverse and reliable process instrumentation and the unavailability of sophisticated software designed for process control. The Integrated Equipment Test (IET) facility was developed by the Consolidated Fuel Reprocessing Program (CFRP) in part to demonstrate new concepts for control of advanced nuclear fuel reprocessing plants. A demonstration of fuel reprocessing equipment automation using advanced instrumentation and a modern, microprocessor-based control system is nearing completion in the facility. This facility provides for the synergistic testing of all chemical process features of a prototypical fuel reprocessing plant that can be attained with unirradiated uranium-bearing feed materials. The unique equipment and mission of the IET facility make it an ideal test bed for automation studies. This effort will provide for the demonstration of the plant automation concept and for the development of techniques for similar applications in a full-scale plant. A set of preliminary recommendations for implementing process automation has been compiled. Some of these concepts are not generally recognized or accepted. The automation work now under way in the IET facility should be useful to others in helping avoid costly mistakes because of the underutilization or misapplication of process automation. 6 figs.

  16. Transfer of radiocarbon liquid releases from the AREVA La Hague spent fuel reprocessing plant in the English Channel.

    PubMed

    Fiévet, Bruno; Voiseux, Claire; Rozet, Marianne; Masson, Michel; Bailly du Bois, Pascal

    2006-01-01

    The recent risk assessment by the North-Cotentin Radioecology Group (, 1999) outlined that (14)C has become one of the major sources of the low dose to man through seafood consumption. It was recommended that more data should be collected about (14)C in the local marine environment. The present study aims to respond to this recommendation. The estimation of (14)C activity in marine species is based on concentration factor values. The values reported here ranged from 1x10(3) to 5x10(3)Bqkg(-1)ww/BqL(-1). A comparison was made between the observed and predicted values. The accuracy of (14)C activity calculations was estimated between underestimation by a factor of 2 and over-estimation by 50% (95% confidence interval). However, the use of the concentration factor parameter is based on the biological and seawater compartments being in steady state. This assumption may not be met at short distances from the point of release of discharges, where rapid changes in seawater concentration may be smoothed out in living organisms due to transfer kinetics. The data processing technique, previously published by Fiévet and Plet (2003. Estimating biological half-lives of radionuclides in marine compartments from environmental time-series measurements. Journal of Environmental Radioactivity 65, 91-107), was used to deal with (14)C transfer kinetics, and carbon half-lives between seawater and a few biological compartments were thus estimated. PMID:16920235

  17. Development of the DIPRES process for the fast breeder reactor fuel cycle

    SciTech Connect

    Collins, E D; Jackson, M D; Griffin, C W; Rasmussen, D E; Norman, R E

    1984-01-01

    In 1979 the Consolidated Fuel Reprocessing Program (CFRP) at ORNL initiated a program for the development of advanced conversion processes with potential for simplifying and improving the conversion/pellet fabrication flowsheet for recycle plutonium. An evaluation of advanced conversion processes led to the selection of DIPRES (DIrect PREss Spheriodized) for development because it has the largest potential for process and product improvements. DIPRES utilizes a gel sphere conversion process and product to provide a spherical feed material for pellet fabrication. The free-flowing nature of the spherical conversion product allows it to be fed directly to pellet presses (i.e., direct press feed) in place of conventional, mechanically blended powder feed. This is advantageous for remote fabrication. The DIPRES feed is prepared by an internal gelation process.

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

    SciTech Connect

    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

  19. Alternative Fuel for Portland Cement Processing

    SciTech Connect

    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

  20. Method for photochemical reduction of uranyl nitrate by tri-N-butyl phosphate and application of this method to nuclear fuel reprocessing

    DOEpatents

    De Poorter, Gerald L.; Rofer-De Poorter, Cheryl K.

    1978-01-01

    Uranyl ion in solution in tri-n-butyl phosphate is readily photochemically reduced to U(IV). The product U(IV) may effectively be used in the Purex process for treating spent nuclear fuels to reduce Pu(IV) to Pu(III). The Pu(III) is readily separated from uranium in solution in the tri-n-butyl phosphate by an aqueous strip.

  1. Results from an initial re-processing of the British Isles continuous GNSS Facility (BIGF) archive of CGPS data for 1997 to 2010

    NASA Astrophysics Data System (ADS)

    Bingley, R.; Hansen, D. N.; Leighton, J.; Teferle, F. N.; David, B.

    2010-12-01

    The IESSG (Institute of Engineering Surveying and Space Geodesy) at the University of Nottingham operates and manages the NERC-funded British Isles continuous GNSS Facility (BIGF). BIGF provides a unique and secure repository for quality assured raw data, and derived products, from a network of continuous GNSS stations throughout the British Isles, and the interface with the scientific community, in serving demand for data and derived products to carry out research. The fundamental value of BIGF is encapsulated in the secure archive of 30 second GNSS RINEX data, supplied to it from currently 155 CGNSS stations. The archive comprises 1,000 station-years of 30 second, primarily GPS data, with some stations operating since 1996/7. Since 2009, BIGF has started to develop derived products, in the form of homogenous time series of parameters including station velocities, tropospheric integrated water vapour and ionospheric activity, to facilitate scientific users who are interested in these parameters but do not want to carry out their own high-level processing of GNSS data. This poster provides details of BIGF and presents the coordinate time series for the period from 1997 to 2010 obtained from an initial re-processing using an in-house modified version of the Bernese GPS software and the GIPSY/OASIS II software, along with re-processed IGS products. Details of how the derived station velocities have been used to form a map of vertical land movements and to study changes in sea levels around the British Isles are provided.

  2. Distributed generation - the fuel processing example

    SciTech Connect

    Victor, R.A.; Farris, P.J.; Maston, V.

    1996-12-31

    The increased costs of transportation and distribution are leading many commercial and industrial firms to consider the on-site generation for energy and other commodities used in their facilities. This trend has been accelerated by the development of compact, efficient processes for converting basic raw materials into finished services at the distributed sites. Distributed generation with the PC25{trademark} fuel cell power plant is providing a new cost effective technology to meet building electric and thermal needs. Small compact on-site separator systems are providing nitrogen and oxygen to many industrial users of these gases. The adaptation of the fuel processing section of the PC25 power plant for on-site hydrogen generation at industrial sites extends distributed generation benefits to the users of industrial hydrogen.

  3. Spectroscopic methods of process monitoring for safeguards of used nuclear fuel separations

    NASA Astrophysics Data System (ADS)

    Warburton, Jamie Lee

    To support the demonstration of a more proliferation-resistant nuclear fuel processing plant, techniques and instrumentation to allow the real-time, online determination of special nuclear material concentrations in-process must be developed. An ideal materials accountability technique for proliferation resistance should provide nondestructive, realtime, on-line information of metal and ligand concentrations in separations streams without perturbing the process. UV-Visible spectroscopy can be adapted for this precise purpose in solvent extraction-based separations. The primary goal of this project is to understand fundamental URanium EXtraction (UREX) and Plutonium-URanium EXtraction (PUREX) reprocessing chemistry and corresponding UV-Visible spectroscopy for application in process monitoring for safeguards. By evaluating the impact of process conditions, such as acid concentration, metal concentration and flow rate, on the sensitivity of the UV-Visible detection system, the process-monitoring concept is developed from an advanced application of fundamental spectroscopy. Systematic benchtop-scale studies investigated the system relevant to UREX or PUREX type reprocessing systems, encompassing 0.01-1.26 M U and 0.01-8 M HNO3. A laboratory-scale TRansUranic Extraction (TRUEX) demonstration was performed and used both to analyze for potential online monitoring opportunities in the TRUEX process, and to provide the foundation for building and demonstrating a laboratory-scale UREX demonstration. The secondary goal of the project is to simulate a diversion scenario in UREX and successfully detect changes in metal concentration and solution chemistry in a counter current contactor system with a UV-Visible spectroscopic process monitor. UREX uses the same basic solvent extraction flowsheet as PUREX, but has a lower acid concentration throughout and adds acetohydroxamic acid (AHA) as a complexant/reductant to the feed solution to prevent the extraction of Pu. By examining

  4. Evaluation of radioactivity release at Rokkasho reprocessing plant

    SciTech Connect

    Sugiyama, Hiroshi; Ishihara, Noriyuki; Maki, Akira

    2007-07-01

    JNFL have been conducting Active Test with spent fuels at Rokkasho Reprocessing Plant (RRP). In Active Test, the evaluation of radioactivity release to the environment (atmosphere and sea) was obtained. (authors)

  5. Pyroprocess for processing spent nuclear fuel

    DOEpatents

    Miller, William E.; Tomczuk, Zygmunt

    2002-01-01

    This is a pyroprocess for processing spent nuclear fuel. The spent nuclear fuel is chopped into pieces and placed in a basket which is lowered in to a liquid salt solution. The salt is rich in ZrF.sub.4 and containing alkali or alkaline earth fluorides, and in particular, the salt chosen was LiF-50 mol % ZrF.sub.4 with a eutectic melting point of 500.degree. C. Prior to lowering the basket, the salt is heated to a temperature of between 550.degree. C. and 700.degree. C. in order to obtain a molten solution. After dissolution the oxides of U, Th, rare earth and other like oxides, the salt bath solution is subject to hydro-fluorination to remove the oxygen and then to a fluorination step to remove U as gaseous UF.sub.6. In addition, after dissolution, the basket contains PuO.sub.2 and undissolved parts of the fuel rods, and the basket and its contents are processed to remove the Pu.

  6. Study of safeguards system on dry reprocessing for fast breeder reactor

    SciTech Connect

    Li, T. K.; Burr, Tom; Menlove, Howard O.; Thomas, K. E.; Fukushima, M.; Hori, M.

    2002-01-01

    A 'Feasibility Study on the Commercialized Fast Breeder Reactor (FBR) Cycle System' is underway at Japan Nuclear Cycle Development Institute (JNC). Concepts to commercialize the FBR fuel cycle are being created together with their necessary research and development (R&D) tasks. 'Dry,' non-aqueous, processes are candidates for FBR fuel reprocessing. Dry reprocessing technology takes advantage of proliferation barriers, due to the lower decontamination factors achievable by the simple pyrochemical processes proposed. The concentration o f highly radioactive impurities and non-fissile materials in products from a dry reprocess is generally significantly larger than the normal aqueous (Purex) process. However, the safeguards of dry reprocesses have not been widely analyzed. In 2000, JNC and Los Alamos National Laboratoiy (LANL) initiated a joint research program to study the safeguards aspects of dry reprocessing. In this study, the safeguardability of the three options: metal electrorefining, oxide electrowinning, and fluoride volatility processes, are assessed. FBR spent fuels are decladded and powdered into mixed oxides (MOX) at the Head-End process either by oxidation-reduction reactions (metal electrorefining and fluoride volatility) or mechanically (oxide electrowinning). At the oxide electrowinning process, the spent MOX he1 powder is transferred to chloride in molten salt and nuclear materials are extracted onto cathode as oxides. For metal electrorefining process, on the other hand, the MOX fuel is converted to chloride in molten salt, and nuclear materials are extracted onto cathode as a metal fomi. At lhe fluoride volatility process, the MOX fuel powder is converted to U&/PuF6 (gaseous form) in a fluidized bed; plutonium and uranium fluorides are separated by volatilization properties and then are converted to oxides. Since the conceptual design of a dry reprocessing plant is incomplete, the operational mode, vessel capacities, residence times, and campaigns

  7. CONSTRUCTION PROGRESS PHOTO SHOWING EXCAVATION PIT FOR MAIN PROCESSING BUILDING ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    CONSTRUCTION PROGRESS PHOTO SHOWING EXCAVATION PIT FOR MAIN PROCESSING BUILDING (CPP-601) LOOKING SOUTH. INL PHOTO NUMBER NRTS-50-693. Unknown Photographer, 1950 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  8. Situ process for making multifunctional fuel additives

    SciTech Connect

    Carrier, R.C.; Allen, B.R.

    1984-02-28

    Disclosed is an in situ or ''one pot'' process for making a fuel additive comprising reacting an excess of at least one N-primary alkylalkylene diamine with maleic anhydride in the presence of from 20 to 36 weight percent of a mineral oil reaction diluent at a temperature ranging from ambient to about 225/sup 0/ F. and recovering a product containing a primary aliphatic hydrocarbon amino alkylene substituted asparagine, an N-primary alkylalkylene diamine in the reaction oil with the product having a by-product succinimide content not in excess of 1.0 weight percent, based on the weight of asparagine present.

  9. In Situ fuel processing in a microbial fuel cell.

    PubMed

    Bahartan, Karnit; Amir, Liron; Israel, Alvaro; Lichtenstein, Rachel G; Alfonta, Lital

    2012-09-01

    A microbial fuel cell (MFC) was designed in which fuel is generated in the cell by the enzyme glucoamylase, which is displayed on the surface of yeast. The enzyme digests starch specifically into monomeric glucose units and as a consequence enables further glucose oxidation by microorganisms present in the MFC anode. The oxidative enzyme glucose oxidase was coupled to the glucoamylase digestive enzyme. When both enzymes were displayed on the surface of yeast cells in a mixed culture, superior fuel-cell performance was observed in comparison with other combinations of yeast cells, unmodified yeast, or pure enzymes. The feasibility of the use of the green macroalgae Ulva lactuca in such a genetically modified MFC was also demonstrated. Herein, we report the performance of such fuel cells as a proof of concept for the enzymatic digestion of complex organic fuels in the anode of MFCs to render the fuel more available to microorganisms. PMID:22833422

  10. Alternative fuels utilization and the automotive emission certification process

    SciTech Connect

    Not Available

    1980-03-01

    The Clean Air Act of 1977 requires that commercially offered automotive fuels and fuel additives be substantially similar to fuels used in certifying model year 1975 and later vehicles. Procedures for certifying that vehicles perform with emissions that meet the Clean Air Act specifications and the impact of this emissions certification process on the use of alternative fuels, such as alcohols, alcohol-gasoline blends and synthetic fuels, in highway vehicles is discussed. (LCL)

  11. Retrospective CMORPH Reprocessing Efforts

    NASA Astrophysics Data System (ADS)

    Yarosh, Y.; Joyce, R.; Xie, P.

    2008-05-01

    constellation, there is enough to retrospectively reprocess CMORPH well beyond the current archive start. Also IR based PMW calibrated rainfall estimates will be calculated as part of the retrospective reprocessing. These estimates will be blended for times and locations that the PMW information is too old for relative accuracy. This blended method (CMORPH-IR) combines the CMORPH and IR based estimates via an error model developed by running test CMORPH processing, albeit withholding random high quality PMW estimates, and determining the error/skill of the CMORPH relative to the IR-based rainfall as a function of season, surface type, region, and age of PMW information in half hourly increments from PMW scan time. The retrospective processing will be performed for Year 2002 and proceed backward. Detailed results will be reported at the meeting.

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

    SciTech Connect

    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).

  13. MOPITT V5 reprocessing

    Atmospheric Science Data Center

    2013-08-06

    ... parameters.   While the MOPITT team believes these technical problems would only concern a small number of MOPITT data users, ... the text string 'L1V3.36' whereas the reprocessed L1 files include 'L1V3.37'. The original L2 filenames included the text string ...

  14. MONOLITHIC FUEL FABRICATION PROCESS DEVELOPMENT AT THE IDAHO NATIONAL LABORATORY

    SciTech Connect

    Glenn A. Moore; Francine J. Rice; Nicolas E. Woolstenhulme; W. David SwanK; DeLon C. Haggard; Jan-Fong Jue; Blair H. Park; Steven E. Steffler; N. Pat Hallinan; Michael D. Chapple; Douglas E. Burkes

    2008-10-01

    Within the Reduced Enrichment for Research and Test Reactors (RERTR) program directed by the US Department of Energy (DOE), UMo fuel-foils are being developed in an effort to realize high density monolithic fuel plates for use in high-flux research and test reactors. Namely, targeted are reactors that are not amenable to Low Enriched Uranium (LEU) fuel conversion via utilization of high density dispersion-based fuels, i.e. 8-9 gU/cc. LEU conversion of reactors having a need for >8-9 gU/cc fuel density will only be possible by way of monolithic fuel forms. The UMo fuel foils under development afford fuel meat density of ~16 gU/cc and thus have the potential to facilitate LEU conversions without any significant reactor-performance penalty. Two primary challenges have been established with respect to UMo monolithic fuel development; namely, fuel element fabrication and in-reactor fuel element performance. Both issues are being addressed concurrently at the Idaho National Laboratory. An overview is provided of the ongoing monolithic UMo fuel development effort at the Idaho National Laboratory (INL); including development of complex/graded fuel foils. Fabrication processes to be discussed include: UMo alloying and casting, foil fabrication via hot rolling, fuel-clad interlayer application via co-rolling and thermal spray processes, clad bonding via Hot Isostatic Pressing (HIP) and Friction Bonding (FB), and fuel plate finishing.

  15. Overview of fuel processing options for polymer electrolyte fuel cell systems

    SciTech Connect

    Kumar, R.

    1995-12-31

    The polymer electrolyte fuel cell (PEFC) is being developed for use in heavy- and light-duty transportation applications. While this fuel cell has been used successfully in buses and vans with compressed hydrogen as the on-board fuel [1,2], the fuel cell system must incorporate fuel processing (reforming) for any other on-board fuel to produce the hydrogen or hydrogen-rich fuel gas to be fed to the fuel cell stack. This is true even for alternative methods of storing hydrogen, such as use of a metal hydride or liquefied hydrogen. The ``fuel processing`` needed to recover the hydrogen includes providing the heat of dissociation of the hydride and cooling the hydrogen to the temperature of the fuel cell stack. Discussed below are some of the options being considered for processing of on-board fuels (other than compressed hydrogen) to generate the fuel cell anode gas, and the effects of fuel processing on system design, efficiency, steady-state and dynamic performance, and other factors.

  16. CONSTRUCTION PROGRESS PHOTO SHOWING MAIN PROCESSING BUILDING (CPP601) LOOKING NORTH. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    CONSTRUCTION PROGRESS PHOTO SHOWING MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTH. INL PHOTO NUMBER NRTS-51-1387. Unknown Photographer, 1/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  17. CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP601) LOOKING NORTHWEST. INL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTHWEST. INL PHOTO NUMBER NRTS-51-1390. Unknown Photographer, 1/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  18. CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP601) LOOKING EAST. INL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP-601) LOOKING EAST. INL PHOTO NUMBER NRTS-51-1547. Unknown Photographer, 2/28/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  19. SOUTH ELEVATION OF MAIN PROCESSING BUILDING (CPP601) LOOKING NORTH. INL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    SOUTH ELEVATION OF MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTH. INL PHOTO NUMBER HD-22-5-3. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  20. Process for producing fluid fuel from coal

    DOEpatents

    Hyde, Richard W.; Reber, Stephen A.; Schutte, August H.; Nadkarni, Ravindra M.

    1977-01-01

    Process for producing fluid fuel from coal. Moisture-free coal in particulate form is slurried with a hydrogen-donor solvent and the heated slurry is charged into a drum wherein the pressure is so regulated as to maintain a portion of the solvent in liquid form. During extraction of the hydrocarbons from the coal, additional solvent is added to agitate the drum mass and keep it up to temperature. Subsequently, the pressure is released to vaporize the solvent and at least a portion of the hydrocarbons extracted. The temperature of the mass in the drum is then raised under conditions required to crack the hydrocarbons in the drum and to produce, after subsequent stripping, a solid coke residue. The hydrocarbon products are removed and fractionated into several cuts, one of which is hydrotreated to form the required hydrogen-donor solvent while other fractions can be hydrotreated or hydrocracked to produce a synthetic crude product. The heaviest fraction can be used to produce ash-free coke especially adapted for hydrogen manufacture. The process can be made self-sufficient in hydrogen and furnishes as a by-product a solid carbonaceous material with a useful heating value.

  1. SAFEGUARDS EXPERIENCE ON THE DUPIC FUEL CYCLE PROCESS

    SciTech Connect

    J. HONG; H. KIM; ET AL

    2001-02-01

    Safeguards have been applied to the R and D process for directly fabricating CANDU fuel with PWR spent fuel material. Safeguards issues to be resolved were identified in the areas such as international cooperation on handling foreign origin nuclear material, technology development of operator's measurement system of the bulk handling process of spent fuel material, and a built-in C/S system for independent verification of material flow. The fuel cycle concept (Direct Use of PWR spent fuel in CANDU, DUPIC) was developed in consideration of reutilization of over-flowing spent fuel resources at PWR sites and a reduction of generated high level wastes. All those safeguards issues have been finally resolved, and the first batch of PWR spent fuel material was successfully introduced in the DUPIC lab facility and has been in use for routine process development.

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

    SciTech Connect

    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.

  3. 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

    SciTech Connect

    Riley, Brian J.; Pierce, David A.; Frank, Steven M.; Matyáš, Josef; Burns, Carolyne A.

    2015-04-01

    This paper describes the various approaches attempted to make solution-derived sodalite with a LiCl-Li2O oxide reduction salt used to dissolve used uranium oxide fuel so the uranium can be recovered and recycled. The approaches include modified sol-gel and solutionbased 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 in the crystalline product (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt but that the incorporation of Li into the sodalite is low.

  4. 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

    DOE PAGESBeta

    Riley, Brian J.; Pierce, David A.; Frank, Steven M.; Matyáš, Josef; Burns, Carolyne A.

    2015-04-01

    This paper describes the various approaches attempted to make solution-derived sodalite with a LiCl-Li2O oxide reduction salt used to dissolve used uranium oxide fuel so the uranium can be recovered and recycled. The approaches include modified sol-gel and solutionbased 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 Li2Omore » 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 in the crystalline product (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt but that the incorporation of Li into the sodalite is low.« less

  5. 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

    SciTech Connect

    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.

  6. 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

    NASA Astrophysics Data System (ADS)

    Riley, Brian J.; Pierce, David A.; Frank, Steven M.; Matyáš, Josef; Burns, Carolyne A.

    2015-04-01

    This paper describes the various approaches evaluated for making solution-derived sodalite with a LiCl-Li2O 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 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 in the crystalline product (∼92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt but that the incorporation of Li into the sodalite is low.

  7. Fuel Cell Stations Automate Processes, Catalyst Testing

    NASA Technical Reports Server (NTRS)

    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. Reprocessed uranium exposure and lung cancer risk.

    PubMed

    Canu, Irina Guseva; Jacob, Sophie; Cardis, Elisabeth; Wild, Pascal; Caër-Lorho, Sylvaine; Auriol, Bernard; Laurier, Dominique; Tirmarche, Margot

    2010-09-01

    This study investigated the risk of lung cancer in regards to protracted occupational exposure to reprocessed uranium compounds. Two thousand seven hundred and nine male workers employed at the AREVA NC uranium processing plant between 1960 and 2005 in France were included in the cohort. Historical exposure to reprocessed uranium compounds classified by their solubility type was assessed on the basis of the plant's specific job-exposure matrix. Cox proportional hazard models adjusted for attained age, calendar period, and socioeconomic status were used to estimate relative risks in regards of each type of uranium compound. The relative risk of lung cancer tended to increase with decreasing solubility of reprocessed uranium compounds. The highest-though not statistically significant-relative risk was observed among workers exposed to slowly soluble reprocessed uranium dioxide. This study is the first suggesting an increasing risk of lung cancer associated with exposure to reprocessed uranium. Our results are consistent with data from experimental studies of biokinetics and the action mechanism of slowly soluble uranium compounds, but need to be confirmed in larger studies with more detailed dose-response analyses. PMID:20699691

  9. Renewable hydrogen production for fossil fuel processing

    SciTech Connect

    Greenbaum, E.; Lee, J.W.; Tevault, C.V.

    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.

  10. Solid oxide fuel cell process and apparatus

    DOEpatents

    Cooper, Matthew Ellis; Bayless, David J.; Trembly, Jason P.

    2011-11-15

    Conveying gas containing sulfur through a sulfur tolerant planar solid oxide fuel cell (PSOFC) stack for sulfur scrubbing, followed by conveying the gas through a non-sulfur tolerant PSOFC stack. The sulfur tolerant PSOFC stack utilizes anode materials, such as LSV, that selectively convert H.sub.2S present in the fuel stream to other non-poisoning sulfur compounds. The remaining balance of gases remaining in the completely or near H.sub.2S-free exhaust fuel stream is then used as the fuel for the conventional PSOFC stack that is downstream of the sulfur-tolerant PSOFC. A broad range of fuels such as gasified coal, natural gas and reformed hydrocarbons are used to produce electricity.

  11. 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

    SciTech Connect

    Amoroso, J. W.; Marra, J. C.

    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).

  12. 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

    SciTech Connect

    Amoroso, J. W.; Marra, J. C.

    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).

  13. A high quality reprocessed ground-based GPS dataset for atmospheric process studies, radiosonde and model evaluation, and reanalysis of HYMEX Special Observing Period

    NASA Astrophysics Data System (ADS)

    BOCK, Olivier; Bosser, Pierre; Pacione, Rosa; Nuret, Mathieu; Fourrié, Nadia

    2015-04-01

    Data from more than 1000 ground-based GPS receivers in the north-western Mediterranean area have been reprocessed in a consistent way using GIPSY-OASIS II software for the period from 1st September 2012 to 31 March 2013 which encompasses the Special Observation Periods (SOPs) 1 and 2 of the HYMEX project. The reprocessed GPS ZTD data were screened converted to IWV. The ZTD data were used to assess the accuracy of the near real time ZTD data assimilated for operational weather forecasting. The mean of delay differences between the operational and reprocessed solutions is about 0 +/- 3 mm (mean +/- standard deviation of bias over all stations) and the standard deviation of delay differences ranges between 4 and 8 mm. Significant bias reduction is thus expected from a reanalysis ingesting the reprocessed delay data. Various methods and auxiliary data (surface pressure and weighted mean temperature) are investigated for the conversion of ZTD data into IWV. The final IWV dataset is used to evaluate radiosonde humidity observations and operational analyses produced with the AROME model. The spatial and temporal distribution of IWV is also studied with a focus on heavy precipitation events in the north-western Mediterranean area during the HYMEX SOP1.

  14. Transformative monitoring approaches for reprocessing.

    SciTech Connect

    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.

  15. TIGA Tide Gauge Data Reprocessing at GFZ

    NASA Astrophysics Data System (ADS)

    Deng, Zhiguo; Schöne, Tilo; Gendt, Gerd

    2014-05-01

    To analyse the tide gauge measurements for the purpose of global long-term sea level change research a well-defined absolute reference frame is required by oceanographic community. To create such frame the data from a global GNSS network located at or near tide gauges are processed. For analyzing the GNSS data on a preferably continuous basis the International GNSS Service (IGS) Tide Gauge Benchmark Monitoring Working Group (TIGA-WG) is responsible. As one of the TIGA Analysis Centers the German Research Centre for Geosciences (GFZ) is contributing to the IGS TIGA Reprocessing Campaign. The solutions of the TIGA Reprocessing Campaign will also contribute to 2nd IGS Data Reprocessing Campaign with GFZ IGS reprocessing solution. After the first IGS reprocessing finished in 2010 some improvements were implemented into the latest GFZ software version EPOS.P8: reference frame IGb08 based on ITRF2008, antenna calibration igs08.atx, geopotential model (EGM2008), higher-order ionospheric effects, new a priori meteorological model (GPT2), VMF mapping function, and other minor improvements. GPS data of the globally distributed tracking network of 794 stations for the time span from 1994 until end of 2012 are used for the TIGA reprocessing. To handle such large network a new processing strategy is developed and described in detail. In the TIGA reprocessing the GPS@TIGA data are processed in precise point positioning (PPP) mode to clean data using the IGS reprocessing orbit and clock products. To validate the quality of the PPP coordinate results the rates of 80 GPS@TIGA station vertical movement are estimated from the PPP results using Maximum Likelihood Estimation (MLE) method. The rates are compared with the solution of University of LaRochelle Consortium (ULR) (named ULR5). 56 of the 80 stations have a difference of the vertical velocities below 1 mm/yr. The error bars of PPP rates are significant larger than those of ULR5, which indicates large time correlated noise in

  16. Electrocatalytic and fuel processing studies for portable fuel cells

    NASA Astrophysics Data System (ADS)

    Matter, Paul H.

    In the field of catalysis, the development of alternative catalysts for the oxygen reduction reaction (ORR) in Polymer Electrolyte Membrane Fuel Cell (PEMFC) cathodes has been an ongoing task for researchers over the past two decades. PEM fuel cells are considered to be potential replacements for internal combustion engines in automobiles, and their reduced emissions and better efficiency would have huge payoffs for our environment, and in reducing our nation's dependence on foreign oil. To date, PEMFC cathode over-potentials are still significant, and the only materials discovered to be highly active and stable catalysts in an acidic environment are platinum-based. Despite several major advances in recent years in reducing platinum loading in fuel cell electrodes, the high expense and low availability of platinum will hinder the large-scale commercialization of PEM fuel cells. The most hopeful advances being made in replacing platinum are related to pyrolyzed organic macrocycles with transition metal centers (such as Fe or Co porphyrins and phthalocyanines). Encouragingly, it has recently been discovered that active electrodes could be prepared by heat-treating metal and nitrogen precursors (not necessarily organic macrocycles) together in the presence of a carbon support. In the first study of this dissertation, catalysts for the Oxygen Reduction Reaction (ORR) were prepared by the pyrolysis of acetonitrile over various supports. The supports used included Vulcan Carbon, high purity alumina, silica, magnesia, and these same supports impregnated with Fe, Co, or Ni in the form of acetate salt. The catalysts were characterized by BET surface area analysis, BJH Pore Size Distribution (PSD), conductivity testing, Transmission Electron Microscopy (TEM), Temperature Programmed Oxidation (TPO), Thermo-Gravimetric Analysis (TGA), X-Ray Diffraction (XRD), X-ray Photo-electron Spectroscopy (XPS), Mossbauer Spectroscopy, Rotating Disk Electrode (RDE) half cell testing, and

  17. 77 FR 823 - Guidance for Fuel Cycle Facility Change Processes

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-06

    ... Information DG-3037 was published in the Federal Register on July 14, 2011 (76 FR 41527). The public comment... COMMISSION Guidance for Fuel Cycle Facility Change Processes AGENCY: Nuclear Regulatory Commission. ACTION... issuing a new regulatory guide (RG) 3.74, ``Guidance for Fuel Cycle Facility Change Processes.''...

  18. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 4 2010-04-01 2010-04-01 false Reprocessing. 211.115 Section 211.115 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) DRUGS: GENERAL CURRENT GOOD MANUFACTURING PRACTICE FOR FINISHED PHARMACEUTICALS Production and Process Controls §...

  19. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 4 2011-04-01 2011-04-01 false Reprocessing. 211.115 Section 211.115 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) DRUGS: GENERAL CURRENT GOOD MANUFACTURING PRACTICE FOR FINISHED PHARMACEUTICALS Production and Process Controls §...

  20. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 21 Food and Drugs 4 2012-04-01 2012-04-01 false Reprocessing. 211.115 Section 211.115 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) DRUGS: GENERAL CURRENT GOOD MANUFACTURING PRACTICE FOR FINISHED PHARMACEUTICALS Production and Process Controls §...

  1. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 21 Food and Drugs 4 2013-04-01 2013-04-01 false Reprocessing. 211.115 Section 211.115 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) DRUGS: GENERAL CURRENT GOOD MANUFACTURING PRACTICE FOR FINISHED PHARMACEUTICALS Production and Process Controls §...

  2. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 4 2014-04-01 2014-04-01 false Reprocessing. 211.115 Section 211.115 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) DRUGS: GENERAL CURRENT GOOD MANUFACTURING PRACTICE FOR FINISHED PHARMACEUTICALS Production and Process Controls §...

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

    SciTech Connect

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

    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.

  4. Fuel quality/processing study. Volume 4: On site processing studies

    NASA Technical Reports Server (NTRS)

    Jones, G. E., Jr.; Cutrone, M.; Doering, H.; Hickey, J.

    1981-01-01

    Fuel treated at the turbine and the turbine exhaust gas processed at the turbine site are studied. Fuel treatments protect the turbine from contaminants or impurities either in the upgrading fuel as produced or picked up by the fuel during normal transportation. Exhaust gas treatments provide for the reduction of NOx and SOx to environmentally acceptable levels. The impact of fuel quality upon turbine maintenance and deterioration is considered. On site costs include not only the fuel treatment costs as such, but also incremental costs incurred by the turbine operator if a turbine fuel of low quality is not acceptably upgraded.

  5. Analysis of nuclear proliferation resistance reprocessing and recycling technologies

    SciTech Connect

    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

  6. PLOT PLAN OF FUEL STORAGE BUILDING (CPP603) SHOWING STORAGE BASINS ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PLOT PLAN OF FUEL STORAGE BUILDING (CPP-603) SHOWING STORAGE BASINS AND PROPOSED LOCATION OF FUEL ELEMENT CUTTING FACILITY. INL DRAWING NUMBER 200-0603-00-706-051287. ALTERNATE ID NUMBER CPP-C-1287. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  7. FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP603) SHOWING STORAGE BASINS, ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP-603) SHOWING STORAGE BASINS, FUEL ELEMENT CUTTING FACILITY, AND DRY GRAPHITE STORAGE FACILITY. INL DRAWING NUMBER 200-0603-00-030-056329. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  8. Development of remote disassembly technology for liquid-metal reactor (LMR) fuel

    SciTech Connect

    Bradley, E.C.; Evans, J.H.; Metz, C.F. III; Weil, B.S.

    1990-01-01

    A major objective of the Consolidated Fuel Reprocessing Program (CFRP) is to develop equipment and demonstrate technology to reprocess fast breeder reactor fuel. Experimental work on fuel disassembly cutting methods began in the 1970s. High-power laser cutting was selected as the preferred cutting method for fuel disassembly. Remotely operated development equipment was designed, fabricated, installed, and tested at Oak Ridge National Laboratory (ORNL). Development testing included remote automatic operation, remote maintenance testing, and laser cutting process development. This paper summarizes the development work performed at ORNL on remote fuel disassembly. 2 refs., 1 fig.

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

    SciTech Connect

    Li, T.K.; Klosterbuer, S.F.; Menlove, H.O.

    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.

  10. Specialized Disposal Sites for Different Reprocessing Plant Wastes

    SciTech Connect

    Forsberg, Charles W.; Driscoll, Michael J.

    2007-07-01

    Once-through fuel cycles have one waste form: spent nuclear fuel (SNF). In contrast, the reprocessed SNF yields multiple wastes with different chemical, physical, and radionuclide characteristics. The different characteristics of each waste imply that there are potential cost and performance benefits to developing different disposal sites that match the disposal requirements of different waste. Disposal sites as defined herein may be located in different geologies or in a single repository containing multiple sections, each with different characteristics. The paper describes disposal options for specific wastes and the potential for a waste management system that better couples various reprocessing plant wastes with disposal facilities. (authors)

  11. Head-end reprocessing equipment remote maintenance demonstration

    SciTech Connect

    Evans, J.H.; Metz, C.F. III.

    1989-01-01

    Prototype equipment for reprocessing breeder reactor nuclear fuel was installed in the Remote Operation and Maintenance Demonstration (ROMD) area of the Consolidated Fuel Reprocessing Program (CFRP) facility at the Oak Ridge National Laboratory (ORNL) in order to evaluate the design of this equipment in a cold mock-up of a remotely maintained hot cell. This equipment included the Remote Disassembly System (RDS) and the Remote Shear System (RSS). These systems were disassembled and reassembled remotely by using the extensive remote handling systems that are installed in this simulated hot-cell environment. 5 refs., 5 figs.

  12. Fuel quality-processing study. Volume 2: Literature survey

    NASA Technical Reports Server (NTRS)

    Jones, G. E., Jr.; Amero, R.; Murthy, B.; Cutrone, M.

    1981-01-01

    The validity of initial assumptions about raw materials choices and relevant upgrading processing options was confirmed. The literature survey also served to define the on-site (at the turbine location) options for fuel treatment and exhaust gas treatment. The literature survey also contains a substantial compilation of specification and physical property information about liquid fuel products relevant to industrial gas turbines.

  13. Treatment of oxide spent fuel using the lithium reduction process

    SciTech Connect

    Karell, E.J.; Pierce, R.D.; Mulcahey, T.P.

    1996-05-01

    The wide variety in the composition of DOE spent nuclear fuel complicates its long-term disposition because of the potential requirement to individually qualify each type of fuel for repository disposal. Argonne National Laboratory (ANL) has developed the electrometallurgical treatment technique to convert all of these spent fuel types into a single set of disposal forms, simplifying the qualification process. While metallic fuels can be directly processed using the electrometallurgical treatment technique, oxide fuels must first be reduced to the metallic form. The lithium reduction process accomplishes this pretreatment. In the lithium process the oxide components of the fuel are reduced using lithium at 650 C in the presence of molten LiCl, yielding the corresponding metals and Li{sub 2}O. The reduced metal components are then separated from the LiCl salt phase and become the feed material for electrometallurgical treatment. A demonstration test of the lithium reduction process was successfully conducted using a 10-kg batch of simulated oxide spent fuel and engineering-scale equipment specifically constructed for that purpose. This paper describes the lithium process, the equipment used in the demonstration test, and the results of the demonstration test.

  14. Collocation and integration of reprocessing and repositories: implications for aqueous flowsheets and waste management

    SciTech Connect

    Forsberg, C.; Lewis, L.

    2013-07-01

    It is an accident of history that the current model of the fuel cycle is a separate set of facilities connected by transportation. The question is whether collocation and integration of reprocessing and fuel fabrication with the repository significantly reduce the costs of a closed fuel cycle while improving system performance in terms of safety and long-term repository performance. This paper examines the question in terms of higher-level functional requirements of reprocessing systems and geological repositories.

  15. CONSTRUCTION PROGRESS PHOTO SHOWING EMPLACEMENT STEEL BEAMS FUEL STORAGE BUILDING ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    CONSTRUCTION PROGRESS PHOTO SHOWING EMPLACEMENT STEEL BEAMS FUEL STORAGE BUILDING (CPP-603) LOOKING EAST. INL PHOTO NUMBER NRTS-51-1371. Unknown Photographer, 1/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  16. Economic prospects of the Integral Fast Reactor (IFR) fuel cycle

    SciTech Connect

    Chang, Y.I.; Till, C.E.

    1991-01-01

    The IFR fuel cycle based on pyroprocessing involves only few operational steps and the batch-oriented process equipment systems are compact. This results in major cost reductions in all of three areas of reprocessing, fabrication, and waste treatment. This document discusses the economic aspects of this fuel cycle.

  17. Optimizing near real time accountability for reprocessing.

    SciTech Connect

    Cipiti, Benjamin B.

    2010-06-01

    Near Real Time Accountability (NRTA) of actinides at high precision in reprocessing plants has been a long sought-after goal in the safeguards community. Achieving this goal is hampered by the difficulty of making precision measurements in the reprocessing environment, equipment cost, and impact to plant operations. Thus the design of future reprocessing plants requires an optimization of different approaches. The Separations and Safeguards Performance Model, developed at Sandia National Laboratories, was used to evaluate a number of NRTA strategies in a UREX+ reprocessing plant. Strategies examined include the incorporation of additional actinide measurements of internal plant vessels, more use of process monitoring data, and the option of periodic draining of inventory to key tanks. Preliminary results show that the addition of measurement technologies can increase the overall measurement uncertainty due to additional error propagation, so care must be taken when designing an advanced system. Initial results also show that relying on a combination of different NRTA techniques will likely be the best option. The model provides a platform for integrating all the data. The modeling results for the different NRTA options under various material loss conditions will be presented.

  18. Nuclear fuel waste management and disposal concept: Report. Federal environmental assessment review process

    SciTech Connect

    1998-09-01

    The Canadian concept for disposing CANDU reactor waste or high-level nuclear wastes from reprocessing involves underground disposal in sealed containers emplaced in buffer-filled and sealed vaults 500--1,000 meters below ground, in plutonic rock of the Canadian Shield. This document presents the report of a panel whose mandate was to review this concept (rather than a specific disposal project at a specific site) along with a broad range of related policy issues, and to conduct that review in five provinces (including reviews with First Nations groups). It first outlines the review process and then describes the nature of the problem of nuclear waste management. It then presents an overview of the concept being reviewed, its implementation stages, performance assessment analyses performed on the concept, and implications of a facility based on that concept (health, environmental, social, transportation, economic). The fourth section examines the criteria by which the safety and acceptability of the concept should be evaluated. This is followed by a safety and acceptability evaluation from both technical and social perspectives. Section six proposes future steps for building and determining acceptability of the concept, including an Aboriginal participation process, creation of a Nuclear Fuel Waste Management Agency, and a public participation process. The final section discusses some issues outside the panel`s mandate, such as energy policy and renewable energy sources. Appendices include a chronology of panel activities, a review of radiation hazards, comparison between nuclear waste management and the management of other wastes, a review of other countries` approaches to long-term management of nuclear fuel wastes, and details of a siting process proposed by the panel.

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

    DOEpatents

    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.

  20. Use of CAP88 PC to infer differences in the chemical form of 129I emitted from a fuel reprocessing facility.

    PubMed

    Fritz, Brad G; Phillips, Nathan R J

    2013-06-01

    Emissions of (129)I from nuclear fuel separations conducted at the Hanford Site in Washington State have been occurring since the 1940's. Fuel separation on the Hanford Site stopped in 1988, but emissions of (129)I have continued as venting of the PUREX Plant occurred. In this study, atmospheric measurements of (129)I concentrations were coupled with an EPA approved plume dispersion model (CAP88-PC, Version 3.0) to evaluate the effectiveness of the dispersion model for estimating ambient concentrations at the Hanford Site. This evaluation led to the hypothesis that different chemical forms of iodine were being emitted over the years; this hypothesis was developed as an explanation for the model agreeing with measurements over some time periods, but not over all time periods. The model was then run with modified emissions to simulate the short atmospheric half-life of the suspected reactive chemical form of iodine being emitted. This modification resulted in good agreement between the modeled and measured concentrations over the entire 20 year study period (1986-2005), and provided evidence supporting the hypothesis of a reactive form of iodine being emitted. PMID:23395750

  1. Use of CAP88 PC to infer differences in the chemical form of I-129 emitted from a fuel reprocessing facility

    SciTech Connect

    Fritz, Brad G.; Phillips, Nathan RJ

    2013-06-17

    Emissions of 129I from nuclear fuel separations conducted at the Hanford Site in Washington State have been occurring since the 1940’s. Fuel separation on the Hanford Site stopped in 1988, but emissions of 129I have continued as venting of the building occurred. In this study, atmospheric measurements of 129I concentrations were coupled with an EPA approved plume dispersion model (CAP-88PC) to evaluate the effectiveness of the dispersion model for estimating ambient concentrations at the Hanford Site. This evaluation led to the hypothesis of different chemical forms of iodine being emitted over the years; this hypothesis was developed as an explanation for the model agreeing with measurements over some time periods, but not over all time periods. The model was then run with modified emissions to simulate the short atmospheric half-life of the suspected reactive chemical form of iodine being emitted. This modification resulted in good agreement between the modeled and measured concentrations over the entire 20 year study period, and provided evidence that the hypothesis of a reactive form of iodine being emitted was correct.

  2. Electrochemical fluorination for processing of used nuclear fuel

    DOEpatents

    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.

  3. Synergize fuel and petrochemical processing plans with catalytic reforming

    SciTech Connect

    1997-03-01

    Depending on the market, refiner`s plans to produce clean fuels and higher value petrochemicals will weigh heavily on the catalytic reformer`s flexibility. It seems that as soon as a timely article related to catalytic reforming operations is published, a new {open_quotes}boutique{close_quotes} gasoline fuel specification is slapped on to existing fuel standards, affecting reformer operations and processing objectives. Just as importantly, the petrochemical market (such as aromatics) that refiners are targeting, can be very fickle. That`s why process engineers have endeavored to maintain an awareness of the flexibility that technology suppliers are building into modern catalytic reformers.

  4. Renovation of CPF (Chemical Processing Facility) for Development of Advanced Fast Reactor Fuel Cycle System

    SciTech Connect

    Shinichi Aose; Takafumi Kitajima; Kouji Ogasawara; Kazunori Nomura; Shigehiko Miyachi; Yoshiaki Ichige; Tadahiro Shinozaki; Shinichi Ohuchi

    2008-01-15

    CPF (Chemical Processing Facility) was constructed at Nuclear Fuel Cycle Engineering Laboratories of JAEA (Japan Atomic Energy Agency) in 1980 as a basic research field where spent fuel pins from fast reactor (FR) and high level liquid waste can be dealt with. The renovation consists of remodeling of the CA-3 cell and the laboratory A, installation of globe boxes, hoods and analytical equipments to the laboratory C and the analytical laboratory. Also maintenance equipments in the CA-5 cell which had been out of order were repaired. The CA-3 cell is the main cell in which important equipments such as a dissolver, a clarifier and extractors are installed for carrying out the hot test using the irradiated FR fuel. Since the CPF had specialized originally in the research function for the Purex process, it was desired to execute the research and development of such new, various reprocessing processes. Formerly, equipments were arranged in wide space and connected with not only each other but also with utility supply system mainly by fixed stainless steel pipes. It caused shortage of operation space in flexibility for basic experimental study. Old equipments in the CA-3 cell including vessels and pipes were removed after successful decontamination, and new equipments were installed conformably to the new design. For the purpose of easy installation and rearranging the experimental equipments, equipments are basically connected by flexible pipes. Since dissolver is able to be easily replaced, various dissolution experiments is conducted. Insoluble residue generated by dissolution of spent fuel is clarified by centrifugal. This small apparatus is effective to space-saving. Mini mixer settlers or centrifugal contactors are put on to the prescribed limited space in front of the backside wall. Fresh reagents such as solvent, scrubbing and stripping solution are continuously fed from the laboratory A to the extractor by the reagent supply system with semi-automatic observation

  5. Renovation of Chemical Processing Facility for Development of Advanced Fast Reactor Fuel Cycle System in JNC

    SciTech Connect

    Atsushi Aoshima; Shigehiko Miyachi; Takashi Suganuma; Shinichi Nemoto

    2002-07-01

    The CPF had 4 laboratories (operation room A, laboratory A, laboratory C and analysis laboratory) in connection with reprocessing technology. The main laboratory, operation room A, has 5 hot cells. Since equipment in the main cell had been designed for small-scale verification of existing reprocessing steps, it was hardly able to respond flexibly to experimental studies on advanced technology. It was decided to remodel the cell according to the design that was newly laid out in order to ensure the function and space to conduct various basic tests. The other laboratories had no glove boxes for conducting basic experiments of important elements in the advanced reprocessing, such as actinides except U and Pu, lanthanides and so on. In order to meet various requirements of innovative technologies on advanced fuel cycle development, one laboratory is established more for study on dry reprocessing, and glove boxes, hoods and analytical equipment such as NMR, FT-IR, TI-MS are newly installed in the other laboratories in this renovation. After the renovation, hot tests in the CPF will be resumed from April 2002. (authors)

  6. Full-scale experimental facility for the development technologies for the reprocessing of tritium contaminated light and heavy water wastes by CECE process and cryogenic distillation

    SciTech Connect

    Trenin, V.D.; Alekseev, I.A.; Karpov, S.P.; Bondarenko, S.D.; Vasyanina, T.V.; Konoplev, K.A.; Fedorchenko, O.A.; Uborski, V.V.; Voronina, T.

    1995-10-01

    The problem of the formation and accumulation of the tritiated heavy and light water wastes produced under operation of the various nuclear facilities is considered. It is shown that the tritium contaminated wastes may have a wide spectrum of isotope concentrations of H:D:T and correlation one with other. Reprocessing of these wastes is expensive matter due to the small tritium concentration respectfully to other hydrogen isotopes and as well as the small value of separation factor. It requires the development of the versatile technology. The description of the full scale experimental facility constructed at PNPI is given. 18 refs., 1 fig.

  7. Joint Reprocessing of GPS, GLONASS and SLR

    NASA Astrophysics Data System (ADS)

    Fritsche, Mathias; Rodriguez-Solano, Carlos; Steigenberger, Peter; Sosnica, Krzysztof; Wang, Kan; Dietrich, Reinhard; Hugentobler, Urs; Dach, Rolf; Rothacher, Markus

    2013-04-01

    A joint reprocessing of GPS, GLONASS and SLR observations has been carried out at TU Dresden, TU Munich, AIUB and ETH Zurich. Common a priori models have been applied for the processing of all types of observation to ensure both consistent parameter estimates and the rigorous combination of microwave and optical measurements. Based on that reprocessing results, we evaluate the impact of adding GLONASS observations to the standard GPS data processing. In particular, changes in station position time series and day boundary overlaps of consecutive satellite arcs are analyzed. In addition, the GNSS orbits derived from microwave measurements are validated using independent SLR range measurements. Our SLR residuals indicate a significant improvement compared to previous results. Furthermore, we evaluate the performance of our high-rate (30s) combined GNSS satellite clocks and discuss associated zero-difference phase residuals.

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

    SciTech Connect

    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)

  9. Analysis of the ATR fuel element swaging process

    SciTech Connect

    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.

  10. Process for separating water and solids from fuels

    SciTech Connect

    Filho, J.H.; Bachmann, D.L.

    1987-11-17

    A process for separating water and solid particles from a fuel oil feedstock is described comprising: subjecting the feedstock to a first separation in a scroll type centrifugal separator to form a first recovered fuel stream and an oil cake; subjecting at least the first recovered fuel stream to a second separation in a centrifugal disc separator to form a clean fuel stream, an oil-bearing water stream and a sludge stream; treating the oil-bearing water stream in a separator to recover the oil; treating the oil cake removed from the first separation with a solvent in order to form a suspension; mixing the suspension with the sludge stream to form a mixture; feeding the mixture to a filter press to yield a solid reject and a filtrate; separating the filtrate into a decantate and recovered oil; mixing at least a portion of the recovered oil with first recovered fuel stream to form a semi-cleaned fuel stream; and subjecting the semi-cleaned fuel stream to the second separation in a centrifugal disc separator to form the clean fuel stream.

  11. Optimum catalytic process for alcohol fuels from syngas

    SciTech Connect

    Not Available

    1990-04-28

    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. Methanol production and heterogeneous catalysis utilizing transition elements supported on metal oxides with spinel structure are discussed. 12 figs., 16 tabs.

  12. Process of producing liquid hydrocarbon fuels from biomass

    DOEpatents

    Kuester, James L.

    1987-07-07

    A continuous thermochemical indirect liquefaction process to convert various biomass materials into diesel-type transportation fuels which fuels are compatible with current engine designs and distribution systems comprising feeding said biomass into a circulating solid fluidized bed gasification system to produce a synthesis gas containing olefins, hydrogen and carbon monoxide and thereafter introducing the synthesis gas into a catalytic liquefaction system to convert the synthesis gas into liquid hydrocarbon fuel consisting essentially of C.sub.7 -C.sub.17 paraffinic hydrocarbons having cetane indices of 50+.

  13. Process of producing liquid hydrocarbon fuels from biomass

    DOEpatents

    Kuester, J.L.

    1987-07-07

    A continuous thermochemical indirect liquefaction process is described to convert various biomass materials into diesel-type transportation fuels which fuels are compatible with current engine designs and distribution systems comprising feeding said biomass into a circulating solid fluidized bed gasification system to produce a synthesis gas containing olefins, hydrogen and carbon monoxide and thereafter introducing the synthesis gas into a catalytic liquefaction system to convert the synthesis gas into liquid hydrocarbon fuel consisting essentially of C[sub 7]-C[sub 17] paraffinic hydrocarbons having cetane indices of 50+. 1 fig.

  14. Preferential oxidation of methanol and carbon monoxide for gas cleanup during methanol fuel processing

    SciTech Connect

    Birdsell, S.A.; Vanderborgh, N.E.; Inbody, M.A.

    1993-07-01

    Methanol fuel processing generates hydrogen for low-temperature, PEM fuel cell systems now being considered for transportation and other applications. Although liquid methanol fuel is convenient for this application, existing fuel processing techniques generate contaminants that degrade fuel cell performance. Through mathematical models and laboratory experiments chemical processing is described that removes CO and other contaminants from the anode feed stream.

  15. Microwave Processing of Simulated Advanced Nuclear Fuel Pellets

    SciTech Connect

    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.

  16. Process review of lignocellulose biochemical conversion to fuel ethanol

    Technology Transfer Automated Retrieval System (TEKTRAN)

    This review is meant to give a general technical review of the state-of-the-art in process technology for the biochemical conversion of lignocellulose to fuel ethanol. The proceeding details the chemical structure of biomass and basic process steps needed for extracting carbohydrates as sugars and ...

  17. Current Practice of Duodenoscope Reprocessing.

    PubMed

    Kim, Stephen; Muthusamy, V Raman

    2016-10-01

    Numerous outbreaks of duodenoscope-associated transmission of multi-drug resistant bacteria have recently been reported. Unlike prior episodes of endoscope-transmitted infections, the latest outbreaks have occurred despite strict adherence to duodenoscope reprocessing guidelines. The current standard for all flexible endoscope reprocessing includes pre-cleaning, leak testing, an additional manual cleaning step, and high-level disinfection. When these steps are strictly followed, the risk of infection transmission during endoscopy is exceedingly rare. However, due to its complex design, the duodenoscope may not be able to be adequately disinfected using the current reprocessing standards. Supplemental measures to enhance scope reprocessing have subsequently been recommended to reduce the infection risk in patients undergoing endoscopic retrograde cholangiopancreatography. These methods are likely short-term solutions that have yet to be validated regarded their effectiveness. Additional approaches to monitor the quality of duodenoscope reprocessing may also be useful. Ultimately, a definitive, yet logistically feasible, method of duodenoscope reprocessing is required to ensure the safety of our patients. PMID:27595583

  18. Reprocessing single-use devices--the equipment connection.

    PubMed

    Dunn, Debra

    2002-06-01

    This is the second in a series of three articles about reprocessing medical devices labeled as "single use" by the manufacturer. The goal of reprocessing single-use devices (SUDs) is to save money and decrease environmental pollution. Reprocessing can be performed on SUDs that have been used on other patients or opened but not used. In this article, the procedures for reprocessing (e.g., cleaning, inspecting, sterilizing, tracking, testing, validating) and establishing a reuse program are discussed. The first article of the series, published in the May 2002 issue of the Journal, discussed the ethical component of reusing SUDs. The third article, to be published in the July 2002 issue, will discuss the roles of the involved regulatory agencies and organizations guiding the process. PMID:12085405

  19. CONVERTING PYROLYSIS OILS TO RENEWABLE TRANSPORT FUELS: PROCESSING CHALLENGES & OPPORTUNITIES

    SciTech Connect

    Holmgren, Jennifer; Nair, Prabhakar N.; Elliott, Douglas C.; Bain, Richard; Marinangelli, Richard

    2008-03-11

    To enable a sustained supply of biomass-based transportation fuels, the capability to process feedstocks outside the food chain must be developed. Significant industry efforts are underway to develop these new technologies, such as converting cellulosic wastes to ethanol. UOP, in partnership with U.S. Government labs, NREL and PNNL, is developing an alternate route using cellulosic feedstocks. The waste biomass is first subjected to a fast pyrolysis operation to generate pyrolysis oil (pyoil for short). Current efforts are focused on developing a thermochemical platform to convert pyoils to renewable gasoline, diesel and jet fuel. The fuels produced will be indistinguishable from their fossil fuel counterparts and, therefore, will be compatible with existing transport and distribution infrastructure.

  20. Processing of carbon composite paper as electrode for fuel cell

    NASA Astrophysics Data System (ADS)

    Mathur, R. B.; Maheshwari, Priyanka H.; Dhami, T. L.; Sharma, R. K.; Sharma, C. P.

    The porous carbon electrode in a fuel cell not only acts as an electrolyte and a catalyst support, but also allows the diffusion of hydrogen fuel through its fine porosity and serves as a current-carrying conductor. A suitable carbon paper electrode is developed and possesses the characteristics of high porosity, permeability and strength along with low electrical resistivity so that it can be effectively used in proton-exchange membrane and phosphoric acid fuel cells. The electrode is prepared through a combination of two important techniques, viz., paper-making technology by first forming a porous chopped carbon fibre preform, and composite technology using a thermosetting resin matrix. The study reveals an interdependence of one parameter on another and how judicious choice of the processing conditions are necessary to achieve the desired characteristics. The current-voltage performance of the electrode in a unit fuel cell matches that of a commercially-available material.

  1. Motor fuel alkylation process utilizing low acid

    SciTech Connect

    Kocal, J.A.; Imai, T.

    1987-01-06

    A process is described for the alkylation of an isoparaffin with an olefin acting agent comprising contacting the isoparaffin with the olefin acting agent at alkylation conditions in the presence of a catalyst. The catalyst consists essentially of an anhydrous, nonalcoholic mixture of from about 5 to 15 wt. % methyl tert-butyl ether and from 85 to 95 wt. % hydrofluoric acid. The volumetric ratio of hydrofluoric acid to isoparaffin and olefin acting agent is less than 0.75.

  2. Pyrolysis process for producing fuel gas

    NASA Technical Reports Server (NTRS)

    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.

  3. Processing waste fats into a fuel oil substitute

    SciTech Connect

    Pudel, F.; Lengenfeld, P.

    1993-12-31

    Waste fats have a high energy potential. They also contain impurities. For example, fats used for deep-frying contain high contents of solids, water, and chlorides. The process described in this paper removes the impurities by simple processing such as screening, washing, separating, drying, and filtering. The final quality of processed fat allows its use as a fuel oil substitute, and also as a raw material for chemical production.

  4. Reprocessing technology development for irradiated beryllium

    SciTech Connect

    Kawamura, H.; Sakamoto, N.; Tatenuma, K.

    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%.

  5. An assessment of the attractiveness of material associated with thorium/uranium and uranium closed fuel cycles from a safeguards perspective

    SciTech Connect

    Bathke, Charles Gary; Wallace, Richard K; Hase, Kevin R; Sleaford, Brad W; Ebbinghaus, Bartley B; Collins, Brian W; Bradley, Keith S; Prichard, Andrew W; Smith, Brian W

    2010-01-01

    This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with various proposed nuclear fuel cycles. Specifically, this paper examines two closed fuel cycles. The first fuel cycle examined is a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of plutonium/thorium and {sup 233}U/thorium. The used fuel is then reprocessed using the THOREX process and the actinides are recycled. The second fuel cycle examined consists of conventional light water reactors (LWR) whose fuel is reprocessed for actinides that are then fed to and recycled until consumed in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). As reprocessing of LWR fuel has already been examined, this paper will focus on the reprocessing of the scheme's fast-spectrum reactors' fuel. This study will indicate what is required to render these materials as having low utility for use in nuclear weapons. Nevertheless, the results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE). The methodology and key findings will be presented.

  6. Fuel ethanol production from crop residues and processing byproducts

    Technology Transfer Automated Retrieval System (TEKTRAN)

    In 2007, the production of fuel ethanol from corn starch reached 6.5 billion gallons in the U.S.A. Various crop residues such as corn stover, wheat straw, and barley straw, and crop processing byproducts such as corn fiber and rice hulls can serve as low-cost lignocellulosic feedstocks for conversi...

  7. 76 FR 44049 - Guidance for Fuel Cycle Facility Change Processes

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-22

    ... COMMISSION Guidance for Fuel Cycle Facility Change Processes AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; extension of comment period. SUMMARY: On July 14, 2011 (76 FR 41527), the U.S...: Richard.Jervey@nrc.gov . SUPPLEMENTARY INFORMATION: On July 14, 2011 (76 FR 41527), the NRC published...

  8. FUEL ETHANOL PRODUCTION FROM AGRICULTURAL RESIDUES AND PROCESSING BYPRODUCTS

    Technology Transfer Automated Retrieval System (TEKTRAN)

    In 2005, the production of fuel ethanol from corn starch reached 4.5 billion gallons in the U.S. Various agricultural residues such as corn stover and wheat straw, and agricultural processing byproducts such as corn fiber and rice hulls, can serve as low-cost lignocellulosic feedstocks for conversi...

  9. Fate of virginiamycin through the fuel ethanol production process

    Technology Transfer Automated Retrieval System (TEKTRAN)

    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 wa...

  10. PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

    DOEpatents

    Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

    1963-09-01

    A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

  11. OBLIQUE PHOTO OF NORTHWEST CORNER OF FUEL STORAGE BUILDING (CPP603). ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    OBLIQUE PHOTO OF NORTHWEST CORNER OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTHEAST. INL PHOTO NUMBER HD-54-14-4. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  12. OBLIQUE PHOTO OF NORTH ELEVATION OF FUEL STORAGE BUILDING (CPP603). ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    OBLIQUE PHOTO OF NORTH ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTH. INL PHOTO NUMBER HD-54-14-3. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  13. 0BLIQUE PHOTO OF EAST ELEVATION OF FUEL STORAGE BUILDING (CPP603). ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    0BLIQUE PHOTO OF EAST ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING WEST. INL PHOTO NUMBER HD-54-15-1. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  14. INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP603) LOOKING SOUTHWEST SHOWING ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP-603) LOOKING SOUTHWEST SHOWING STORAGE BASIN IN FOREGROUND, TRANSFER CRANE AND UNLOADER TO LEFT OF NORTH SIDE OF HOT CELL. INL PHOTO NUMBER NRTS-58-157. J. Anderson, Photographer, 1/15/1958 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  15. VIEW OF FECF HOT CELL OF FUEL STORAGE BUILDING (CPP603). ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    VIEW OF FECF HOT CELL OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORHTWEST. INL PHOTO NUMBER HD-54-18-3. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  16. Preparation of the Second Shipment of Spent Nuclear Fuel from the Ustav Jaderneho Vyzkumu Rez (UJV Rez), a.s., Czech Republic to the Russian Federation for Reprocessing - 13478

    SciTech Connect

    Trtilek, Radek; Podlaha, Josef

    2013-07-01

    After more than 50 years of operation of the LVR-15 research reactor operated by the UJV Rez, a. s. (formerly Nuclear Research Institute - NRI), a large amount of the spent nuclear fuel (SNF) of Russian origin has been accumulated. In 2005 UJV Rez, a. s. jointed the Russian Research Reactor Fuel Return (RRRFR) program under the United States (US) - Russian Global Threat Reduction Initiative (GTRI) and started the process of SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). In 2007 the first shipment of SNF was realized. In 2011, preparation of the second shipment of spent fuel from the Czech Republic started. The experience obtained from the first shipment will be widely used, but some differences must be taken into the account. The second shipment will be realized in 2013 and will conclude the return transport of all, both fresh and spent, high-enriched nuclear fuel from the Czech Republic to the Russian Federation. After the shipment is completed, there will be only low-enriched nuclear fuel on the territory of the Czech Republic, containing maximum of 20% of U-235, which is the conventionally recognized limit between the low- and high-enriched nuclear materials. The experience (technical, organizational, administrative, logistic) obtained from the each SNF shipment as from the Czech Republic as from other countries using the Russian type research reactors are evaluated and projected onto preparation of next shipment of high enriched nuclear fuel back to the Russian Federation. The results shown all shipments provided by the UJV Rez, a. s. in the frame of the GTRI Program have been performed successfully and safely. It is expected the experience and results will be applied to preparation and completing of the Chinese Miniature Neutron Source Reactors (MNSR) Spent Nuclear Fuel Repatriation in the near future. (authors)

  17. Idle reprocessing plant nuke waste fix moving

    SciTech Connect

    Not Available

    1984-02-01

    A system is discussed for cleanup at a defunct nuclear waste reprocessing plant in upstate New York. The new system, a solidification process, is designed to stabilize 600,000 gal of highly radioactive waste into glass logs. The design for a waste-solidification process consists of a slurry-fed, ceramic kiln where the molten borosilicate glass is mixed with wastes. The mix is poured into steel canisters where it hardens into glass logs to be buried at a federal depository. Also discussed is a new disposal method, called caisson waste disposal, for the low-level radioactive waste generated during cleanup.

  18. Method For Processing Spent (Trn,Zr)N Fuel

    DOEpatents

    Miller, William E.; Richmann, Michael K.

    2004-07-27

    A new process for recycling spent nuclear fuels, in particular, mixed nitrides of transuranic elements and zirconium. The process consists of two electrorefiner cells in series configuration. A transuranic element such as plutonium is reduced at the cathode in the first cell, zirconium at the cathode in the second cell, and nitrogen-15 is released and captured for reuse to make transuranic and zirconium nitrides.

  19. Vitrification of IFR and MSBR halide salt reprocessing wastes

    SciTech Connect

    Siemer, D.D.

    2013-07-01

    Both of the genuinely sustainable (breeder) nuclear fuel cycles (IFR - Integral Fast Reactor - and MSBR - Molten Salt Breeder Reactor -) studied by the USA's national laboratories would generate high level reprocessing waste (HLRW) streams consisting of a relatively small amount ( about 4 mole %) of fission product halide (chloride or fluoride) salts in a matrix comprised primarily (about 95 mole %) of non radioactive alkali metal halide salts. Because leach resistant glasses cannot accommodate much of any of the halides, most of the treatment scenarios previously envisioned for such HLRW have assumed a monolithic waste form comprised of a synthetic analog of an insoluble crystalline halide mineral. In practice, this translates to making a 'substituted' sodalite ('Ceramic Waste Form') of the IFR's chloride salt-based wastes and fluoroapatite of the MSBR's fluoride salt-based wastes. This paper discusses my experimental studies of an alternative waste management scenario for both fuel cycles that would separate/recycle the waste's halide and immobilize everything else in iron phosphate (Fe-P) glass. It will describe both how the work was done and what its results indicate about how a treatment process for both of those wastes should be implemented (fluoride and chloride behave differently). In either case, this scenario's primary advantages include much higher waste loadings, much lower overall cost, and the generation of a product (glass) that is more consistent with current waste management practices. (author)

  20. Reprocessing of used tires into activated carbon and other products

    SciTech Connect

    Teng, H.; Serio, M.A.; Wojtowicz, M.A.; Bassilakis, R.; Solomon, P.R.

    1995-09-01

    Landfilling used tires which are generated each year in the US is increasingly becoming an unacceptable solution. A better approach, from an environmental and economic standpoint, is to thermally reprocess the tires into valuable products such as activated carbon, other solid carbon forms (carbon black, graphite, and carbon fibers), and liquid fuels. In this study, high surface area activated carbons (> 800 m{sup 2}/g solid product) were produced in relatively high yields by pyrolysis of tires at up to 900 C, followed by activation in CO{sub 2} at the same temperature. The surface areas of these materials are comparable with those of commercial activated carbons. The efficiency of the activation process (gain in specific surface area/loss in mass) was greatest (up to 138 m{sup 2}/g original tire) when large pieces of tire material were used ({approximately} 170 mg). Oxygen pretreatment of tires was found to enhance both the yield and the surface area of the carbon product. High-pressure treatment of tires at low temperatures (< 400 C) is an alternative approach if the recovery of carbon black or fuel oils is the primary objective.

  1. Repository disposal requirements for commercial transuranic wastes (generated without reprocessing)

    SciTech Connect

    Daling, P.M.; Ludwick, J.D.; Mellinger, G.B.; McKee, R.W.

    1986-06-01

    This report forms a preliminary planning basis for disposal of commercial transuranic (TRU) wastes in a geologic repository. Because of the unlikely prospects for commercial spent nuclear fuel reprocessing in the near-term, this report focuses on TRU wastes generated in a once-through nuclear fuel cycle. The four main objectives of this study were to: develop estimates of the current inventories, projected generation rates, and characteristics of commercial TRU wastes; develop proposed acceptance requirements for TRU wastes forms and waste canisters that ensure a safe and effective disposal system; develop certification procedures and processing requirements that ensure that TRU wastes delivered to a repository for disposal meet all applicable waste acceptance requirements; and identify alternative conceptual strategies for treatment and certification of commercial TRU first objective was accomplished through a survey of commercial producers of TRU wastes. The TRU waste acceptance and certification requirements that were developed were based on regulatory requirements, information in the literature, and from similar requirements already established for disposal of defense TRU wastes in the Waste Isolation Pilot Plant (WIPP) which were adapted, where necessary, to disposal of commercial TRU wastes. The results of the TRU waste-producer survey indicated that there were a relatively large number of producers of small quantities of TRU wastes.

  2. Kinetic and thermodynamic bases to resolve issues regarding conditioning of uranium metal fuels

    SciTech Connect

    Johnson, A.B.; Ballinger, R.G.; Simpson, K.A.

    1994-12-01

    Numerous uranium - bearing fuels are corroding in fuel storage pools in several countries. At facilities where reprocessing is no longer available, dry storage is being evaluated to preclude aqueous corrosion that is ongoing. It is essential that thermodynamic and kinetic factors are accounted for in transitions of corroding uranium-bearing fuels to dry storage. This paper addresses a process that has been proposed to move Hanford N-Reactor fuel from wet storage to dry storage.

  3. Steps of Reprocessing and Equipments

    PubMed Central

    Lee, Yong Kook

    2013-01-01

    With the increasing interest in endoscopy and the rising number of endoscopic examinations in hospitals, the importance of endoscopic reprocessing is also increasing. Cure facilities that are understaffed and ill-equipped are trying to cope with the problems of insufficient cleaning and high infection risks. To prevent endoscopy-associated infection, the endoscope cleaning, and disinfection guidelines prepared by the Korean Society of Gastrointestinal Endoscopy must be followed. In this review, the steps of endoscopic reprocessing and the equipments required in each step are discussed. PMID:23767039

  4. Process gases for high velocity oxy-fuel thermal spraying

    SciTech Connect

    Creffield, G.K.; Chapman, I.F.; Cole, M.A.; Page, W.J.; McDonough, T.

    1994-12-31

    The importance of fuel and other process gases for high velocity oxy-fuel (HVOF) thermal spraying and especially the delivery of them to the point of use is well recognized. Problems associated with the supply of liquid fuel gases, at the high flow rates and pressures required by this process, have been addressed. Considerable development work has gone into designing an appropriate liquid withdrawal and vaporizer system for propylene, which overcomes these previous difficulties and enables users to maintain adequate fuel gas flow rates in order to ensure optimum operating conditions for the production of high quality coatings. A feature of the thermal spray process is that the temperature of the workpiece is kept low, typically below 150 C, in order to reduce residual stresses in the coating and to protect heat sensitive substrates. Traditionally this has been by compressed air, however, improved cooling has been achieved using carbon dioxide. Specially designed equipment is now available which provides and directs a cold mixture of carbon dioxide gas and solid particles (snow) via suitable nozzles, on the workpiece. The position of the cooling stream can be varied, depending on the application. These developments emphasize the importance now attached to providing dedicated gas installation packages for HVOF.

  5. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    SciTech Connect

    Chad Pope

    2007-05-01

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day

  6. Process for removal of sulfur compounds from fuel gases

    DOEpatents

    Moore, Raymond H.; Stegen, Gary E.

    1978-01-01

    Fuel gases such as those produced in the gasification of coal are stripped of sulfur compounds and particulate matter by contact with molten metal salt. The fuel gas and salt are intimately mixed by passage through a venturi or other constriction in which the fuel gas entrains the molten salt as dispersed droplets to a gas-liquid separator. The separated molten salt is divided into a major and a minor flow portion with the minor flow portion passing on to a regenerator in which it is contacted with steam and carbon dioxide as strip gas to remove sulfur compounds. The strip gas is further processed to recover sulfur. The depleted, minor flow portion of salt is passed again into contact with the fuel gas for further sulfur removal from the gas. The sulfur depleted, fuel gas then flows through a solid absorbent for removal of salt droplets. The minor flow portion of the molten salt is then recombined with the major flow portion for feed to the venturi.

  7. Coal liquefaction process wherein jet fuel, diesel fuel and/or ASTM No. 2 fuel oil is recovered

    DOEpatents

    Bauman, Richard F.; Ryan, Daniel F.

    1982-01-01

    An improved process for the liquefaction of coal and similar solid carbonaceous materials wherein a hydrogen donor solvent or diluent derived from the solid carbonaceous material is used to form a slurry of the solid carbonaceous material and wherein the naphthenic components from the solvent or diluent fraction are separated and used as jet fuel components. The extraction increases the relative concentration of hydroaromatic (hydrogen donor) components and as a result reduces the gas yield during liquefaction and decreases hydrogen consumption during said liquefaction. The hydrogenation severity can be controlled to increase the yield of naphthenic components and hence the yield of jet fuel and in a preferred embodiment jet fuel yield is maximized while at the same time maintaining solvent balance.

  8. Fuel processing for PEM fuel cells: transport and kinetic issues of system design

    NASA Astrophysics Data System (ADS)

    Zalc, J. M.; Löffler, D. G.

    In light of the distribution and storage issues associated with hydrogen, efficient on-board fuel processing will be a significant factor in the implementation of PEM fuel cells for automotive applications. Here, we apply basic chemical engineering principles to gain insight into the factors that limit performance in each component of a fuel processor. A system consisting of a plate reactor steam reformer, water-gas shift unit, and preferential oxidation reactor is used as a case study. It is found that for a steam reformer based on catalyst-coated foils, mass transfer from the bulk gas to the catalyst surface is the limiting process. The water-gas shift reactor is expected to be the largest component of the fuel processor and is limited by intrinsic catalyst activity, while a successful preferential oxidation unit depends on strict temperature control in order to minimize parasitic hydrogen oxidation. This stepwise approach of sequentially eliminating rate-limiting processes can be used to identify possible means of performance enhancement in a broad range of applications.

  9. Analysis of on-board fuel processing designs for PEM fuel cell vehicles

    SciTech Connect

    Kartha, S.; Fischer, S.; Kreutz, T.

    1996-12-31

    As a liquid fuel with weight and volume energy densities comparable to those of gasoline, methanol is an attractive energy carrier for mobile power systems. It is available without contaminants such as sulfur, and can be easily reformed at relatively low temperatures with inexpensive catalysts. This study is concerned with comparing the net efficiencies of PEM fuel cell vehicles fueled with methanol and hydrogen, using fuel cell system models developed using ASPEN chemical process simulation software. For both the methanol and hydrogen systems, base case designs are developed and several variations are considered that differ with respect to the degree of system integration for recovery of heat and compressive work. The methanol systems are based on steam reforming with the water-gas shift reaction and preferential oxidation, and the hydrogen systems are based on compressed hydrogen. This analysis is an exercise in optimizing the system design for each fuel, which ultimately entails balancing system efficiency against a host of other considerations, including system complexity, performance, cost, reliability, weight and volume.

  10. INTERIOR PHOTO OF MAIN PROCESSING BUILDING (CPP601) PROCESS MAKEUP AREA ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    INTERIOR PHOTO OF MAIN PROCESSING BUILDING (CPP-601) PROCESS MAKEUP AREA LOOKING SOUTH. PHOTO TAKEN FROM CENTER OF WEST WALL. INL PHOTO NUMBER HD-50-1-4. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  11. FLOOR PLAN OF MAIN PROCESSING BUILDING (CPP601) BASEMENT SHOWING PROCESS ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    FLOOR PLAN OF MAIN PROCESSING BUILDING (CPP-601) BASEMENT SHOWING PROCESS CORRIDOR AND EIGHTEEN CELLS. TO LEFT IS LABORATORY BUILDING (CPP-602). INL DRAWING NUMBER 200-0601-00-706-051981. ALTERNATE ID NUMBER CPP-E-1981. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  12. INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP601) ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP-601) LOOKING NORTHWEST. PHOTO TAKEN FROM MIDDLE OF CORRIDOR. INL PHOTO NUMBER HD-50-2-3. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  13. INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP601) ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP-601) LOOKING NORTH. PHOTO TAKEN FROM SOUTHWEST CORNER. INL PHOTO NUMBER HD-50-1-3. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  14. INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP601) ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP-601) LOOKING SOUTH. PHOTO TAKEN FROM MIDDLE OF CORRIDOR. INL PHOTO NUMBER HD-50-3-2. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  15. INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP601) ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP-601) LOOKING SOUTHWEST. PHOTO TAKEN FROM NORTHEAST CORNER. INL PHOTO NUMBER HD-50-4-2. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

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

    SciTech Connect

    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.

  17. FUEL ELEMENTS FOR NUCLEAR REACTORS AND PROCESS OF MAKING

    DOEpatents

    Roake, W.E.

    1958-08-19

    A process is described for producing uranium metal granules for use in reactor fuel elements. The granules are made by suspending powdered uramiunn metal or uranium hydride in a viscous, non-reactive liquid, such as paraffin oil, aad pouring the resulting suspension in droplet, on to a bed of powdered absorbent. In this manner the liquid vehicle is taken up by the sorbent and spherical pellets of uranium metal are obtained. The

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

    SciTech Connect

    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.

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

    SciTech Connect

    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

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

    NASA Astrophysics Data System (ADS)

    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

  1. Process for producing fuel from plant sources and fuel blends containing same

    SciTech Connect

    Whitworth, R.D.

    1987-04-04

    A process for producing a fuel from plant sources comprising the steps of: (a) providing a supply of limonene; (b) distilling the limonene and removing the distillate fraction in a temperature range of from about 346/sup 0/F to about 382/sup 0/F; (c) removing water from the distilled limonene; and (d) treating the distilled limonene to preclude the formation of gums therefrom.

  2. Process for converting cellulosic materials into fuels and chemicals

    DOEpatents

    Scott, Charles D.; Faison, Brendlyn D.; Davison, Brian H.; Woodward, Jonathan

    1994-01-01

    A process for converting cellulosic materials, such as waste paper, into fuels and chemicals utilizing enzymatic hydrolysis of the major constituent of paper, cellulose. A waste paper slurry is contacted by cellulase in an agitated hydrolyzer. The cellulase is produced from a continuous, columnar, fluidized-bed bioreactor utilizing immobilized microorganisms. An attritor and a cellobiase reactor are coupled to the agitated hydrolyzer to improve reaction efficiency. The cellulase is recycled by an adsorption process. The resulting crude sugars are converted to dilute product in a fluidized-bed bioreactor utilizing microorganisms. The dilute product is concentrated and purified by utilizing distillation and/or a biparticle fluidized-bed bioreactor system.

  3. Separation of Plutonium from Irradiated Fuels and Targets

    SciTech Connect

    Gray, Leonard W.; Holliday, Kiel S.; Murray, Alice; Thompson, Major; Thorp, Donald T.; Yarbro, Stephen; Venetz, Theodore J.

    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.

  4. Reprocessing system with nuclide separation based on chromatography in hydrochloric acid solution

    SciTech Connect

    Suzuki, Tatsuya; Tachibana, Yu; Koyama, Shi-ichi

    2013-07-01

    We have proposed the reprocessing system with nuclide separation processes based on the chromatographic technique in the hydrochloric acid solution system. Our proposed system consists of the dissolution process, the reprocessing process, the minor actinide separation process, and nuclide separation processes. In the reprocessing and separation processes, the pyridine resin is used as a main separation media. It was confirmed that the dissolution in the hydrochloric acid solution is easily achieved by the plasma voloxidation and by the addition of oxygen peroxide into the hydrochloric acid solution.

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

    SciTech Connect

    Brown, L.F.

    1996-03-01

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

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

    SciTech Connect

    Hertzler, T. )

    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.

  7. Preliminary analysis of treatment strategies for transuranic wastes from reprocessing plants

    SciTech Connect

    Ross, W.A.; Schneider, K.J.; Swanson, J.L.; Yasutake, K.M.; Allen, R.P.

    1985-07-01

    This document provides a comparison of six treatment options for transuranic wastes (TRUW) resulting from the reprocessing of commercial spent fuel. Projected transuranic waste streams from the Barnwell Nuclear Fuel Plant (BNFP), the reference fuel reprocessing plant in this report, were grouped into the five categories of hulls and hardware, failed equipment, filters, fluorinator solids, and general process trash (GPT) and sample and analytical cell (SAC) wastes. Six potential treatment options were selected for the five categories of waste. These options represent six basic treatment objectives: (1) no treatment, (2) minimum treatment (compaction), (3) minimum number of processes and products (cementing or grouting), (4) maximum volume reduction without decontamination (melting, incinerating, hot pressing), (5) maximum volume reduction with decontamination (decontamination, treatment of residues), and (6) noncombustible waste forms (melting, incinerating, cementing). Schemes for treatment of each waste type were selected and developed for each treatment option and each type of waste. From these schemes, transuranic waste volumes were found to vary from 1 m/sup 3//MTU for no treatment to as low as 0.02 m/sup 3//MTU. Based on conceptual design requirements, life-cycle costs were estimated for treatment plus on-site storage, transportation, and disposal of both high-level and transuranic wastes (and incremental low-level wastes) from 70,000 MTU. The study concludes that extensive treatment is warranted from both cost and waste form characteristics considerations, and that the characteristics of most of the processing systems used are acceptable. The study recommends that additional combinations of treatment methods or strategies be evaluated and that in the interim, melting, incineration, and cementing be further developed for commercial TRUW. 45 refs., 9 figs., 32 tabs.

  8. Plate-Based Fuel Processing System Final Report

    SciTech Connect

    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

  9. Fuel quality/processing study. Volume 2: Appendix. Task 1 literature survey

    NASA Technical Reports Server (NTRS)

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

    1981-01-01

    The results of a literature survey of fuel processing and fuel quality are given. Liquid synfuels produced from coal and oil shale are discussed. Gas turbine fuel property specifications are discussed. On-site fuel pretreatment and emissions from stationary gas turbines are discussed. Numerous data tables and abstracts are given.

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

    PubMed

    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. PMID:24997379

  11. Microbial fuel cell treatment of ethanol fermentation process water

    SciTech Connect

    Borole, Abhijeet P.

    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.

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

    SciTech Connect

    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.

  13. Closed Fuel Cycle Waste Treatment Strategy

    SciTech Connect

    Vienna, J. D.; Collins, E. D.; Crum, J. V.; Ebert, W. L.; Frank, S. M.; Garn, T. G.; Gombert, D.; Jones, R.; Jubin, R. T.; Maio, V. C.; Marra, J. C.; Matyas, J.; Nenoff, T. M.; Riley, B. J.; Sevigny, G. J.; Soelberg, N. R.; Strachan, D. M.; Thallapally, P. K.; Westsik, J. H.

    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.

  14. PROCESS OF MAKING SHAPED FUEL FOR NUCLEAR REACTORS

    DOEpatents

    O'Leary, W.J.; Fisher, E.A.

    1964-02-11

    A process for making uranium dioxide fuel of great strength, density, and thermal conductivity by mixing it with 0.1 to 1% of a densifier oxide (tin, aluminum, zirconium, ferric, zinc, chromium, molybdenum, titanium, or niobium oxide) and with a plasticizer (0.5 to 3% of bentonite and 0.05 to 2% of methylcellulose, propylene glycol alginate, or ammonium alginate), compacting the mixture obtained, and sintering the bodies in an atmosphere of carbon monoxide or carbon dioxide, with or without hydrogen, or of a nitrogen-hydrogen mixture is described. (AEC)

  15. PROCESS OF MAKING A NEUTRONIC REACTOR FUEL ELEMENT COMPOSITION

    DOEpatents

    Alter, H.W.; Davidson, J.K.; Miller, R.S.; Mewherter, J.L.

    1959-01-13

    A process is presented for making a ceramic-like material suitable for use as a nuclear fuel. The material consists of a solid solution of plutonium dioxide in uranium dioxide and is produced from a uranyl nitrate -plutonium nitrate solution containing uraniunm and plutonium in the desired ratio. The uranium and plutonium are first precipitated from the solution by addition of NH/ sub 4/OH and the dried precipitate is then calcined at 600 C in a hydrogen atmosphere to yield the desired solid solution of PuO/sub 2/ in UO/sub 2/.

  16. Automated catalyst processing for cloud electrode fabrication for fuel cells

    DOEpatents

    Goller, Glen J.; Breault, Richard D.

    1980-01-01

    A process for making dry carbon/polytetrafluoroethylene floc material, particularly useful in the manufacture of fuel cell electrodes, comprises of the steps of floccing a co-suspension of carbon particles and polytetrafluoroethylene particles, filtering excess liquids from the co-suspension, molding pellet shapes from the remaining wet floc solids without using significant pressure during the molding, drying the wet floc pellet shapes within the mold at temperatures no greater than about 150.degree. F., and removing the dry pellets from the mold.

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

    SciTech Connect

    Not Available

    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.

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

    SciTech Connect

    Herrmann, F.J.; Herrmann, B.; Hoeflich, V.

    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.

  19. DIRECT DISMANTLING OF REPROCESSING PLANT CELLS THE EUREX PLANT EXPERIENCEe2d12c

    SciTech Connect

    Gili, M.; Troiani, F.; Risoluti, P.

    2003-02-27

    After finishing the reprocessing campaigns in 1970-1983, the EUREX pilot reprocessing plant of ENEA Saluggia Research Center started into a new phase, aiming to materials and irradiated fuel systemation and radioactive wastes conditioning. In 1997 the project ''CORA'' for a vitrification plant for the high and intermediate liquid radioactive wastes started. The ''CORA'' plant will be hosted in some dismantled cells of the EUREX plant, reusing many of the EUREX plant auxiliary systems, duly refurbished, saving money and construction time and avoiding a new nuclear building in the site. Two of the cells that will be reused were part of the EUREX chemical process (solvent recovery and 2nd extraction cycle) and the components were obviously internally contaminated. In 2000 the direct (hands-on) dismantling of one of them started and has been completed in summer 2002; the second one will be dismantled in the next year and then the ''CORA'' plant will be assembled inside the cells. Special care w as taken to avoid spread of contamination in the cells, where ''CORA'' installation activities will start in the next years, during the dismantling process The analysis of data and results collected during the dismantling of the first cell shows that direct dismantling can be achieved with careful choice of tools, procedures and techniques, to reduce volumes of wastes to be disposed and radiological burden.

  20. NEW MATERIAL NEEDS FOR HYDROCARBON FUEL PROCESSING: Generating Hydrogen for the PEM Fuel Cell

    NASA Astrophysics Data System (ADS)

    Farrauto, R.; Hwang, S.; Shore, L.; Ruettinger, W.; Lampert, J.; Giroux, T.; Liu, Y.; Ilinich, O.

    2003-08-01

    The hydrogen economy is fast approaching as petroleum reserves are rapidly consumed. The fuel cell promises to deliver clean and efficient power by combining hydrogen and oxygen in a simple electrochemical device that directly converts chemical energy to electrical energy. Hydrogen, the most plentiful element available, can be extracted from water by electrolysis. One can imagine capturing energy from the sun and wind and/or from the depths of the earth to provide the necessary power for electrolysis. Alternative energy sources such as these are the promise for the future, but for now they are not feasible for power needs across the globe. A transitional solution is required to convert certain hydrocarbon fuels to hydrogen. These fuels must be available through existing infrastructures such as the natural gas pipeline. The present review discusses the catalyst and adsorbent technologies under development for the extraction of hydrogen from natural gas to meet the requirements for the proton exchange membrane (PEM) fuel cell. The primary market is for residential applications, where pipeline natural gas will be the source of H2 used to power the home. Other applications including the reforming of methanol for portable power applications such as laptop computers, cellular phones, and personnel digital equipment are also discussed. Processing natural gas containing sulfur requires many materials, for example, adsorbents for desulfurization, and heterogeneous catalysts for reforming (either autothermal or steam reforming) water gas shift, preferential oxidation of CO, and anode tail gas combustion. All these technologies are discussed for natural gas and to a limited extent for reforming methanol.

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

    PubMed

    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. PMID:18781691

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

    NASA Astrophysics Data System (ADS)

    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.

  3. Deposition behavior of UO2 and noble-metal elements in oxide-electrowinning reprocessing

    NASA Astrophysics Data System (ADS)

    Kosugi, K.; Fukushima, M.; Myochin, M.; Mizuguchi, K.; Oomori, T.

    2005-02-01

    As a candidate process for future reprocessing technology of nuclear spent fuel, oxide-electrowinning method has been studied. In this method, the uranium is collected on the cathode in the form of UO2 by electrolysis in the molten chloride. Thereby, the noble metal (NM) elements accompany the uranium deposition, because of very close redox potential between NM elements and UO2. To clarify the electrolysis behavior of the uranium and NM elements in the low-current-density electrolysis, the laboratory scale experiments were performed under various conditions of cathode current density and solutes concentration in the chloride melt, and the separation efficiency and the morphology of the deposition were investigated. It was found that the separation of Pd from uranium was more difficult than that of Rh. The presence of U4+ greatly influenced current efficiency of the electrolysis process.

  4. Synthesis of Diopside by Solution Combustion Process Using Glycine Fuel

    NASA Astrophysics Data System (ADS)

    Sherikar, Baburao N.; Umarji, A. M.

    Nano ceramic Diopside (CaMgSi2O6) powders are synthesized by Solution Combustion Process(SCS) using Calcium nitrate, Magnesium nitrate as oxidizer and glycine as fuel, fumed silica as silica source. Ammonium nitrate (AN) is used as extra oxidizer. Effect of AN on Diopside phase formation is investigated. The adiabatic flame temperatures are calculated theoretically for varying amount of AN according to thermodynamic concept and correlated with the observed flame temperatures. A “Multi channel thermocouple setup connected to computer interfaced Keithley multi voltmeter 2700” is used to monitor the thermal events during the process. An interpretation based on maximum combustion temperature and the amount of gases produced during reaction for various AN compositions has been proposed for the nature of combustion and its correlation with the characteristics of as synthesized powder. These powders are characterized by XRD, SEM showing that the powders are composed of polycrystalline oxides with crystallite size of 58nm to 74nm.

  5. Process for converting cellulosic materials into fuels and chemicals

    DOEpatents

    Scott, C.D.; Faison, B.D.; Davison, B.H.; Woodward, J.

    1994-09-20

    A process is described for converting cellulosic materials, such as waste paper, into fuels and chemicals utilizing enzymatic hydrolysis of the major constituent of paper, cellulose. A waste paper slurry is contacted by cellulase in an agitated hydrolyzer. The cellulase is produced from a continuous, columnar, fluidized-bed bioreactor utilizing immobilized microorganisms. An attrition mill and a cellobiase reactor are coupled to the agitated hydrolyzer to improve reaction efficiency. The cellulase is recycled by an adsorption process. The resulting crude sugars are converted to dilute product in a fluidized-bed bioreactor utilizing microorganisms. The dilute product is concentrated and purified by utilizing distillation and/or a biparticle fluidized-bed bioreactor system. 1 fig.

  6. Commercial Nuclear Reprocessing in the United States

    SciTech Connect

    Sherrill, Charles Leland; Balatsky, Galya Ivanovna

    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.

  7. Safeguards and Non-proliferation Issues as Related to Advanced Fuel Cycle and Advanced Fast Reactor Development with Processing of Reactor Fuel

    SciTech Connect

    Rahmat Aryaeinejad; Jerry D. Cole; Mark W. Drigert; Dee E. Vaden

    2006-10-01

    The goal of this work is to establish basic data and techniques to enable safeguards appropriate to a new generation of nuclear power systems that will be based on fast spectrum reactors and mixed actinide fuels containing significant quantities of "minor" actinides, possibly due to reprocessing, and determination of what new radiation signatures and parameters need to be considered. The research effort focuses on several problems associated with the use of fuel having significantly different actinide inventories that current practice and on the development of innovative techniques using new radiation signatures and other parameters useful for safeguards and monitoring. In addition, the development of new distinctive radiation signatures as an aid in controlling proliferation of nuclear materials has parallel applications to support Gen-IV and current advanced fuel cycle initiative (AFCI) goals as well as the anticipated Global Nuclear Energy Partnership (GNEP).

  8. Endoscope Reprocessing: Update on Controversial Issues

    PubMed Central

    Choi, Hyun Ho

    2015-01-01

    Several issues concerning endoscope reprocessing remain unresolved based on currently available data. Thus, further studies are required to confirm standard practices including safe endoscope shelf life, proper frequency of replacement of some accessories including water bottles and connecting tubes, and microbiological surveillance testing of endoscopes after reprocessing. The efficacy and cost-effectiveness of newer technology that allows automated cleaning and disinfection is one such controversial issue. In addition, there are no guidelines on whether delayed reprocessing and extended soaking may harm endoscope integrity or increase the bioburden on the external or internal device surfaces. In this review, we discuss the unresolved and controversial issues regarding endoscope reprocessing. PMID:26473115

  9. Technetium-99 and strontium-90: Abundance determination at ultratrace sensitivity by AMS as signatures of undeclared nuclear reprocessing activity

    SciTech Connect

    McAninch, J.E.; Proctor, I.D.

    1995-03-01

    The purpose of this White Paper is to examine the use of the ultratrace technique Accelerator Mass Spectrometry (AMS) to lower detection limits for {sup 99}Tc and {sup 90}Sr, and to examine the utility of these isotopes as signatures of a convert reprocessing facility. The International Atomic Energy Agency (IAEA) has committed to improving the effectiveness of the IAEA Safeguards System. This is in some degree a result of the discovery in 1991 of an undeclared Iraqi EMIS program. Recommendations from the March 1993 Consultants Group Meeting have resulted in several studies and follow on field trials to identify environmental signatures from covert nuclear fuel reprocessing activity. In particular, the April, 1993 reports of the Standing Advisory Group on Safeguards Implementation (SAGSI) identified the long-lived radioisotopes Technetium-99 and strontium-90 as two reliable signatures of fuel reprocessing activity. This report also suggested pathways in the chemical processing of irradiated fuel where these elements would be volatilized and potentially released in amounts detectable with ultratrace sensitivity techniques. Based on measured {sup 99}Tc background levels compiled from a variety of sources, it is estimated that AMS can provide 10% measurements of environmental levels of {sup 99}Tc in a few minutes using modestly sized samples: a few grams for soils, plants, or animal tissues; one to several liters for rain or seawater samples; and tens to hundreds of cubic meters for air sampling. Small sample sizes and high sample throughput result in significant increases in feasibility, cost effectiveness, and quality of data for a regional monitoring program. Similar results are expected for {sup 90}Sr.

  10. Review of solar fuel-producing quantum conversion processes

    NASA Astrophysics Data System (ADS)

    Peterson, D. B.; Biddle, J. R.; Fujita, T.

    1984-05-01

    The status and potential of fuel-producing solar photochemical processes are discussed. Research focused on splitting water to produce dihydrogen and is at a relatively early stage of development. Current emphasis is primarily directed toward understanding the basic chemistry underlying such quantum conversion processes. Theoretical analyses by various investigators predict a limiting thermodynamic efficiency of 31% for devices with a single photosystem operating with unfocused sunlight at 300 K. When non-idealities are included, it appears unlikely that actual devices will have efficiencies greater than 12 to 15%. Observed efficiencies are well below theoretical limits. Cyclic homogeneous photochemical processes for splitting water have efficiencies considerably less than 1%. Efficiency can be significantly increased by addition of a sacrificial reagent; however, such systems are no longer cyclic and it is doubtful that they would be economical on a commercial scale. The observed efficiencies for photoelectrochemical processes are also low but such systems appear more promising than homogeneous photochemical systems. Operating and systems options, including operation at elevated temperature and hybrid and coupled quantum-thermal conversion processes, are also considered.

  11. Review of solar fuel-producing quantum conversion processes

    NASA Technical Reports Server (NTRS)

    Peterson, D. B.; Biddle, J. R.; Fujita, T.

    1984-01-01

    The status and potential of fuel-producing solar photochemical processes are discussed. Research focused on splitting water to produce dihydrogen and is at a relatively early stage of development. Current emphasis is primarily directed toward understanding the basic chemistry underlying such quantum conversion processes. Theoretical analyses by various investigators predict a limiting thermodynamic efficiency of 31% for devices with a single photosystem operating with unfocused sunlight at 300 K. When non-idealities are included, it appears unlikely that actual devices will have efficiencies greater than 12 to 15%. Observed efficiencies are well below theoretical limits. Cyclic homogeneous photochemical processes for splitting water have efficiencies considerably less than 1%. Efficiency can be significantly increased by addition of a sacrificial reagent; however, such systems are no longer cyclic and it is doubtful that they would be economical on a commercial scale. The observed efficiencies for photoelectrochemical processes are also low but such systems appear more promising than homogeneous photochemical systems. Operating and systems options, including operation at elevated temperature and hybrid and coupled quantum-thermal conversion processes, are also considered.

  12. Reprocessing of lithium titanate pebbles by graphite bed method

    NASA Astrophysics Data System (ADS)

    Hong, Ming; Zhang, Yingchun; Xiang, Maoqiao; Zhang, Yun

    2015-04-01

    Lithium titanate enriched by 6Li isotope is considered as a candidate of tritium breeding materials for fusion reactors due to its excellent performance. The reuse of burned Li2TiO3 pebbles is an important issue because of the high costs of 6Li-enriched materials and waste considerations. For this purpose, reprocessing of Li2TiO3 pebbles by graphite bed method was developed. Simulative Li2TiO3 pebbles with low-lithium content according to the expected lithium burn-up were fabricated. After that, Li2TiO3 pebbles were re-fabricated with lithium carbonate as lithium additives, in order to gain the composition of lithium titanate with a Li/Ti ratio of 2. The process was optimized to obtain reprocessed Li2TiO3 pebbles that were suitable for reuse as ceramic breeder. Density, porosity, grain size and crushing load of the reprocessed pebbles were characterized. This process did not deteriorate the properties of the reprocessed pebbles and was almost no waste generation.

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

    NASA Astrophysics Data System (ADS)

    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.

  14. Environmental assessment for radioisotope heat source fuel processing and fabrication

    SciTech Connect

    Not Available

    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.

  15. Sulfur tolerant molten carbonate fuel cell anode and process

    DOEpatents

    Remick, Robert J.

    1990-01-01

    Molten carbonate fuel cell anodes incorporating a sulfur tolerant carbon monoxide to hydrogen water-gas-shift catalyst provide in situ conversion of carbon monoxide to hydrogen for improved fuel cell operation using fuel gas mixtures of over about 10 volume percent carbon monoxide and up to about 10 ppm hydrogen sulfide.

  16. Thermodynamic analysis and comparison on oxy-fuel power generation process - article no. 053001

    SciTech Connect

    Deng, S.M.; Hynes, R.

    2009-09-15

    In this paper, pressurized oxy-fuel combustion power generation processes are modeled and analyzed based on a 350 MW subcritical reheat boiler associated with a condensing steam turbine. The performance results are obtained. Furthermore, the influences of slurry concentration and coal properties on power plant performance are investigated. An oxy-fuel configuration operating at ambient pressure is studied to compare the performance with pressurized oxy-fuel configuration. Thermodynamic analysis reveals the true potentials of the pressurized oxy-fuel process. Based on the system integration, an improved configuration is proposed in which plant efficiency of pressurized oxy-fuel process is increased by 1.36%.

  17. CONSTRUCTION PROGRESS PHOTO SHOWING EXCAVATION PIT FOR MAIN PROCESSING BUILDING ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    CONSTRUCTION PROGRESS PHOTO SHOWING EXCAVATION PIT FOR MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTHWEST. INL PHOTO NUMBER NRTS-50-885. Unknown Photographer, 10/30/1950 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  18. AERIAL VIEW OF MAIN PROCESSING BUILDING SHOWING CONSTRUCTION PROGRESS AND ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    AERIAL VIEW OF MAIN PROCESSING BUILDING SHOWING CONSTRUCTION PROGRESS AND EXCAVATION FOR LABORATORY ON LEFT. INL PHOTO NUMBER NRTS-51-1759. Unknown Photographer, 3/28/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  19. PLAN SECTIONS AND DETAILS OF CELL HATCHES MAIN PROCESSING BUILDING ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PLAN SECTIONS AND DETAILS OF CELL HATCHES MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103256. ALTERNATE ID NUMBER 542-11-F-302. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  20. ARCHITECTURAL FLOOR PLAN OF PROCESS AND ACCESS AREAS HOT PILOT ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    ARCHITECTURAL FLOOR PLAN OF PROCESS AND ACCESS AREAS HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-0640-00-279-111679. ALTERNATE ID NUMBER 8952-CPP-640-A-2. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  1. The chemical state of fission products in oxide fuels at different stages of the nuclear fuel cycle

    SciTech Connect

    Kleykamp, H.

    1988-03-01

    A survey of work at the Kernforschungszentrum Karlsruhe is presented on the chemical state of selected fission products that are relevant in the fuel cycle of light water reactor (LWR) and fast breeder reactor fuels. The influence of fuel type and irradiation progress on the composition of the Mo-Tc-Ru-Rh-Pd fission product alloys precipitated in the oxide matrix is examined using the respective multicomponent phase diagrams. The kinetics of dissolution of these phases in nitric acid at the reprocessing stage is discussed. Composition and structure of the residues, and the reprecipitation phenomena from highly active waste (HAW), are elucidated. A second metamorphosis of the fission products is recognized during the vitrification process. The formation of Ru(Rh) oxide and Pd(Rh, U, Te) alloys in simulated vitrified HAW concentrate and in HAW concentrate from the reprocessing of irradiated LWR fuels in interpreted on the basis of heterogeneous equilibria.

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

    NASA Astrophysics Data System (ADS)

    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

  3. Survey of Endoscope Reprocessing in Korea

    PubMed Central

    Park, Jeong Bae; Yang, Jae Nam; Koo, Ja Seol; Jang, Jae Young; Park, Sang Hoon; Hong, Su Jin; Kim, Sang-Woo; Chun, Hoon Jai

    2015-01-01

    Background/Aims There is a growing emphasis on quality management in endoscope reprocessing. Previous surveys conducted in 2002 and 2004 were not practitioner-oriented. Therefore, this survey is significant for being the first to target actual participants in endoscope reprocessing in Korea. Methods This survey comprised 33 self-filled questions, and was personally delivered to nurses and nursing auxiliaries in the endoscopy departments of eight hospitals belonging to the society. The anonymous responses were collected after 1 week either by post or in person by committee members. Results The survey included 100 participants. In the questionnaire addressing compliance rates with the reprocessing guideline, the majority (98.9%) had a high compliance rate compared to 27% of respondents in 2002 and 50% in 2004. The lowest rate of compliance with a reprocessing procedure was reported for transporting the contaminated endoscope in a sealed container. Automated endoscope reprocessors were available in all hospitals. Regarding reprocessing time, more than half of the subjects replied that reprocessing took more than 15 minutes (63.2%). Conclusions The quality management of endoscope reprocessing has improved as since the previous survey. A national survey expanded to include primary clinics is required to determine the true current status of endoscope reprocessing. PMID:25674525

  4. Assessment of sulfur removal processes for advanced fuel cell systems

    SciTech Connect

    Lorton, G.A.

    1980-01-01

    This study consisted of a technical evaluation and economic comparison of sulfur removal processes for integration into a coal gasification-molten carbonate (CGMC) fuel cell power plant. Initially, the performance characteristics of potential sulfur removal processes were evaluated and screened for conformance to the conditions and requirements expected in commercial CGMC power plants. Four of these processes, the Selexol process, the Benfield process, the Sulfinol process, and the Rectisol process, were selected for detailed technical and economic comparison. The process designs were based on a consistent set of technical criteria for a grass roots facility with a capacity of 10,000 tons per day of Illinois No. 6 coal. Two raw gas compositions, based on oxygen-blown and air-blown Texaco gasification, were used. The bulk of the sulfur was removed in the sulfur removal unit, leaving a small amount of sulfur compounds in the gas (1 ppMv or 25 ppMv). The remaining sulfur compounds were removed by reaction with zinc oxide in the sulfur polishing unit. The impact of COS hydrolysis pretreatment on sulfur removal was evaluated. Comprehensive capital and O and M cost estimates for each of the process schemes were developed for the essentially complete removal of sulfur compounds. The impact on the overall plant performance was also determined. The total capital requirement for sulfur removal schemes ranged from $59.4/kW to $84.8/kW for the oxygen-blown cases and from $89.5/kW to $133/kW for the air-blown cases. The O and M costs for sulfur removal for 70% plant capacity factor ranged from 0.82 mills/kWh to 2.76 mills/kWh for the oxygen-blown cases and from 1.77 mills/kWh to 4.88 mills/kWh for the air-blown cases. The Selexol process benefitted the most from the addition of COS hydrolysis pretreatment.

  5. Environmental Impacts on Nuclear Reprocessing Solvents

    NASA Astrophysics Data System (ADS)

    Gillens, A. R.; Fessenden, J. E.

    2009-12-01

    Nuclear tests have been employed ever since the first nuclear explosion in Alamogordo, NM during the mid-1940s. Nuclear weapons pose a threat to civil society and result in extensive biological (medical) damages. For this reason, treaties banning nuclear tests and weapons have been employed since the 1960s to cease proliferation of weapons. However, as nuclear tests continue in secrecy and actinides, such as plutonium and uranium, are eligible for theft, nuclear forensics is needed to prevent weapons proliferation. In this study, solvents [tributyl phosphate (TBP), dodecane, decanol] used in reprocessing spent nuclear fuel are analyzed using an isotope ratio mass spectrometer, which provides indisputable evidence in identifying the operation in which solvents were used. Solvent samples are observed under variable conditions in the laboratory for different time periods. It is assumed that their carbon isotope values (δ13C) will become more positive (shift heavy) with time. It is found that the solvents are hygroscopic. TBP leaves the most robust signature compared to the other solvents studied and the isotope values for all solvents under all conditions become more positive with time. This study serves as primary research in understanding how solvents behave under variable conditions in the laboratory and how this could be translated to the environment in fate and transport studies.

  6. X-ray reprocessing in binaries

    NASA Astrophysics Data System (ADS)

    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.

  7. FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION

    SciTech Connect

    Carter, J.

    2011-01-03

    The United States (U.S.) currently utilizes a once-through fuel cycle where used nuclear fuel (UNF) is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. Within the Department of Energy's (DOE) Office of Nuclear Energy (DOE-NE), the Fuel Cycle Research and Development Program (FCR&D) develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development of advanced fuel cycles, including modified open and closed cycles. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Yet, the routine disposal of used nuclear fuel and radioactive waste remains problematic. Advanced fuel cycles will generate different quantities and forms of waste than the current LWR fleet. This study analyzes the quantities and characteristics of potential waste forms including differing waste matrices, as a function of a variety of potential fuel cycle alternatives including: (1) Commercial UNF generated by uranium fuel light water reactors (LWR). Four once through fuel cycles analyzed in this study differ by varying the assumed expansion/contraction of nuclear power in the U.S. (2) Four alternative LWR used fuel recycling processes analyzed differ in the reprocessing method (aqueous vs. electro-chemical), complexity (Pu only or full transuranic (TRU) recovery) and waste forms generated. (3) Used Mixed Oxide (MOX) fuel derived from the recovered Pu utilizing a single reactor pass. (4) Potential waste forms generated by the reprocessing of fuels derived from recovered TRU utilizing multiple reactor passes.

  8. Process Developed for Fabricating Engineered Pore Structures for High- Fuel-Utilization Solid Oxide Fuel Cells

    NASA Technical Reports Server (NTRS)

    Sofie, Stephen W.; Cable, Thomas L.; Salamone, Sam M.

    2005-01-01

    Solid oxide fuel cells (SOFCs) have tremendous commercial potential because of their high efficiency, high energy density, and flexible fuel capability (ability to use fossil fuels). The drive for high-power-utilizing, ultrathin electrolytes (less than 10 microns), has placed an increased demand on the anode to provide structural support, yet allow sufficient fuel entry for sustained power generation. Concentration polarization, a condition where the fuel demand exceeds the supply, is evident in all commercial-based anode-supported cells, and it presents a significant roadblock to SOFC commercialization.

  9. Fate of virginiamycin through the fuel ethanol production process.

    PubMed

    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. PMID:27038946

  10. Process for making a martensitic steel alloy fuel cladding product

    DOEpatents

    Johnson, Gerald D.; Lobsinger, Ralph J.; Hamilton, Margaret L.; Gelles, David S.

    1990-01-01

    This is a very narrowly defined martensitic steel alloy fuel cladding material for liquid metal cooled reactors, and a process for making such a martensitic steel alloy material. The alloy contains about 10.6 wt. % chromium, about 1.5 wt. % molybdenum, about 0.85 wt. % manganese, about 0.2 wt. % niobium, about 0.37 wt. % silicon, about 0.2 wt. % carbon, about 0.2 wt. % vanadium, 0.05 maximum wt. % nickel, about 0.015 wt. % nitrogen, about 0.015 wt. % sulfur, about 0.05 wt. % copper, about 0.007 wt. % boron, about 0.007 wt. % phosphorous, and with the remainder being essentially iron. The process utilizes preparing such an alloy and homogenizing said alloy at about 1000.degree. C. for 16 hours; annealing said homogenized alloy at 1150.degree. C. for 15 minutes; and tempering said annealed alloy at 700.degree. C. for 2 hours. The material exhibits good high temperature strength (especially long stress rupture life) at elevated temperature (500.degree.-760.degree. C.).

  11. Disposition of salt-waste from pyrochemical nuclear fuel processing

    SciTech Connect

    Vance, E.R.

    2007-07-01

    Waste salts from pyrochemical processing of nuclear fuel can be immobilised in sodalite if consolidated by hot isostatic pressing (HIP) at {approx}750 deg. C/100 MPa in thick stainless steel 316 cans. Other canning materials for this purpose also look possible. Spodiosite-based waste forms do not look promising in terms of leach resistance and their incorporation of alkali ions and compatibility with other phases which could potentially accommodate fission products, such as NaZr{sub 2}(PO{sub 4}){sub 3} or alumino-phosphate glass. Chloro- or fluor-apatite-based waste forms however have been reported to successfully accommodate fission products and alkalis which would be derived from either chloride- or fluoride-based waste pyro-processing salts. The presence of 10 or 20 wt% of additional Whitlockite, Ca{sub 3}(PO{sub 4}){sub 2}, should allow chemical flexibility to maintain the same qualitative phase assemblage when there are variations in the waste feed and in the waste/precursor ratios. Experimental verification of incorporation of the full complement of waste salts and fission products is not yet complete however. Apatite-rich samples could likely be HIPed in Inconel 600 cans. Other candidate HIP canning materials such as Alloy 22 or Inconel 625 are under study by encapsulating them in the candidate waste form and studying their interaction or otherwise with the waste form. (author)

  12. Potential synergy: the thorium fuel cycle and rare earths processing

    SciTech Connect

    Ault, T.; Wymer, R.; Croff, A.; Krahn, S.

    2013-07-01

    The use of thorium in nuclear power programs has been evaluated on a recurring basis. A concern often raised is the lack of 'thorium infrastructure'; however, for at least a part of a potential thorium fuel cycle, this may less of a problem than previously thought. Thorium is frequently encountered in association with rare earth elements and, since the U.S. last systematically evaluated the large-scale use of thorium (the 1970's,) the use of rare earth elements has increased ten-fold to approximately 200,000 metric tons per year. Integration of thorium extraction with rare earth processing has been previously described and top-level estimates have been done on thorium resource availability; however, since ores and mining operations differ markedly, what is needed is process flowsheet analysis to determine whether a specific mining operation can feasibly produce thorium as a by-product. Also, the collocation of thorium with rare earths means that, even if a thorium product stream is not developed, its presence in mining waste streams needs to be addressed and there are previous instances where this has caused issues. This study analyzes several operational mines, estimates the mines' ability to produce a thorium by-product stream, and discusses some waste management implications of recovering thorium. (authors)

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

    NASA Astrophysics Data System (ADS)

    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

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

    SciTech Connect

    E.T. 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.

  15. Advanced reprocessing developments in Europe contribution of European projects ACSEPT and ACTINET-I3

    SciTech Connect

    Bourg, S.; Poinssot, C.; Geist, A.; Cassayre, L.; Rhodes, C.; Ekberg, C.

    2012-07-01

    Nuclear energy has more than ever to demonstrate that it can contribute safely and on a sustainable way to answer the international increase in energy needs. Actually, in addition to an increased safety of the reactors themselves, its acceptance is still closely associated to our capability to reduce the lifetime of the nuclear waste, to manage them safely and to propose options for a better use of the natural resources. Spent fuel reprocessing can help to reach these objectives. But this cannot be achieved only by optimizing industrial processes through engineering studies. It is of a primary importance to increase our fundamental knowledge in actinide sciences in order to build the future of nuclear energy on reliable and scientifically-founded results, and therefore meet the needs of the future fuel cycles in terms of fabrication and performance of fuels, reprocessing and waste management. At the European level, both the collaborative project ACSEPT and the Integrated Infrastructure Initiative ACTINET-I3 work together to improve our knowledge in actinides chemistry and therefore develop advanced separation processes. These tools are complementary and work in close connection on some specific issues such as the understanding of the selectivity of extracting organic ligands. By offering trans-national access to the main nuclear research facility in Europe, ACTINET-I3 aims at increasing the knowledge in actinide sciences by gathering all the expertise available in European nuclear research institutes or university and giving them the opportunity to come and work in hot-labs (ITU, Atalante...) or beamlines (ESFR, ANKA, PSI) ACSEPT is focused on the development of advanced separation processes, both aqueous and pyrochemical. Head-end steps, fuel re-fabrication, solvent treatment, waste management are also taken into account. In aqueous process development, the SANEX and innovative SANEX flowsheets demonstration were successfully achieved. Chemical systems were

  16. Thermochemical Processing of Radioactive Waste Using Powder Metal Fuels

    SciTech Connect

    Ojovan, M. I.; Sobolev, I. A.; Dmitriev, S. A.; Panteleev, V. I.; Karlina, O. K.; Klimov. V. L.

    2003-02-25

    Problematic radioactive wastes were generated during various activities of both industrial facilities and research institutions usually in relative small amounts. These can be spent ion exchange resins, inorganic absorbents, wastes from research nuclear reactors, irradiated graphite, mixed, organic or chlorine-containing radioactive waste, contaminated soils, un-burnable heavily surface-contaminated materials, etc. Conventional treatment methods encounter serious problems concerning processing efficiency of such waste, e.g. complete destruction of organic molecules and avoiding of possible emissions of radionuclides, heavy metals and chemically hazardous species. Some contaminations cannot be removed from surface using common decontamination methods. Conditioning of ash residues obtained after treatment of solid radioactive waste including ashes received from treating problematic wastes also is a complicated task. Moreover due to relative small volume of specific type radioactive waste the development of target treatment procedures and facilities to conduct technological processes and their deployment could be economically unexpedient and ecologically no justified. Thermochemical processing technologies are used for treating and conditioning problematic radioactive wastes. The thermochemical processing uses powdered metal fuels (PMF) that are specifically formulated for the waste composition and react chemically with the waste components. The composition of the PMF is designed in such a way as to minimize the release of hazardous components and radionuclides in the off gas and to confine the contaminants in the ash residue. The thermochemical procedures allow decomposition of organic matter and capturing hazardous radionuclides and chemical species simultaneously. A significant advantage of thermochemical processing is its autonomy. Thermochemical treatment technologies use the energy of exothermic reactions in the mixture of radioactive or hazardous waste with PMF

  17. Fuel quality-processing study. Volume 1: Overview and results

    NASA Technical Reports Server (NTRS)

    Jones, G. E., Jr.

    1982-01-01

    The methods whereby the intermediate results were obtained are outlined, and the evaluation of the feasible paths from liquid fossil fuel sources to generated electricity is presented. The segments from which these paths were built are the results from the fuel upgrading schemes, on-site treatments, and exhaust gas treatments detailed in the subsequent volumes. The salient cost and quality parameters are included.

  18. Process for making film-bonded fuel cell interfaces

    DOEpatents

    Kaufman, Arthur; Terry, Peter L.

    1990-07-03

    An improved interface configuration for use between adjacent elements of a fuel cell stack. The interface is impervious to gas and liquid and provides resistance to corrosion by the electrolyte of the fuel cell. A multi-layer arrangement for the interface provides bridging electrical contact with a hot-pressed resin filling the void space.

  19. Process simulation of a PEM fuel cell system

    SciTech Connect

    Ledjeff-Hey, K.; Roes, J.; Formanski, V.; Gieshoff, J.; Vogel, B.

    1996-01-01

    The thermodynamic performance of a PEM fuel cell system for producing electrical power from natural gas is investigated by considering the flows of energy and energy through the various steps of the whole system. The flows of energy are evaluated using a computer code for energy and energy analyses. The fuel cell system is designed to produce a hydrogen volumetric flow of nearly 5.0 m{sup 3} {sub NTP}/h, provided to the fuel cell at an absolute pressure of 2.9 bar. The fuel cell itself is working with an efficiency of about 60 % at an operating temperature of 65 - 75{degrees} C with an air ratio of four and provides a maximum electric power of 9 kW. Taking into consideration only the produced electric power as useful output of the fuel cell system a total efficiency of 42.2 % is calculated using the simulation results.

  20. Fuel Quality/Processing Study. Volume II. Appendix, Task I, literature survey

    SciTech Connect

    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.

  1. Heterogeneous catalytic process for alcohol fuels from syngas

    SciTech Connect

    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.

  2. AFCI Transmutation Fuel Processes and By-Products Planning: Interim Report

    SciTech Connect

    Eric L. Shaber

    2005-09-01

    dictates the need for detailed process flows, mass balances, batch size data, and radiological dose estimates. Full definition of the materials that will need to be handled in the facility as feed material inputs, in-process fuel, scrap recycle, scrap requiring recovery, and by-product wastes is required. The feed material for demonstrating transmutation fuel fabrication will need to come from the separations of actinides from spent nuclear fuel processed in the same AFCF.

  3. Reprocessed polylactide: studies of thermo-oxidative decomposition.

    PubMed

    Badia, J D; Santonja-Blasco, L; Martínez-Felipe, A; Ribes-Greus, A

    2012-06-01

    The combustion process of virgin and reprocessed polylactide (PLA) was simulated by multi-rate linear non-isothermal thermogravimetric experiments under O(2). A complete methodology that accounted on the thermal stability and emission of gases was thoroughly developed. A new model, Thermal Decomposition Behavior, and novel parameters, the Zero-Decomposition Temperatures, were used to test the thermal stability of the materials under any linear heating rate. The release of gases was monitored by Evolved Gas Analysis with in-line FT-IR analysis. In addition, a kinetic analysis methodology that accounted for variable activation parameters showed that the decomposition process could be driven by the formation of bubbles in the melt. It was found that the combustion technologies for virgin PLA could be transferred for the energetic valorization of its recyclates. Combustion was pointed out as appropriate for the energetic valorization of PLA submitted to more than three successive reprocessing cycles. PMID:22481003

  4. Co-Rolled U10Mo/Zirconium-Barrier-Layer Monolithic Fuel Foil Fabrication Process

    SciTech Connect

    G. A. Moore; M. C. Marshall

    2010-01-01

    Integral to the current UMo fuel foil processing scheme being developed at Idaho National Laboratory (INL) is the incorporation of a zirconium barrier layer for the purpose of controlling UMo-Al interdiffusion at the fuel-meat/cladding interface. A hot “co-rolling” process is employed to establish a ~25-µm-thick zirconium barrier layer on each face of the ~0.3-mm-thick U10Mo fuel foil.

  5. Drying studies for corroded DOE aluminum plate fuels

    SciTech Connect

    Lords, R.E.; Windes, W.E.; Crepeau, J.C.; Sidwell, R.W.

    1996-05-01

    The Idaho National Engineering Laboratory (INEL) currently stores a wide variety of spent nuclear fuel. The fuel was originally intended to be stored underwater for a short period of thermal cooling, then removed and reprocessed. However, it has been stored underwater for much longer thank originally anticipated. During this time dust and airborne desert soil have entered the oldest INEL pool, accumulating on the fuel. Also, the aluminum fuel cladding has corroded compromising the exposed surfaces of the fuel. Plans are now underway to move some the the more vulnerable aluminum plate type fuels into dry storage in an existing vented and filtered fuel storage facility. In preparation for dry storage of the fuel a drying and canning station is being built at the INEL. The two primary objectives of this facility are to determine the influence of corrosion products on the drying process and to establish temperature distribution inside the canister during heating.

  6. On the Reprocessing and Reanalysis of Observations for Climate

    NASA Technical Reports Server (NTRS)

    Bosilovich, Michael G.; Kennedy, John; Dee, Dick; Allan, R.; O'Neill, Alan

    2013-01-01

    The long observational record is critical to our understanding of the Earths climate, but most observing systems were not developed with a climate objective in mind. As a result, tremendous efforts have gone into assessing and reprocessing the data records to improve their usefulness in climate studies. The purpose of this paper is to both review recent progress in reprocessing and reanalyzing observations, and to summarize the challenges that must be overcome in order to improve our understanding of climate and variability. Reprocessing improves data quality through more scrutiny and improved retrieval techniques for individual observing systems, while reanalysis merges many disparate observations with models through data assimilation, yet both aim to provide an climatology of Earth processes. Many challenges remain, such as tracking the improvement of processing algorithms and limited spatial coverage. Reanalyses have fostered significant research, yet reliable global trends in many physical fields are not yet attainable, despite significant advances in data assimilation and numerical modeling. Oceanic reanalyses have made significant advances in recent years, but will only be discussed here in terms of progress toward integrated Earth system analyses. Climate data sets are generally adequate for process studies and large-scale climate variability. Communication of the strengths, limitations and uncertainties of reprocessed observations and reanalysis data, not only among the community of developers, but also with the extended research community, including the new generations of researchers and the decision makers is crucial for further advancement of the observational data records. It must be emphasized that careful investigation of the data and processing methods are required to use the observations appropriately.

  7. Automated brush plating process for solid oxide fuel cells

    SciTech Connect

    Long, Jeffrey William

    2003-01-01

    A method of depositing a metal coating (28) on the interconnect (26) of a tubular, hollow fuel cell (10) contains the steps of providing the fuel cell (10) having an exposed interconnect surface (26); contacting the inside of the fuel cell (10) with a cathode (45) without use of any liquid materials; passing electrical current through a contacting applicator (46) which contains a metal electrolyte solution; passing the current from the applicator (46) to the cathode (45) and contacting the interconnect (26) with the applicator (46) and coating all of the exposed interconnect surface.

  8. BIOMASS AND FOSSIL FUEL TO METHANOL AND CARBON VIA THE HYDROCARB PROCESS

    EPA Science Inventory

    The paper discusses the production of methanol and carbon from biomass and fossil fuels, utilizing the Hydrocarbon process. This process has the potential to minimize dependence on imported fuels for the transportation and utility sectors by increasing the yield and reducing the ...

  9. Numerical approach for the voloxidation process of an advanced spent fuel conditioning process (ACP)

    SciTech Connect

    Park, Byung Heung; Jeong, Sang Mun; Seo, Chung-Seok

    2007-07-01

    A voloxidation process is adopted as the first step of an advanced spent fuel conditioning process in order to prepare the SF oxide to be reduced in the following electrolytic reduction process. A semi-batch type voloxidizer was devised to transform a SF pellet into powder. In this work, a simple reactor model was developed for the purpose of correlating a gas phase flow rate with an operation time as a numerical approach. With an assumption that a solid phase and a gas phase are homogeneous in a reactor, a reaction rate for an oxidation was introduced into a mass balance equation. The developed equation can describe a change of an outlet's oxygen concentration including such a case that a gas flow is not sufficient enough to continue a reaction at its maximum reaction rate. (authors)

  10. Fluid-bed fluoride volatility process recovers uranium from spent uranium alloy fuels

    NASA Technical Reports Server (NTRS)

    Barghusen, J. J.; Chilenskas, A. A.; Gunderson, G. E.; Holmes, J. T.; Jonke, A. A.; Kincinas, J. E.; Levitz, N. M.; Potts, G. L.; Ramaswami, D.; Stethers, H.; Turner, K. S.

    1967-01-01

    Fluid-bed fluoride volatility process recovers uranium from uranium fuels containing either zirconium or aluminum. The uranium is recovered as uranium hexafluoride. The process requires few operations in simple, compact equipment, and eliminates aqueous radioactive wastes.

  11. Deep Burn Develpment of Transuranic Fuel for High-Temperature Helium-Cooled Reactors - July 2010

    SciTech Connect

    Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D; Bell, Gary L

    2010-08-01

    The DB Program Quarterly Progress Report for April - June 2010, ORNL/TM/2010/140, was distributed to program participants on August 4. This report discusses the following: (1) TRU (transuranic elements) HTR (high temperature helium-cooled reactor) Fuel Modeling - (a) Thermochemical Modeling, (b) 5.3 Radiation Damage and Properties; (2) TRU HTR Fuel Qualification - (a) TRU Kernel Development, (b) Coating Development, (c) ZrC Properties and Handbook; and (3) HTR Fuel Recycle - (a) Recycle Processes, (b) Graphite Recycle, (c) Pyrochemical Reprocessing - METROX (metal recovery from oxide fuel) Process Development.

  12. Features, Events and Processes for the Used Fuel Disposition Campaign

    SciTech Connect

    Blink, J A; Greenberg, H R; Caporuscio, F A; Houseworth, J E; Freeze, G A; Mariner, P; Cunnane, J C

    2010-12-15

    The Used Fuel Disposition (UFD) Campaign within DOE-NE is evaluating storage and disposal options for a range of waste forms and a range of geologic environments. To assess the potential performance of conceptual repository designs for the combinations of waste form and geologic environment, a master set of Features, Events, and Processes (FEPs) has been developed and evaluated. These FEPs are based on prior lists developed by the Yucca Mountain Project (YMP) and the international repository community. The objective of the UFD FEPs activity is to identify and categorize FEPs that are important to disposal system performance for a variety of disposal alternatives (i.e., combinations of waste forms, disposal concepts, and geologic environments). FEP analysis provides guidance for the identification of (1) important considerations in disposal system design, and (2) gaps in the technical bases. The UFD FEPs also support the development of performance assessment (PA) models to evaluate the long-term performance of waste forms in the engineered and geologic environments of candidate disposal system alternatives. For the UFD FEP development, five waste form groups and seven geologic settings are being considered. A total of 208 FEPs have been identified, categorized by the physical components of the waste disposal system as well as cross-cutting physical phenomena. The combination of 35 waste-form/geologic environments and 208 FEPs is large; however, some FEP evaluations can cut across multiple waste/environment combinations, and other FEPs can be categorized as not-applicable for some waste/environment combinations, making the task of FEP evaluation more tractable. A FEP status tool has been developed to document progress. The tool emphasizes three major areas that can be statused numerically. FEP Applicability documents whether the FEP is pertinent to a waste/environment combination. FEP Completion Status documents the progress of the evaluation for the FEP

  13. H2S removal with ZnO during fuel processing for PEM fuel cell applications

    SciTech Connect

    Li, Liyu; King, David L.

    2006-09-15

    The possibility of using ZnO as a H2S absorbent to protect catalysts in the gasoline and diesel fuel processor for PEM fuel cell applications was studied. It is possible to use commercial ZnO absorbent as a guard bed to protect the PROX catalyst and PEM fuel cell. However, it is not feasible to use ZnO to protect high and low temperature WGS catalysts, most likely due to COS formation via reactions CO + H2S = COS + H2 and CO2 + H2S = COS + H2O.

  14. Sensitivity analysis of a dry-processed Candu fuel pellet's design parameters

    SciTech Connect

    Choi, Hangbok; Ryu, Ho Jin

    2007-07-01

    Sensitivity analysis was carried out in order to investigate the effect of a fuel pellet's design parameters on the performance of a dry-processed Canada deuterium uranium (CANDU) fuel and to suggest the optimum design modifications. Under a normal operating condition, a dry-processed fuel has a higher internal pressure and plastic strain due to a higher fuel centerline temperature when compared with a standard natural uranium CANDU fuel. Under a condition that the fuel bundle dimensions do not change, sensitivity calculations were performed on a fuel's design parameters such as the axial gap, dish depth, gap clearance and plenum volume. The results showed that the internal pressure and plastic strain of the cladding were most effectively reduced if a fuel's element plenum volume was increased. More specifically, the internal pressure and plastic strain of the dry-processed fuel satisfied the design limits of a standard CANDU fuel when the plenum volume was increased by one half a pellet, 0.5 mm{sup 3}/K. (authors)

  15. Properties of Aluminum Deposited by a High-Velocity Oxygen-Fueled Process

    SciTech Connect

    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.

  16. Hallam fuel decladding. Program summary report

    SciTech Connect

    Dennison, W.F.

    1980-04-01

    This report summarizes the program of decladding the 150 Hallam fuel assemblies, removal of the sodium, and the packaging and shipment of the recovered fuel to Savannah River for eventual reprocessing.

  17. CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP601) ON THE RIGHT ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP-601) ON THE RIGHT AND LABORATORY (CPP-602) ON THE LEFT. INL PHOTO NUMBER NRTS-51-3373. Unknown Photographer, 9/28/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  18. EAST ELEVATION OF MAIN PROCESSING BUILDING (CPP601) LOOKING NORTHWEST. MAINTENANCE ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    EAST ELEVATION OF MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTHWEST. MAINTENANCE SHOP AND OFFICE BUILDING (CPP-630) ON RIGHT IN PHOTO. INL PHOTO NUMBER HD-22-3-2. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  19. SOUTH SECTION OF WEST ELEVATION OF MAIN PROCESSING BUILDING (CPP601) ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    SOUTH SECTION OF WEST ELEVATION OF MAIN PROCESSING BUILDING (CPP-601) LOOKING EAST. HEADEND PLANT BUILDING (CPP-640) APPEARS ON LEFT IN PHOTO. INL PHOTO NUMBER HD-22-3-3. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  20. EQUIPMENT LAYOUT OF MAIN PROCESSING BUILDING (CPP601) LCELL PLAN AND ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    EQUIPMENT LAYOUT OF MAIN PROCESSING BUILDING (CPP-601) L-CELL PLAN AND SECTION SHOWS COMPLEXITY OF CELLS. INL DRAWING NUMBER 200-0601-00-098-105687. ALTERNATE ID NUMBER 4289-20-301. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  1. SOUTH ELEVATION AND DETAILS OF MAIN PROCESSING BUILDING (CPP601). INL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    SOUTH ELEVATION AND DETAILS OF MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103082. ALTERNATE ID NUMBER 542-12-B-76. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  2. BUILDING DETAILS AND SECTIONS OF MAIN PROCESSING BUILDING (CPP601). INL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    BUILDING DETAILS AND SECTIONS OF MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103080. ALTERNATE ID NUMBER 542-11-B-74. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  3. STRUCTURAL DETAILS AND SECTIONS OF MAIN PROCESSING BUILDING (CPP601). INL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    STRUCTURAL DETAILS AND SECTIONS OF MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103079. ALTERNATE ID NUMBER 542-11-B-73. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  4. EAST AND WEST ELEVATIONS OF MAIN PROCESSING BUILDING (CPP601). INL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    EAST AND WEST ELEVATIONS OF MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103081. ALTERNATE ID NUMBER 542-11-B-75. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

  5. Reprocessing of LiH in Molten Chlorides

    NASA Astrophysics Data System (ADS)

    Masset, Patrick J.; Gabriel, Armand; Poignet, Jean-Claude

    2008-06-01

    LiH was used as inactive material to stimulate the reprocessing of lithium tritiate in molten chlorides. The electrochemical properties (diffusion coefficients, apparent standard potentials) were measured by means of transient electrochemical techniques (cyclic voltammetry and chronopotentiometry). At 425 ºC the diffusion coefficient and the apparent standard potential were 2.5 · 10-5 cm2 s-1 and -1.8 V vs. Ag/AgCl, respectively. For the process design the LiH solubility was measured by means of DTA to optimize the LiH concentration in the molten phase. In addition electrolysis tests were carried out at 460 ºC with current densities up to 1 A cm-2 over 24 h. These results show that LiH may be reprocessed in molten chlorides consisting in the production of hydrogen gas at the anode and molten metallic lithium at the cathode.

  6. On the Reprocessing and Reanalysis of Observations for Climate

    NASA Technical Reports Server (NTRS)

    Bosilovich, Michael G.; Kennedy, John; Dee, Dick; ONeill, Alan

    2012-01-01

    The long observational record is critical to our understanding of the Earth s climate, but most observing systems were not developed with a climate objective in mind. As a result, tremendous efforts have gone into assessing and reprocessing the data records to improve their usefulness in climate studies. Many challenges remain, such as tracking the improvement of processing algorithms and limited spatial coverage. Reanalyses have fostered significant research, yet reliable global trends in many physical fields are not yet attainable, despite significant advances in data assimilation and numerical modeling. Communication of the strengths, limitations and uncertainties of reprocessed observations and reanalysis data, not only among the community of developers, but also with the extended research community, including the new generations of researchers and the decision makers is crucial for further advancement of the observational data records. WCRP provides the means to bridge the different motivating objectives on which national efforts focus.

  7. Plutonium production story at the Hanford site: processes and facilities history

    SciTech Connect

    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.

  8. Automatic inspection for remotely manufactured fuel elements

    SciTech Connect

    Reifman, J.; Vitela, J.E.; Gibbs, K.S.; Benedict, R.W.

    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.

  9. Technical and economic feasibility of alternative fuel use in process heaters and small boilers

    SciTech Connect

    Not Available

    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.

  10. Material flow simulation in a nuclear chemical process

    SciTech Connect

    Mahgerefteh, M.

    1984-01-01

    At a nuclear fuel reprocessing plant the special nuclear materials (SNM) are received as constituents of spent fuel assemblies, are converted to liquid form, and undergo a series of chemical processes. Uncertainties in measurements of SNM at each stage of reprocessing limit the accuracy of simple material balance accounting as a safeguards method. To be effective, a formal safeguards program must take into account all sources of measurement error yet detect any diversion of SNM. An analytical method for assessing the accountability of selected constituent SNM is demonstrated. A combined discrete-continuous, time-dependent model using the GASP IV simulation language is developed to simulate mass flow, material accountability and measurement error at each stage of the reprocessing plant.

  11. The Nuclear Energy Advanced Modeling and Simulation Safeguards and Separations Reprocessing Plant Toolkit

    SciTech Connect

    McCaskey, Alex; Billings, Jay Jay; de Almeida, Valmor F

    2011-08-01

    This report details the progress made in the development of the Reprocessing Plant Toolkit (RPTk) for the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. RPTk is an ongoing development effort intended to provide users with an extensible, integrated, and scalable software framework for the modeling and simulation of spent nuclear fuel reprocessing plants by enabling the insertion and coupling of user-developed physicochemical modules of variable fidelity. The NEAMS Safeguards and Separations IPSC (SafeSeps) and the Enabling Computational Technologies (ECT) supporting program element have partnered to release an initial version of the RPTk with a focus on software usability and utility. RPTk implements a data flow architecture that is the source of the system's extensibility and scalability. Data flows through physicochemical modules sequentially, with each module importing data, evolving it, and exporting the updated data to the next downstream module. This is accomplished through various architectural abstractions designed to give RPTk true plug-and-play capabilities. A simple application of this architecture, as well as RPTk data flow and evolution, is demonstrated in Section 6 with an application consisting of two coupled physicochemical modules. The remaining sections describe this ongoing work in full, from system vision and design inception to full implementation. Section 3 describes the relevant software development processes used by the RPTk development team. These processes allow the team to manage system complexity and ensure stakeholder satisfaction. This section also details the work done on the RPTk ``black box'' and ``white box'' models, with a special focus on the separation of concerns between the RPTk user interface and application runtime. Section 4 and 5 discuss that application runtime component in more detail, and describe the dependencies, behavior, and rigorous testing of its constituent components.

  12. Ventilation systems for a spent LWR fuel recycle complex

    SciTech Connect

    Not Available

    1981-01-01

    A conceptual design study has been made of a facility to recycle spent Light Water Reactor fuel. This study was based on coprocessing of plutonium and uranium where plutonium is never available as a separate material. The design of the fuel reprocessing facilities is based on remote operation and remote maintenance. The experience of many years of safe and dependable operation of government fuel processing facilities at Savannah River and Hanford was used in the design. A requirement of the study was that the facilities be licensable under Title 10 and Title 40 of the Code of Federal Regulations.

  13. Literature on fabrication of tungsten for application in pyrochemical processing of spent nuclear fuels

    SciTech Connect

    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.

  14. Processes for converting biomass-derived feedstocks to chemicals and liquid fuels

    DOEpatents

    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.

  15. Flowsheet for shear/leach processing of N Reactor fuel at PUREX

    SciTech Connect

    Enghusen, M.B.

    1995-04-13

    This document was originally prepared to support the restart of the PUREX plant using a new Shear/Leach head end process. However, the PUREX facility was shutdown and processing of the remaining N Reactor fuel is no longer considered an alternative for fuel disposition. This document is being issued for reference only to document the activities which were investigated to incorporate the shear/leach process in the PUREX plant.

  16. Processing of FRG mixed oxide fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements

    SciTech Connect

    Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Tischer, H.E.

    1980-11-01

    The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment section of the agreement, FRG fuel spheres were recently sent for processing in the Department of Energy sponsored cold pilot plant for High-Temperature Gas-Cooled Reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles. These particles were in turn crushed and burned to recover the fuel-bearing kernels for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated the applicability of the US HTGR fuel treatment flowsheet to FRG fuel processing. 10 figures.

  17. Processing of FRG high-temperature gas-cooled reactor fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements

    SciTech Connect

    Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Drake, R.N.

    1981-11-01

    The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects of gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment development section of the agreement, both FRG mixed uranium/ thorium and low-enriched uranium fuel spheres have been processed in the Department of Energy-sponsored cold pilot plant for high-temperature gas-cooled reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles suitable for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated certain modifications to the US HTGR fuel burining process necessary for FRG fuel treatment. Results of the tests will be used in the design of a US/FRG joint prototype headend facility for HTGR fuel.

  18. Process technologies for production of fuel ethanol from lignocellulosic biomass

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Lignocellulosic biomass such as corn stover, wheat straw, rice straw, and switchgrass can serve as low-cost feedstock for production of fuel ethanol. These feedstocks contain complex carbohydrates (cellulose and hemicelluloses) which need to be converted to fermentable sugars and then these sugars b...

  19. COAL PROCESSING TECHNOLOGY: ENVIRONMENTAL IMPACT OF SYNTHETIC FUELS DEVELOPMENT

    EPA Science Inventory

    This paper describes a portion of a comprehensive, long term EPA program to evaluate the environmental effects of energy systems that produce gaseous and liquid fuels from coal. The technologies being assessed include low-, medium-, and high-Btu gasification and liquefaction. One...

  20. Energy and exergy analysis of an ethanol reforming process for solid oxide fuel cell applications.

    PubMed

    Tippawan, Phanicha; Arpornwichanop, Amornchai

    2014-04-01

    The fuel processor in which hydrogen is produced from fuels is an important unit in a fuel cell system. The aim of this study is to apply a thermodynamic concept to identify a suitable reforming process for an ethanol-fueled solid oxide fuel cell (SOFC). Three different reforming technologies, i.e., steam reforming, partial oxidation and autothermal reforming, are considered. The first and second laws of thermodynamics are employed to determine an energy demand and to describe how efficiently the energy is supplied to the reforming process. Effect of key operating parameters on the distribution of reforming products, such as H2, CO, CO2 and CH4, and the possibility of carbon formation in different ethanol reformings are examined as a function of steam-to-ethanol ratio, oxygen-to-ethanol ratio and temperatures at atmospheric pressure. Energy and exergy analysis are performed to identify the best ethanol reforming process for SOFC applications. PMID:24561628

  1. Reprocessing of metallurgical slag into materials for the building industry

    SciTech Connect

    Pioro, L.S.; Pioro, I.L

    2004-07-01

    Several methods of reprocessing metallurgical (blast furnace) slag into materials for the building industry, based on melting aggregates with submerged combustion, were developed and tested. The first method involves melting hot slag with some additives directly in a slag ladle with a submerged gas-air burner, with the objective of producing stabilized slag or glass-ceramic. The second method involves direct draining of melted slag from a ladle into the slag receiver, with subsequent control of the slag draining into the converter where special charging materials are added to the melt, with the objective of producing glass-ceramic. A third method involves melting cold slag with some additives inside a melting converter with submerged gas-air burners, with the objective of producing glass-ceramic fillers for use in road construction. Specific to the melting process is the use of a gas-air mixture with direct combustion inside the melt. This feature provides melt bubbling to help achieve maximum heat transfer from combustion products to the melt, improve mixing (and therefore homogeneity of the melt), and increases the rate of chemical reactions. The experimental data for different aspects of the proposed methods are presented. The reprocessed blast-furnace slag in the form of granules can be used as fillers for concretes, asphalts, and as additives in the production of cement, bricks and other building materials. As well, reprocessed blast-furnace slag can be poured into forms for the production of glass-ceramic tiles.

  2. Reprocessing of metallurgical slag into materials for the building industry.

    PubMed

    Pioro, L S; Pioro, I L

    2004-01-01

    Several methods of reprocessing metallurgical (blast furnace) slag into materials for the building industry, based on melting aggregates with submerged combustion, were developed and tested. The first method involves melting hot slag with some additives directly in a slag ladle with a submerged gas-air burner, with the objective of producing stabilized slag or glass-ceramic. The second method involves direct draining of melted slag from a ladle into the slag receiver, with subsequent control of the slag draining into the converter where special charging materials are added to the melt, with the objective of producing glass-ceramic. A third method involves melting cold slag with some additives inside a melting converter with submerged gas-air burners, with the objective of producing glass-ceramic fillers for use in road construction. Specific to the melting process is the use of a gas-air mixture with direct combustion inside the melt. This feature provides melt bubbling to help achieve maximum heat transfer from combustion products to the melt, improve mixing (and therefore homogeneity of the melt), and increases the rate of chemical reactions. The experimental data for different aspects of the proposed methods are presented. The reprocessed blast-furnace slag in the form of granules can be used as fillers for concretes, asphalts, and as additives in the production of cement, bricks and other building materials. As well, reprocessed blast-furnace slag can be poured into forms for the production of glass-ceramic tiles. PMID:15081065

  3. CHEMICALLY ACTIVE FLUID BED PROCESS FOR SULPHUR REMOVAL DURING GASIFICATION OF CARBONACEOUS FUELS

    EPA Science Inventory

    The report covers the final 3 years of a 9-year program to evaluate the Chemically Active Fluid Bed (CAFB) process for gasification and desulfurization of liquid and solid fuels in a fluidized bed of hot lime. A range of alternative fuels, including three coals and a lignite, wer...

  4. SYSTEM AND PROCESS FOR PRODUCTION OF METHANOL FROM COMBINED WIND TURBINE AND FUEL CELL POWER

    EPA Science Inventory

    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 ...

  5. 10 CFR 140.13a - Amount of financial protection required for plutonium processing and fuel fabrication plants.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... plutonium processing and fuel fabrication plant is required to have and maintain financial protection in the... use plutonium at two or more plutonium processing and fuel fabrication plants at the same location... protection covers all such plants at the location....

  6. 10 CFR 140.13a - Amount of financial protection required for plutonium processing and fuel fabrication plants.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... plutonium processing and fuel fabrication plant is required to have and maintain financial protection in the... use plutonium at two or more plutonium processing and fuel fabrication plants at the same location... protection covers all such plants at the location....

  7. 10 CFR 140.13a - Amount of financial protection required for plutonium processing and fuel fabrication plants.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... plutonium processing and fuel fabrication plant is required to have and maintain financial protection in the... use plutonium at two or more plutonium processing and fuel fabrication plants at the same location... protection covers all such plants at the location....

  8. SAF: the next generation process for radiotoxic material handling in the nuclear fuel industry

    SciTech Connect

    Nyman, D.H.; Graham, R.A.

    1984-07-19

    In 1980 the Secure Automated Fabrication (SAF) Project was established with the goal to design, build, and operate a remote process for manufacturing breeder reactor fuel pins. The SAF line will be housed in the Fuels and Materials Examination Facility (FMEF) at the Hanford site. The fabrication system and supporting operations are designed for computer-controlled operation from a centralized control room. In addition to improved worker protection, remote and automated fuel fabrication operations will result in enhanced safeguards and accountability of fuel material, improved product quality, and increased productivity. Installation of the SAF line equipment has started. Qualification runs are scheduled to begin in 1986 with production commencing in 1987.

  9. Fuel-Flexible Combustion System for Refinery and Chemical Plant Process Heaters

    SciTech Connect

    2010-06-01

    Funded by the American Recovery and Reinvestment Act of 2009 ENVIRON International Corporation, in collaboration with Callidus Technologies by Honeywell and Shell Global Solutions, Inc., will develop and demonstrate a full-scale fuel blending and combustion system. This system will allow a broad range of opportunity fuel compositions, including syngas, biogas, natural gas, and refinery fuel gas, to be safely, cost-effectively, and efficiently utilized while generating minimal emissions of criteria pollutants. The project will develop a commercial technology for application in refinery and chemical plant process heaters where opportunity fuels are used.

  10. Development of Hot Pressing as a Low Cost Processing Technique for Fuel Cell Fabrication

    SciTech Connect

    Sarin, V

    2003-01-14

    Dependable, plentiful, and economical energy has been the driving force for financial, industrial, and political growth in the US since the mid 19th century. For a country whose progress is so deeply rooted in abundant energy and whose current political agenda involves stabilizing world fossil fuel prices, the development of a reliable, efficient and environmentally friendly power generating source seems compulsory. The maturing of high technology fuel cells may be the panacea the country will find indispensable to free itself from foreign dependence. Fuel cells offer an efficient, combustion-less, virtually pollution-free power source, capable of being sited in downtown urban areas or in remote regions. Fuel cells have few moving parts and run almost silently. Fuel cells are electrochemical devices that convert the chemical energy of a fuel directly to electrical energy. Unlike batteries, which store a finite amount of energy, fuel cells will generate electricity continuously, as long as fuel and oxidant are available to the electrodes. Additionally, fuel cells offer clean, efficient, and reliable power and they can be operated using a variety of fuels. Hence, the fuel cell is an extremely promising technology. Over the course of this research, the fundamental knowledge related to ceramic processing, sintering, and hot pressing to successfully hot press a single operational SOFC in one step has been developed. Ceramic powder processing for each of the components of an SOFC has bene tailored towards this goal. Processing parameter for the electrolyte and cathode have been studied and developed until they converted. Several anode fabrication techniques have been developed. Additionally, a novel anode structured has been developed and refined. These individual processes have been cultivated until a single cell SOFC has been fabricated in one step.

  11. 40 CFR 80.1416 - Petition process for evaluation of new renewable fuels pathways.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... make it, and the production process. The justification must include process modeling flow charts. (iii... 40 Protection of Environment 16 2010-07-01 2010-07-01 false Petition process for evaluation of new... Petition process for evaluation of new renewable fuels pathways. (a) Pursuant to this section, a party...

  12. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    DOE PAGESBeta

    Collette, R.; King, J.; Buesch, C.; Keiser, Jr., D. D.; Williams, W.; Miller, B. D.; Schulthess, J.

    2016-04-01

    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 whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

  13. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    NASA Astrophysics Data System (ADS)

    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.

  14. Spent Nuclear Fuel (SNF) Process Validation Technical Support Plan

    SciTech Connect

    SEXTON, R.A.

    2000-05-09

    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.

  15. Electrical start-up for diesel fuel processing in a fuel-cell-based auxiliary power unit

    NASA Astrophysics Data System (ADS)

    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.

  16. Chemical Engineering Division fuel cycle programs. Quarterly progress report, April-June 1979. [Pyrochemical/dry processing; waste encapsulation in metal; transport in geologic media

    SciTech Connect

    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.

  17. Development of pyro-processing technology for thorium-fuelled molten salt reactor

    SciTech Connect

    Uhlir, J.; Straka, M.; Szatmary, L.

    2012-07-01

    The Molten Salt Reactor (MSR) is classified as the non-classical nuclear reactor type based on the specific features coming out from the use of liquid fuel circulating in the MSR primary circuit. Other uniqueness of the reactor type is based on the fact that the primary circuit of the reactor is directly connected with the on-line reprocessing technology, necessary for keeping the reactor in operation for a long run. MSR is the only reactor system, which can be effectively operated within the {sup 232}Th- {sup 233}U fuel cycle as thorium breeder with the breeding factor significantly higher than one. The fuel cycle technologies proposed as ford the fresh thorium fuel processing as for the primary circuit fuel reprocessing are pyrochemical and mainly fluoride. Although these pyrochemical processes were never previously fully verified, the present-day development anticipates an assumption for the successful future deployment of the thorium-fuelled MSR technology. (authors)

  18. Fundamental phenomena on fuel decomposition and boundary layer combustion processes with applications to hybrid rocket motors

    NASA Astrophysics Data System (ADS)

    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.

  19. Fundamental phenomena on fuel decomposition and boundary layer combustion processes with applications to hybrid rocket motors

    NASA Technical Reports Server (NTRS)

    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.

  20. System Design Description and Requirements for Modeling the Off-Gas Systems for Fuel Recycling Facilities

    SciTech Connect

    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.

  1. TRISO-Coated Fuel Processing to Support High Temperature Gas-Cooled Reactors

    SciTech Connect

    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 technologies used in

  2. Solid recovered fuel: influence of waste stream composition and processing on chlorine content and fuel quality.

    PubMed

    Velis, Costas; Wagland, Stuart; Longhurst, Phil; Robson, Bryce; Sinfield, Keith; Wise, Stephen; Pollard, Simon

    2012-02-01

    Solid recovered fuel (SRF) produced by mechanical-biological treatment (MBT) of municipal waste can replace fossil fuels, being a CO(2)-neutral, affordable, and alternative energy source. SRF application is limited by low confidence in quality. We present results for key SRF properties centered on the issue of chlorine content. A detailed investigation involved sampling, statistical analysis, reconstruction of composition, and modeling of SRF properties. The total chlorine median for a typical plant during summer operation was 0.69% w/w(d), with lower/upper 95% confidence intervals of 0.60% w/w(d) and 0.74% w/w(d) (class 3 of CEN Cl indicator). The average total chlorine can be simulated, using a reconciled SRF composition before shredding to <40 mm. The relative plastics vs paper mass ratios in particular result in an SRF with a 95% upper confidence limit for ash content marginally below the 20% w/w(d) deemed suitable for certain power plants; and a lower 95% confidence limit of net calorific value (NCV) at 14.5 MJ kg(ar)(-1). The data provide, for the first time, a high level of confidence on the effects of SRF composition on its chlorine content, illustrating interrelationships with other fuel properties. The findings presented here allow rational debate on achievable vs desirable MBT-derived SRF quality, informing the development of realistic SRF quality specifications, through modeling exercises, needed for effective thermal recovery. PMID:22191490

  3. Development and control of the process for the manufacture of zircaloy-4 tubing for LWBR fuel rods

    SciTech Connect

    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.

  4. Advanced Fuel Cycle Cost Basis

    SciTech Connect

    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.

  5. Advanced Fuel Cycle Cost Basis

    SciTech Connect

    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.

  6. Advanced Fuel Cycle Cost Basis

    SciTech Connect

    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.

  7. Hydrodeoxygenation processes: advances on catalytic transformations of biomass-derived platform chemicals into hydrocarbon fuels.

    PubMed

    De, Sudipta; Saha, Basudeb; Luque, Rafael

    2015-02-01

    Lignocellulosic biomass provides an attractive source of renewable carbon that can be sustainably converted into chemicals and fuels. Hydrodeoxygenation (HDO) processes have recently received considerable attention to upgrade biomass-derived feedstocks into liquid transportation fuels. The selection and design of HDO catalysts plays an important role to determine the success of the process. This review has been aimed to emphasize recent developments on HDO catalysts in effective transformations of biomass-derived platform molecules into hydrocarbon fuels with reduced oxygen content and improved H/C ratios. Liquid hydrocarbon fuels can be obtained by combining oxygen removal processes (e.g. dehydration, hydrogenation, hydrogenolysis, decarbonylation etc.) as well as by increasing the molecular weight via C-C coupling reactions (e.g. aldol condensation, ketonization, oligomerization, hydroxyalkylation etc.). Fundamentals and mechanistic aspects of the use of HDO catalysts in deoxygenation reactions will also be discussed. PMID:25443804

  8. Progress on matrix SiC processing and properties for fully ceramic microencapsulated fuel form

    NASA Astrophysics Data System (ADS)

    Terrani, K. A.; Kiggans, J. O.; Silva, C. M.; Shih, C.; Katoh, Y.; Snead, L. L.

    2015-02-01

    The consolidation mechanism and resulting properties of the silicon carbide (SiC) matrix of fully ceramic microencapsulated (FCM) fuel form are discussed. The matrix is produced via the nano-infiltration transient eutectic-forming (NITE) process. Coefficient of thermal expansion, thermal conductivity, and strength characteristics of this SiC matrix have been characterized in the unirradiated state. An ad hoc methodology for estimation of thermal conductivity of the neutron-irradiated NITE-SiC matrix is also provided to aid fuel performance modeling efforts specific to this concept. Finally, specific processing methods developed for production of an optimal and reliable fuel form using this process are summarized. These various sections collectively report the progress made to date on production of optimal FCM fuel form to enable its application in light water and advanced reactors.

  9. Assessment of sulfur removal processes for advanced fuel cell systems

    NASA Astrophysics Data System (ADS)

    Lorton, G. A.

    1980-01-01

    The performance characteristics of potential sulfur removal processes were evaluated and four of these processes, the Selexol process, the Benfield process, the Sulfinol process, and the Rectisol process, were selected for detailed technical and economic comparison. The process designs were based on a consistent set of technical criteria for a grass roots facility with a capacity of 10,000 tons per day of Illinois No. 6 coal. Two raw gas compositions, based on oxygen blown and air blown Texaco gasification, were used. The bulk of the sulfur was removed in the sulfur removal unit, leaving a small amount of sulfur compounds in the gas. The remaining sulfur compounds were removed by reaction with zinc oxide in the sulfur polishing unit. The impact of COS hydrolysis pretreatment on sulfur removal was evaluated. Comprehensive capital and O and M cost estimates for each of the process schemes were developed.

  10. Multiphase CFD-based models for chemical looping combustion process: Fuel reactor modeling

    SciTech Connect

    Jung, Jonghwun; Gamwo, I.K.

    2008-04-21

    Chemical looping combustion (CLC) is a flameless two-step fuel combustion that produces a pure CO2 stream, ready for compression and sequestration. The process is composed of two interconnected fluidized bed reactors. The air reactor which is a conventional circulating fluidized bed and the fuel reactor which is a bubbling fluidized bed. The basic principle is to avoid the direct contact of air and fuel during the combustion by introducing a highly-reactive metal particle, referred to as oxygen carrier, to transport oxygen from the air to the fuel. In the process, the products from combustion are kept separated from the rest of the flue gases namely nitrogen and excess oxygen. This process eliminates the energy intensive step to separate the CO2 from nitrogen-rich flue gas that reduce the thermal efficiency. Fundamental knowledge of multiphase reactive fluid dynamic behavior of the gas–solid flow is essential for the optimization and operation of a chemical looping combustor. Our recent thorough literature review shows that multiphase CFD-based models have not been adapted to chemical looping combustion processes in the open literature. In this study, we have developed the reaction kinetics model of the fuel reactor and implemented the kinetic model into a multiphase hydrodynamic model, MFIX, developed earlier at the National Energy Technology Laboratory. Simulated fuel reactor flows revealed high weight fraction of unburned methane fuel in the flue gas along with CO2 and H2O. This behavior implies high fuel loss at the exit of the reactor and indicates the necessity to increase the residence time, say by decreasing the fuel flow rate, or to recirculate the unburned methane after condensing and removing CO2.

  11. Coal fueled ported kiln direct reduction process in Norway

    SciTech Connect

    Rierson, D.W.

    1994-12-31

    Allis Mineral Systems (AMS), formerly the minerals processing group at Allis-Chalmers Corporation, developed a ported kiln process in the 1960`s specifically for the direct reduction of iron ore. The process is called ACCAR. This ported kiln technology has more recently been coupled with AMS` GRATE-KILN System for iron oxide pelletizing into the GRATE-CAR Process, for minerals reduction. The GRATE-CAR Process can handle a fine grained ore concentrate through the steps of agglomeration, induration and reduction in a single production line.

  12. WATER REUSE IN A PAPER REPROCESSING PLANT

    EPA Science Inventory

    This project was undertaken to determine the feasibility of water reuse in a paper reprocessing plant with the goal being to 'close the loop' or to demonstrate zero discharge technology. Before the project began, Big Chief Roofing Company at Ardmore, OK, was discharging 7.89 1/se...

  13. Effects of reprocessing on nanoalumina polymer composites

    NASA Astrophysics Data System (ADS)

    Huang, Chunchia

    The life cycle of reprocessed polymer nanocomposites is a critical factor associated with their growing use, but the limited work on reprocessing of nanocomposites has focused solely on the effects of organoclays. This research investigated of the structure and property changes during reprocessing of polypropylene (PP) and polycarbonate (PC) nanocomposites with 3 wt. % nanoalumina. Neat PP and PC were used as controls. Reprocessing of the neat polymers and nanocomposites produced no indication of oxidation (in FTIR), no changes in the glass transition temperature of PC and the melting temperatures of PP, and no changes in thermal stability (as measured using thermogravimetric analysis). Significant decreases, however, occurred in the melt viscosity of the materials. The introduction of nanoalumina during twin screw extrusion also produced a significant decrease in the viscosity and a 10°C decrease in the glass transition temperature of the PC nanocomposite. Color changes did not correspond to the chain scission in PP and PC; neat PP and PP composite yellowed, neat PC turned brown, and the PC nanocomposite did not change color. Dispersion of the nanoalumina in both PP and PC improved with repeated reprocessing, the crystallinity in the PP/nanoalumina composites remained constant. The Young's moduli of both the PP and PP/A12O3 nanocomposite were similar, whereas the Young's modulus values of the PC/A1 2O3 nanocomposite was slighted lower than that of the neat PC. In contrast, the elongations at break of the PP/A12O3 and PC/A12O3 nanocomposites were, respectively, 50% and 16% of the values measured for the neat resins. All modulus and elongation at break values, however, remained constant over five reprocessing cycles. This behavior suggests that the major degradation mechanism during reprocessing of neat PP and PP nanocomposites was thermal-mechanical polymer chain scission and that the nanoalumina enhanced this degradation in the PC/nanoalumina. This additional

  14. Control of radio-iodine at the German reprocessing plant WAK during operation and after shutdown

    SciTech Connect

    Herrmann, F.J.; Herrmann, B.; Kuhn, K.D.

    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.

  15. Simulation of primary fuel atomization processes at subcritical pressures.

    SciTech Connect

    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 Poisson's 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.

  16. Toward mechanistic understanding of nuclear reprocessing chemistries by quantifying lanthanide solvent extraction kinetics via microfluidics with constant interfacial area and rapid mixing.

    PubMed

    Nichols, Kevin P; Pompano, Rebecca R; Li, Liang; Gelis, Artem V; Ismagilov, Rustem F

    2011-10-01

    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

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

    NASA Astrophysics Data System (ADS)

    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.

  18. Iodine Pathways and Off-Gas Stream Characteristics for Aqueous Reprocessing Plants – A Literature Survey and Assessment

    SciTech Connect

    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.

  19. Vision for the development of an international nuclear fuel recycling program

    SciTech Connect

    Kok, Kenneth D.

    2007-07-01

    The purpose of the development of an international nuclear fuel recycle program is to: - Demonstrate advanced recycling by working to prove the technologies needed to close the fuel cycle, minimize waste, and obtain more energy benefit for each unit of fuel. - Build a global vision by enlisting partners to limit the spread of sensitive nuclear technologies in a way that enables nuclear power to meet global challenges. The program will begin with the establishment of a smaller scale secure fuel cycle facility that would serve as a model for international nuclear fuel reprocessing centers that would eventually be built in several countries world wide. The operating process plants will provide the secure and safe guarded environment for the recycle of spent fuel from nuclear power stations around the world. The demonstration site will provide for developing and testing processes that would lead to the more complete use of the energy available in nuclear fuels and the minimization of long lived nuclear waste. (author)

  20. Dissolution process for ZrO.sub.2 -UO.sub.2 -CaO fuels

    DOEpatents

    Paige, Bernice E.

    1976-06-22

    The present invention provides an improved dissolution process for ZrO.sub.2 -UO.sub.2 -CaO-type pressurized water reactor fuels. The zirconium cladding is dissolved with hydrofluoric acid, immersing the ZrO.sub.2 -UO.sub.2 -CaO fuel wafers in the resulting zirconium-dissolver-product in the dissolver vessel, and nitric acid is added to the dissolver vessel to facilitate dissolution of the uranium from the ZrO.sub.2 -UO.sub.2 -CaO fuel wafers.