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Sample records for mixed oxide mox

  1. Mixed Oxide (MOX) Fuel Fabrication Facility | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration | (NNSA) fieldoffices / Savannah River Field Office Mixed Oxide (MOX) Fuel Fabrication Facility Documents related to the project: Plutonium Disposition Study Options Independent Assessment Phase 1 Report, April 13, 2015 Plutonium Disposition Study Options Independent Assessment Phase 2 Report, August 20, 2015 Final Report of the Plutonium Disposition Red Team, August 13, 2015 Commentary on Report by High Bridge Associates, Inc., Feb. 12, 2016 Related Topics Mixed Oxide Fuel

  2. An integrated approach for the verification of fresh mixed oxide fuel (MOX) assemblies at light water reactor MOX recycle reactors

    SciTech Connect (OSTI)

    Menlove, Howard O; Lee, Sang - Yoon

    2009-01-01

    This paper presents an integrated approach for the verification of mixed oxide (MOX) fuel assemblies prior to their being loaded into the reactor. There is a coupling of the verification approach that starts at the fuel fabrication plant and stops with the transfer of the assemblies into the thermal reactor. The key measurement points are at the output of the fuel fabrication plant, the receipt at the reactor site, and the storage in the water pool as fresh fuel. The IAEA currently has the capability to measure the MOX fuel assemblies at the output of the fuel fabrication plants using a passive neutron coincidence counting systems of the passive neutron collar (PNCL) type. Also. at the MOX reactor pool, the underwater coincidence counter (UWCC) has been developed to measure the MOX assemblies in the water. The UWCC measurement requires that the fuel assembly be lifted about two meters up in the storage rack to avoid interference from the fuel that is stored in the rack. This paper presents a new method to verify the MOX fuel assemblies that are in the storage rack without the necessity of moving the fuel. The detector system is called the Underwater MOX Verification System (UMVS). The integration and relationship of the three measurements systems is described.

  3. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Mertyurek, Ugur; Gauld, Ian C.

    2015-12-24

    In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

  4. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

    SciTech Connect (OSTI)

    Mertyurek, Ugur; Gauld, Ian C.

    2015-12-24

    In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

  5. Mixed Oxide Fuel Fabrication Facility

    National Nuclear Security Administration (NNSA)

    0%2A en Mixed Oxide (MOX) Fuel Fabrication Facility http:nnsa.energy.govfieldofficessavannah-river-field-officemixed-oxide-mox-fuel-fabrication-facility

  6. Monte-Carlo Code (MCNP) Modeling of the Advanced Test Reactor Applicable to the Mixed Oxide (MOX) Test Irradiation

    SciTech Connect (OSTI)

    G. S. Chang; R. C. Pederson

    2005-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, and 40 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energys Fissile Materials Disposition Program (FMDP). The fuel burnup analyses presented in this study were performed using MCWO, a welldeveloped tool that couples the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements.

  7. Remote-controlled NDA (nondestructive assay) systems for feed and product storage at an automated MOX (mixed oxide) facility

    SciTech Connect (OSTI)

    Menlove, H.O.; Augustson, R.H.; Ohtani, T.; Seya, M.; Takahashi, S.; Abedin-Zadeh, R.; Hassan, B.; Napoli, S.

    1989-01-01

    Nondestructive assay (NDA) systems have been developed for use in an automated mixed oxide (MOX) fabrication facility. Unique features have been developed for the NDA systems to accommodate robotic sample handling and remote operation. In addition, the systems have been designed to obtain International Atomic Energy Agency inspection data without the need for an inspector at the facility at the time of the measurements. The equipment is being designed to operate continuously in an unattended mode with data storage for periods of up to one month. The two systems described in this paper include a canister counter for the assay of MOX powder at the input to the facility and a capsule counter for the assay of complete liquid-metal fast breeder reactor fuel assemblies at the output of the plant. The design, performance characteristics, and authentication of the two systems will be described. The data related to reliability, precision, and stability will be presented. 5 refs., 10 figs., 4 tabs.

  8. Actual Scale MOX Powder Mixing Test for MOX Fuel Fabrication Plant in Japan

    SciTech Connect (OSTI)

    Osaka, Shuichi; Kurita, Ichiro; Deguchi, Morimoto; Ito, Masanori; Goto, Masakazu

    2007-07-01

    Japan Nuclear Fuel Ltd. (hereafter, JNFL) promotes a program of constructing a MOX fuel fabrication plant (hereafter, J-MOX) to fabricate MOX fuels to be loaded in domestic light water reactors. Since Japanese fiscal year (hereafter, JFY) 1999, JNFL, to establish the technology for a smooth start-up and the stable operation of J-MOX, has executed an evaluation test for technology to be adopted at J-MOX. JNFL, based on a consideration that J-MOX fuel fabrication comes commercial scale production, decided an introduction of MIMAS technology into J-MOX main process, from powder mixing through pellet sintering, well recognized as mostly important to achieve good quality product of MOX fuel, since it achieves good results in both fuel production and actual reactor irradiation in Europe, but there is one difference that JNFL is going to use Japanese typical plutonium and uranium mixed oxide powder converted with the micro-wave heating direct de-nitration technology (hereafter, MH-MOX) but normal PuO{sub 2} of European MOX fuel fabricators. Therefore, in order to evaluate the suitability of the MH-MOX powder for the MIMAS process, JNFL manufactured small scale test equipment, and implemented a powder mixing evaluation test up until JFY 2003. As a result, the suitability of the MH-MOX powder for the MIMAS process was positively evaluated and confirmed It was followed by a five-years test named an 'actual test' from JFY 2003 to JFY 2007, which aims at demonstrating good operation and maintenance of process equipment as well as obtaining good quality of MOX fuel pellets. (authors)

  9. Evaluation of existing United States` facilities for use as a mixed-oxide (MOX) fuel fabrication facility for plutonium disposition

    SciTech Connect (OSTI)

    Beard, C.A.; Buksa, J.J.; Chidester, K.; Eaton, S.L.; Motley, F.E.; Siebe, D.A.

    1995-12-31

    A number of existing US facilities were evaluated for use as a mixed-oxide fuel fabrication facility for plutonium disposition. These facilities include the Fuels Material Examination Facility (FMEF) at Hanford, the Washington Power Supply Unit 1 (WNP-1) facility at Hanford, the Barnwell Nuclear Fuel Plant (BNFP) at Barnwell, SC, the Fuel Processing Facility (FPF) at Idaho National Engineering Laboratory (INEL), the Device Assembly Facility (DAF) at the Nevada Test Site (NTS), and the P-reactor at the Savannah River Site (SRS). The study consisted of evaluating each facility in terms of available process space, available building support systems (i.e., HVAC, security systems, existing process equipment, etc.), available regional infrastructure (i.e., emergency response teams, protective force teams, available transportation routes, etc.), and ability to integrate the MOX fabrication process into the facility in an operationally-sound manner that requires a minimum amount of structural modifications.

  10. An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

    SciTech Connect (OSTI)

    Evans, Louise G; Croft, Stephen; Swinhoe, Martyn T; Tobin, S. J.; Boyer, B. D.; Menlove, H. O.; Schear, M. A.; Worrall, Andrew

    2010-11-24

    Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

  11. An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

    SciTech Connect (OSTI)

    Evans, Louise G; Croft, Stephen; Swinhoe, Martyn T; Tobin, S. J.; Menlove, H. O.; Schear, M. A.; Worrall, Andrew

    2011-01-13

    Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/ or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

  12. Environmental Excellence Program Recognizes MOX Services | National...

    National Nuclear Security Administration (NNSA)

    of NNSA's Mixed Oxide (MOX) Fuel Fabrication Facility at the Savannah River Site. ... The MOX Fuel Fabrication Facility, currently under construction at the Savannah River Site ...

  13. MOX

    National Nuclear Security Administration (NNSA)

    as MOX fuel will be significantly more expensive than anticipated. Given a lifecycle cost estimate for the program of approximately 30 billion or more and a challenging budget...

  14. All About MOX

    ScienceCinema (OSTI)

    None

    2014-08-06

    In 1999, the Nuclear Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

  15. NNSA Holds Groundbreaking at MOX Facility | National Nuclear...

    National Nuclear Security Administration (NNSA)

    NNSA's plutonium disposition program moved another step forward with the start of site preparation for its Mixed Oxide (MOX) Fuel Fabrication Facility at the Savannah River Site. ...

  16. Mixed Oxide Fuel Fabrication Facility | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration Mixed Oxide Fuel Fabrication Facility Mixed Oxide (MOX) Fuel Fabrication Facility Documents related to the project: Plutonium Disposition Study Options Independent Assessment Phase 1 Report, April 13, 2015 Plutonium Disposition Study Options Independent Assessment Phase 2 Report, August 20, 2015 Final Report of the Plutonium Disposition Red Team, August 13, 2015 Commentary on

  17. MOX Two-Year Construction Anniversary | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration | (NNSA) Two-Year Construction Anniversary MOX Two-Year Construction Anniversary MOX Two-Year Construction Anniversary In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus

  18. MOX | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    MOX Mixed Oxide (MOX) Fuel Fabrication Facility Documents related to the project: Plutonium Disposition Study Options Independent Assessment Phase 1 Report, April 13, 2015 Plutonium Disposition Study Options Independent Assessment Phase 2 Report, August 20, 2015 Final Report of the Plutonium Disposition Red Team, August 13, 2015 Commentary on... Analysis of Surplus Weapons-Grade Plutonium Disposition Options The Administration remains firmly committed to disposing of surplus weapon-grade

  19. NNSA B-Roll: MOX Facility

    SciTech Connect (OSTI)

    2010-05-21

    In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

  20. NNSA B-Roll: MOX Facility

    ScienceCinema (OSTI)

    None

    2010-09-01

    In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

  1. MOX Three-Year Construction Anniversary | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration | (NNSA) Three-Year Construction Anniversary August 1, 2010 marks the third year of successful construction at the Mixed Oxide (MOX) Fuel Fabrication Facility at the Savannah River Site near Aiken, SC. Shaw AREVA MOX Services, LLC is under contract with the National Nuclear Security Administration (NNSA) to design, build, and operate MOX Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium.

  2. Mixed oxide solid solutions

    DOE Patents [OSTI]

    Magno, Scott; Wang, Ruiping; Derouane, Eric

    2003-01-01

    The present invention is a mixed oxide solid solution containing a tetravalent and a pentavalent cation that can be used as a support for a metal combustion catalyst. The invention is furthermore a combustion catalyst containing the mixed oxide solid solution and a method of making the mixed oxide solid solution. The tetravalent cation is zirconium(+4), hafnium(+4) or thorium(+4). In one embodiment, the pentavalent cation is tantalum(+5), niobium(+5) or bismuth(+5). Mixed oxide solid solutions of the present invention exhibit enhanced thermal stability, maintaining relatively high surface areas at high temperatures in the presence of water vapor.

  3. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    SciTech Connect (OSTI)

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  4. MOX and MOX with 237Np/241Am Inert Fission Gas Generation Comparison in ATR

    SciTech Connect (OSTI)

    G. S. Chang; M. Robel; W. J. Carmack; D. J. Utterbeck

    2006-06-01

    The treatment of spent fuel produced in nuclear power generation is one of the most important issues to both the nuclear community and the general public. One of the viable options to long-term geological disposal of spent fuel is to extract plutonium, minor actinides (MA), and potentially long-lived fission products from the spent fuel and transmute them into short-lived or stable radionuclides in currently operating light-water reactors (LWR), thus reducing the radiological toxicity of the nuclear waste stream. One of the challenges is to demonstrate that the burnup-dependent characteristic differences between Reactor-Grade Mixed Oxide (RG-MOX) fuel and RG-MOX fuel with MA Np-237 and Am 241 are minimal, particularly, the inert gas generation rate, such that the commercial MOX fuel experience base is applicable. Under the Advanced Fuel Cycle Initiative (AFCI), developmental fuel specimens in experimental assembly LWR-2 are being tested in the northwest (NW) I-24 irradiation position of the Advanced Test Reactor (ATR). The experiment uses MOX fuel test hardware, and contains capsules with MOX fuel consisting of mixed oxide manufactured fuel using reactor grade plutonium (RG-Pu) and mixed oxide manufactured fuel using RG-Pu with added Np/Am. This study will compare the fuel neutronics depletion characteristics of Case-1 RG-MOX and Case-2 RG-MOX with Np/Am.

  5. Mixed Acid Oxidation

    SciTech Connect (OSTI)

    Pierce, R.A.

    1999-10-26

    Several non-thermal processes have been developed to destroy organic waste compounds using chemicals with high oxidation potentials. These efforts have focused on developing technologies that work at low temperatures, relative to incineration, to overcome many of the regulatory issues associated with obtaining permits for waste incinerators. One such technique with great flexibility is mixed acid oxidation. Mixed acid oxidation, developed at the Savannah River Site, uses a mixture of an oxidant (nitric acid) and a carrier acid (phosphoric acid). The carrier acid acts as a non-volatile holding medium for the somewhat volatile oxidant. The combination of acids allows appreciable amounts of the concentrated oxidant to remain in the carrier acid well above the oxidant''s normal boiling point.

  6. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    SciTech Connect (OSTI)

    Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S.

    1995-09-01

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO{sub 2}) mixed with urania (UO{sub 2}). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified.

  7. The manufacture and performance of homogeneous microstructure SBR MOX fuel

    SciTech Connect (OSTI)

    Barker, Matthew A.; Stephenson, Keith; Weston, Rebecca

    2007-07-01

    In the early 1980's, British experience in the manufacture of mixed-oxide fast reactor fuel was used to develop a new thermal MOX manufacturing route called the Short Binder-less Route (SBR). Laboratory- scale development led to the manufacture of commercial PWR fuel in a small pilot plant, and the construction of the full-scale dual-line Sellafield MOX Plant (SMP). SMP's first MOX assemblies are now under irradiation. SBR MOX is manufactured with 100% co-milled feedstock, leading to a microstructure dominated by a solid solution of (U,Pu)O{sub 2} at the nominal enrichment. A comprehensive fuel performance research programme has demonstrated the benign performance of SBR MOX up to 54 MWd/kgHM. In particular, the homogeneous microstructure is believed to be instrumental in the favourable fission gas retention and PCI resistance properties. (authors)

  8. Mixed Oxide (MOX) Fuel Fabrication Facility | National Nuclear...

    National Nuclear Security Administration (NNSA)

    Assessment Phase 2 Report, August 20, 2015 Final Report of the Plutonium Disposition Red Team, August 13, 2015 Commentary on Report by High Bridge Associates, Inc., Feb. 12, ...

  9. Energy Secretary Bodman Commends Key Milestone In MOX Program | Department

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    of Energy Commends Key Milestone In MOX Program Energy Secretary Bodman Commends Key Milestone In MOX Program April 1, 2005 - 11:28am Addthis WASHINGTON, DC - In response to the Nuclear Regulatory Commission's (NRC) authorization of the construction of a U.S. Mixed-Oxide (MOX) Fuel Fabrication Facility at the Department of Energy's Savannah River Site in South Carolina, Secretary of Energy Samuel W. Bodman today released the following statement: "Issuing the permit for construction of a

  10. MOX fuel arrangement for nuclear core

    DOE Patents [OSTI]

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-07-17

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  11. Mox fuel arrangement for nuclear core

    DOE Patents [OSTI]

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-05-15

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

  12. MOX fuel arrangement for nuclear core

    DOE Patents [OSTI]

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    1998-01-01

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  13. MOX fuel arrangement for nuclear core

    DOE Patents [OSTI]

    Kantrowitz, M.L.; Rosenstein, R.G.

    1998-10-13

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

  14. MOX Cross-Section Libraries for ORIGEN-ARP

    SciTech Connect (OSTI)

    Gauld, I.C.

    2003-07-01

    The use of mixed-oxide (MOX) fuel in commercial nuclear power reactors operated in Europe has expanded rapidly over the past decade. The predicted characteristics of MOX fuel such as the nuclide inventories, thermal power from decay heat, and radiation sources are required for design and safety evaluations, and can provide valuable information for non-destructive safeguards verification activities. This report describes the development of computational methods and cross-section libraries suitable for the analysis of irradiated MOX fuel with the widely-used and recognized ORIGEN-ARP isotope generation and depletion code of the SCALE (Standardized Computer Analyses for Licensing Evaluation) code system. The MOX libraries are designed to be used with the Automatic Rapid Processing (ARP) module of SCALE that interpolates appropriate values of the cross sections from a database of parameterized cross-section libraries to create a problem-dependent library for the burnup analysis. The methods in ORIGEN-ARP, originally designed for uranium-based fuels only, have been significantly upgraded to handle the larger number of interpolation parameters associated with MOX fuels. The new methods have been incorporated in a new version of the ARP code that can generate libraries for low-enriched uranium (LEU) and MOX fuel types. The MOX data libraries and interpolation algorithms in ORIGEN-ARP have been verified using a database of declared isotopic concentrations for 1042 European MOX fuel assemblies. The methods and data are validated using a numerical MOX fuel benchmark established by the Organization for Economic Cooperation and Development (OECD) Working Group on burnup credit and nuclide assay measurements for irradiated MOX fuel performed as part of the Belgonucleaire ARIANE International Program.

  15. Mixed oxide fuel development

    SciTech Connect (OSTI)

    Leggett, R.D.; Omberg, R.P.

    1987-05-08

    This paper describes the success of the ongoing mixed-oxide fuel development program in the United States aimed at qualifying an economical fuel system for liquid metal cooled reactors. This development has been the cornerstone of the US program for the past 20 years and has proceeded in a deliberate and highly disciplined fashion with high emphasis on fuel reliability and operational safety as major features of an economical fuel system. The program progresses from feature testing in EBR-II to qualifying full size components in FFTF under fully prototypic conditions to establish a basis for extending allowable lifetimes. The development program started with the one year (300 EFPD) core, which is the FFTF driver fuel, continued with the demonstration of a two year (600 EFPD) core and is presently evaluating a three year (900 EFPD) fuel system. All three of these systems, consistent with other LMR fuel programs around the world, use fuel pellets gas bonded to a cladding tube that is assembled into a bundle and fitted into a wrapper tube or duct for ease of insertion into a core. The materials of construction progressed from austenitic CW 316 SS to lower swelling austenitic D9 to non swelling ferritic/martensitic HT9. 6 figs., 2 tabs.

  16. Development of a fresh MOX fuel transport package for disposition of weapons plutonium

    SciTech Connect (OSTI)

    Ludwig, S.B.; Pope, R.B.; Shappert, L.B.; Michelhaugh, R.D.; Chae, S.M.

    1998-11-01

    The US Department of Energy announced its Record of Decision on January 14, 1997, to embark on a dual-track approach for disposition of surplus weapons-usable plutonium using immobilization in glass or ceramics and burning plutonium as mixed-oxide (MOX) fuel in reactors. In support of the MOX fuel alternative, Oak Ridge National Laboratory initiated development of conceptual designs for a new package for transporting fresh (unirradiated) MOX fuel assemblies between the MOX fabrication facility and existing commercial light-water reactors in the US. This paper summarizes progress made in development of new MOX transport package conceptual designs. The development effort has included documentation of programmatic and technical requirements for the new package and development and analysis of conceptual designs that satisfy these requirements.

  17. NNSA Marks Two-Year Construction Milestone at MOX Facility in South

    National Nuclear Security Administration (NNSA)

    Carolina | National Nuclear Security Administration | (NNSA) Marks Two-Year Construction Milestone at MOX Facility in South Carolina July 31, 2009 WASHINGTON, D.C. - The National Nuclear Security Administration (NNSA) today marked the second full year of successful construction of the Mixed Oxide (MOX) Fuel Fabrication Facility at the Savannah River Site (SRS) near Aiken, South Carolina, by launching a new online multimedia kit on NNSA's website. The kit, which includes a photo gallery that

  18. Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

    SciTech Connect (OSTI)

    Carbajo, J.J.

    2005-05-27

    The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.

  19. APPLICATION OF COLUMN EXTRACTION METHOD FOR IMPURITIES ANALYSIS ON HB-LINE PLUTONIUM OXIDE IN SUPPORT OF MOX FEED PRODUCT SPECIFICATIONS

    SciTech Connect (OSTI)

    Jones, M.; Diprete, D.; Wiedenman, B.

    2012-03-20

    The current mission at H-Canyon involves the dissolution of an Alternate Feedstocks 2 (AFS-2) inventory that contains plutonium metal. Once dissolved, HB-Line is tasked with purifying the plutonium solution via anion exchange, precipitating the Pu as oxalate, and calcining to form plutonium oxide (PuO{sub 2}). The PuO{sub 2} will provide feed product for the Mixed Oxide (MOX) Fuel Fabrication Facility, and the anion exchange raffinate will be transferred to H-Canyon. The results presented in this report document the potential success of the RE resin column extraction application on highly concentrated Pu samples to meet MOX feed product specifications. The original 'Hearts Cut' sample required a 10000x dilution to limit instrument drift on the ICP-MS method. The instrument dilution factors improved to 125x and 250x for the sample raffinate and sample eluent, respectively. As noted in the introduction, the significantly lower dilutions help to drop the total MRL for the analyte. Although the spike recoveries were half of expected in the eluent for several key elements, they were between 94-98% after Nd tracer correction. It is seen that the lower ICD limit requirements for the rare earths are attainable because of less dilution. Especially important is the extremely low Ga limit at 0.12 {mu}g/g Pu; an ICP-MS method is now available to accomplish this task on the sample raffinate. While B and V meet the column A limits, further development is needed to meet the column B limits. Even though V remained on the RE resin column, an analysis method is ready for investigation on the ICP-MS, but it does not mean that V cannot be measured on the ICP-ES at a low dilution to meet the column B limits. Furthermore, this column method can be applicable for ICP-ES as shown in Table 3-2, in that it trims the sample of Pu, decreasing and sometimes eliminating Pu spectral interferences.

  20. The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

    SciTech Connect (OSTI)

    G. W. Eccleston; H. O. Menlove; M. Abhold; M. Baker; J. Pecos

    1999-05-01

    This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency {sup 3}He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the {sup 240}Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual.

  1. Examination of Risk Analysis Methods for MOX Land Transport in Japan

    SciTech Connect (OSTI)

    HOHNSTREITER, GLENN FREDRICK; PIERCE, JIM D.

    2003-04-01

    This report presents background information and methodology for a risk assessment of mixed oxide (MOX) reactor fuel transport in the nation of Japan to support their nuclear energy program. This work includes an extensive literature review, a review of other MOX activities worldwide, a survey of the statutory requirements for transporting nuclear materials, a discussion of risk assessment methodology, and calculation results for specific examples. Typical risk evaluations are given to provide guidance for later risk analyses specific to MOX fuel transport in Japan. This report also includes specific information that will be required for routes, cask types, accident-rate statistics, and population densities along specified routes, along with other detailed information needed for risk analysis studies pertinent to MOX transport in Japan. This information will be used in future specific risk studies.

  2. Neutron Emission Characteristics of Two Mixed-Oxide Fuels: Simulations and Initial Experiments

    SciTech Connect (OSTI)

    D. L. Chichester; S. A. Pozzi; J. L. Dolan; M. Flaska; J. T. Johnson; E. H. Seabury; E. M. Gantz

    2009-07-01

    Simulations and experiments have been carried out to investigate the neutron emission characteristics of two mixed-oxide (MOX) fuels at Idaho National Laboratory (INL). These activities are part of a project studying advanced instrumentation techniques in support of the U.S. Department of Energy's Fuel Cycle Research and Development program and it's Materials Protection, Accounting, and Control for Transmutation (MPACT) campaign. This analysis used the MCNP-PoliMi Monte Carlo simulation tool to determine the relative strength and energy spectra of the different neutron source terms within these fuels, and then used this data to simulate the detection and measurement of these emissions using an array of liquid scintillator neutron spectrometers. These calculations accounted for neutrons generated from the spontaneous fission of the actinides in the MOX fuel as well as neutrons created via (alpha,n) reactions with oxygen in the MOX fuel. The analysis was carried out to allow for characterization of both neutron energy as well as neutron coincidences between multiple detectors. Coincidences between prompt gamma rays and neutrons were also analyzed. Experiments were performed at INL with the same materials used in the simulations to benchmark and begin validation tests of the simulations. Data was collected in these experiments using an array of four liquid scintillators and a high-speed waveform digitizer. Advanced digital pulse-shape discrimination algorithms were developed and used to collect this data. Results of the simulation and modeling studies are presented together with preliminary results from the experimental campaign.

  3. Americium characterization by X-ray fluorescence and absorption spectroscopy in plutonium uranium mixed oxide

    SciTech Connect (OSTI)

    Degueldre, Claude Cozzo, Cedric; Martin, Matthias; Grolimund, Daniel; Mieszczynski, Cyprian

    2013-06-01

    Plutonium uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regard to their environment and the coolant. In this work the study of the atomic structure and next-neighbour environment of Am in the (Pu,U)O? lattice in an irradiated (60 MW d kg?) MOX sample was performed employing micro-X-ray fluorescence (-XRF) and micro-X-ray absorption fine structure (-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Am (~0.66 wt%) are determined from the experimental data gained for the irradiated fuel material examined in its peripheral zone (rim) of the fuel. In the irradiated sample Am builds up as Am? species within an [AmO?]? coordination environment (e.g. >90%) and no (<10%) Am(IV) or (V) can be detected in the rim zone. The occurrence of americium dioxide is avoided by the redox buffering activity of the uranium dioxide matrix. - Graphical abstract: Americium LIII XAFS spectra recorded for the irradiated MOX sub-sample in the rim zone for a 300 ?m300 ?m beam size area investigated over six scans of 4 h. The records remain constant during multi-scan. The analysis of the XAFS signal shows that Am is found as trivalent in the UO? matrix. This analytical work shall open the door of very challenging analysis (speciation of fission product and actinides) in irradiated nuclear fuels. - Highlights: Americium was characterized by microX-ray absorption spectroscopy in irradiated MOX fuel. The americium redox state as determined from XAS data of irradiated fuel material was Am(III). In the sample, the Am? face an AmO??coordination environment in the (Pu,U)O? matrix. The americium dioxide is reduced by the uranium dioxide matrix.

  4. ANALYSIS AND EXAMINATION OF MOX FUEL FROM NONPROLIFERATION PROGRAMS

    SciTech Connect (OSTI)

    McCoy, Kevin; Machut, Dr McLean; Morris, Robert Noel; Blanpain, Patrick; Hemrick, James Gordon

    2013-01-01

    The U.S. Department of Energy has decided to dispose of a portion of the nation s surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. Four lead assemblies were manufactured and irradiated to a maximum fuel rod burnup of 47.3 MWd/kg heavy metal. This was the first commercial irradiation of MOX fuel with a 240Pu/239Pu ratio of less than 0.10. Five fuel rods with varying burnups and plutonium contents were selected from one of the assemblies and shipped to Oak Ridge National Laboratory for hot cell examination. The performance of the rods was analyzed with AREVA s next-generation GALILEO code. The results of the analysis confirmed that the fuel rods had performed safely and predictably, and that GALILEO is applicable to MOX fuel with a low 240Pu/239Pu ratio as well as to standard MOX. The results are presented and compared to the GALILEO database. In addition, the fuel cladding was tested to confirm that traces of gallium in the fuel pellets had not affected the mechanical properties of the cladding. The irradiated cladding was found to remain ductile at both room temperature and 350 C for both the axial and circumferential directions.

  5. Benchmark of SCALE (SAS2H) isotopic predictions of depletion analyses for San Onofre PWR MOX fuel

    SciTech Connect (OSTI)

    Hermann, O.W.

    2000-02-01

    The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, after discharge from reactors is of significant interest to the Fissile Materials Disposition Program. The validation of the SCALE (SAS2H) depletion code for use in the prediction of isotopic compositions of MOX fuel, similar to previous validation studies on uranium-only fueled reactors, has corresponding significance. The EEI-Westinghouse Plutonium Recycle Demonstration Program examined the use of MOX fuel in the San Onofre PWR, Unit 1, during cycles 2 and 3. Isotopic analyses of the MOX spent fuel were conducted on 13 actinides and {sup 148}Nd by either mass or alpha spectrometry. Six fuel pellet samples were taken from four different fuel pins of an irradiated MOX assembly. The measured actinide inventories from those samples has been used to benchmark SAS2H for MOX fuel applications. The average percentage differences in the code results compared with the measurement were {minus}0.9% for {sup 235}U and 5.2% for {sup 239}Pu. The differences for most of the isotopes were significantly larger than in the cases for uranium-only fueled reactors. In general, comparisons of code results with alpha spectrometer data had extreme differences, although the differences in the calculations compared with mass spectrometer analyses were not extremely larger than that of uranium-only fueled reactors. This benchmark study should be useful in estimating uncertainties of inventory, criticality and dose calculations of MOX spent fuel.

  6. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    SciTech Connect (OSTI)

    Ade, Brian J; Gauld, Ian C

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

  7. Mixed-oxide fuel decay heat analysis for BWR LOCA safety evaluation

    SciTech Connect (OSTI)

    Chiang, R. T.

    2013-07-01

    The mixed-oxide (MOX) fuel decay heat behavior is analyzed for Boiling Water Reactor (BWR) Loss of Coolant Accident (LOCA) safety evaluation. The physical reasoning on why the decay heat power fractions of MOX fuel fission product (FP) are significantly lower than the corresponding decay heat power fractions of uranium-oxide (UOX) fuel FP is illustrated. This is primarily due to the following physical phenomena. -The recoverable energies per fission of plutonium (Pu)-239 and Pu-241 are significantly higher than those of uranium (U)-235 and U-238. Consequently, the fission rate required to produce the same amount of power in MOX fuel is significantly lower than that in UOX fuel, which leads to lower subsequent FP generation rate and associated decay heat power in MOX fuel than those in UOX fuel. - The effective FP decay energy per fission of Pu-239 is significantly lower than the corresponding effective FP decay energy per fission of U-235, e.g., Pu-239's 10.63 Mega-electron-Volt (MeV) vs. U-235's 12.81 MeV at the cooling time 0.2 second. This also leads to lower decay heat power in MOX fuel than that in UOX fuel. The FP decay heat is shown to account for more than 90% of the total decay heat immediately after shutdown. The FP decay heat results based on the American National Standard Institute (ANSI)/American Nuclear Society (ANS)-5.1-1979 standard method are shown very close to the corresponding FP decay heat results based on the ANSI/ANS-5.1-2005 standard method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method are shown very close to but mostly slightly lower than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1971 method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method or the ANSI/ANS-5.1-1971 method are shown significantly larger than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1979 standard method or the ANSI/ANS-5.1-2005 standard method. (authors)

  8. Mixed oxide nanoparticles and method of making

    DOE Patents [OSTI]

    Lauf, Robert J.; Phelps, Tommy J.; Zhang, Chuanlun; Roh, Yul

    2002-09-03

    Methods and apparatus for producing mixed oxide nanoparticulates are disclosed. Selected thermophilic bacteria cultured with suitable reducible metals in the presence of an electron donor may be cultured under conditions that reduce at least one metal to form a doped crystal or mixed oxide composition. The bacteria will form nanoparticles outside the cell, allowing easy recovery. Selection of metals depends on the redox potentials of the reducing agents added to the culture. Typically hydrogen or glucose are used as electron donors.

  9. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect (OSTI)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

  10. Performance of Cladding on MOX Fuel with Low 240Pu/239Pu Ratio

    SciTech Connect (OSTI)

    McCoy, Kevin; Blanpain, Patrick; Morris, Robert Noel

    2014-01-01

    The U.S. Department of Energy has decided to dispose of a portion of its surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. As part of fuel qualification, four lead assemblies were manufactured and irradiated to a maximum fuel rod average burnup of 47.3 MWd/kg heavy metal. This was the world s first commercial irradiation of MOX fuel with a 240Pu/239Pu ratio less than 0.10. Five fuel rods with varying burnups and plutonium contents were selected from one of the assemblies and shipped to Oak Ridge National Laboratory for hot cell examination. This paper discusses the results of those examinations with emphasis on cladding performance. Exams relevant to the cladding included visual and eddy current exams, profilometry, microscopy, hydrogen analysis, gallium analysis, and mechanical testing. There was no discernible effect of the type of MOX fuel on the performance of the cladding.

  11. MOX Lead Assembly Fabrication at the Savannah River Site

    SciTech Connect (OSTI)

    Geddes, R.L.; Spiker, D.L.; Poon, A.P.

    1997-12-01

    The U. S. Department of Energy (DOE) announced its intent to prepare an Environmental Impact Statement (EIS) under the National Environmental Policy Act (NEPA) on the disposition of the nations weapon-usable surplus plutonium.This EIS is tiered from the Storage and Disposition of Weapons-Usable Fissile Material Programmatic Environmental Impact Statement issued in December 1996,and the associated Record of Decision issued on January, 1997. The EIS will examine reasonable alternatives and potential environmental impacts for the proposed siting, construction, and operation of three types of facilities for plutonium disposition. The three types of facilities are: a pit disassembly and conversion facility, a facility to immobilize surplus plutonium in a glass or ceramic form for disposition, and a facility to fabricate plutonium oxide into mixed oxide (MOX) fuel.As an integral part of the surplus plutonium program, Oak Ridge National Laboratory (ORNL) was tasked by the DOE Office of Fissile Material Disposition(MD) as the technical lead to organize and evaluate existing facilities in the DOE complex which may meet MD`s need for a domestic MOX fuel fabrication demonstration facility. The Lead Assembly (LA) facility is to produce 1 MT of usable test fuel per year for three years. The Savannah River Site (SRS) as the only operating plutonium processing site in the DOE complex, proposes two options to carry out the fabrication of MOX fuel lead test assemblies: an all Category I facility option and a combined Category I and non-Category I facilities option.

  12. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    SciTech Connect (OSTI)

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O`Connor, D.G.; Carrell, R.D.; Jaeger, C.D.; Thompson, M.L.; Strasser, A.A.

    1998-03-01

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET.

  13. Modeling of the performance of weapons MOX fuel in light water reactors

    SciTech Connect (OSTI)

    Alvis, J.; Bellanger, P.; Medvedev, P.G.; Peddicord, K.L.; Gellene, G.I.

    1999-05-01

    Both the Russian Federation and the US are pursing mixed uranium-plutonium oxide (MOX) fuel in light water reactors (LWRs) for the disposition of excess plutonium from disassembled nuclear warheads. Fuel performance models are used which describe the behavior of MOX fuel during irradiation under typical power reactor conditions. The objective of this project is to perform the analysis of the thermal, mechanical, and chemical behavior of weapons MOX fuel pins under LWR conditions. If fuel performance analysis indicates potential questions, it then becomes imperative to assess the fuel pin design and the proposed operating strategies to reduce the probability of clad failure and the associated release of radioactive fission products into the primary coolant system. Applying the updated code to anticipated fuel and reactor designs, which would be used for weapons MOX fuel in the US, and analyzing the performance of the WWER-100 fuel for Russian weapons plutonium disposition are addressed in this report. The COMETHE code was found to do an excellent job in predicting fuel central temperatures. Also, despite minor predicted differences in thermo-mechanical behavior of MOX and UO{sub 2} fuels, the preliminary estimate indicated that, during normal reactor operations, these deviations remained within limits foreseen by fuel pin design.

  14. Evaluation of Co-precipitation Processes for the Synthesis of Mixed-Oxide Fuel Feedstock Materials

    SciTech Connect (OSTI)

    Collins, Emory D; Voit, Stewart L; Vedder, Raymond James

    2011-06-01

    The focus of this report is the evaluation of various co-precipitation processes for use in the synthesis of mixed oxide feedstock powders for the Ceramic Fuels Technology Area within the Fuels Cycle R&D (FCR&D) Program's Advanced Fuels Campaign. The evaluation will include a comparison with standard mechanical mixing of dry powders and as well as other co-conversion methods. The end result will be the down selection of a preferred sequence of co-precipitation process for the preparation of nuclear fuel feedstock materials to be used for comparison with other feedstock preparation methods. A review of the literature was done to identify potential nitrate-to-oxide co-conversion processes which have been applied to mixtures of uranium and plutonium to achieve recycle fuel homogeneity. Recent studies have begun to study the options for co-converting all of the plutonium and neptunium recovered from used nuclear fuels, together with appropriate portions of recovered uranium to produce the desired mixed oxide recycle fuel. The addition of recycled uranium will help reduce the safeguard attractiveness level and improve proliferation resistance of the recycled fuel. The inclusion of neptunium is primarily driven by its chemical similarity to plutonium, thus enabling a simple quick path to recycle. For recycle fuel to thermal-spectrum light water reactors (LWRs), the uranium concentration can be {approx}90% (wt.), and for fast spectrum reactors, the uranium concentration can typically exceed 70% (wt.). However, some of the co-conversion/recycle fuel fabrication processes being developed utilize a two-step process to reach the desired uranium concentration. In these processes, a 50-50 'master-mix' MOX powder is produced by the co-conversion process, and the uranium concentration is adjusted to the desired level for MOX fuel recycle by powder blending (milling) the 'master-mix' with depleted uranium oxide. In general, parameters that must be controlled for co

  15. Process for etching mixed metal oxides

    DOE Patents [OSTI]

    Ashby, C.I.H.; Ginley, D.S.

    1994-10-18

    An etching process is described using dicarboxylic and tricarboxylic acids as chelating etchants for mixed metal oxide films such as high temperature superconductors and ferroelectric materials. Undesirable differential etching rates between different metal oxides are avoided by selection of the proper acid or combination of acids. Feature sizes below one micron, excellent quality vertical edges, and film thicknesses in the 100 Angstrom range may be achieved by this method. 1 fig.

  16. Process for etching mixed metal oxides

    DOE Patents [OSTI]

    Ashby, Carol I. H.; Ginley, David S.

    1994-01-01

    An etching process using dicarboxylic and tricarboxylic acids as chelating etchants for mixed metal oxide films such as high temperature superconductors and ferroelectric materials. Undesirable differential etching rates between different metal oxides are avoided by selection of the proper acid or combination of acids. Feature sizes below one micron, excellent quality vertical edges, and film thicknesses in the 100 Angstom range may be achieved by this method.

  17. Improved mixed oxide fuel calculations with the evaluated nuclear data library JEFF-3.2

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Noguere, G.; Bernard, D.; Blaise, P.; Bouland, O.; Leal, Luiz C.; Leconte, P.; Litaize, O.; Peneliau, Y.; Roque, B.; Santamarina, A.; et al

    2016-02-01

    In this study, an overestimation of the keff values for mixed oxide (MOX) fuels was identified with Monte Carlo (TRIPOLI-4) and deterministic (APOLLO2) calculations based on the Joint Evaluated Fission and Fusion (JEFF) evaluated nuclear data library. The overestimation becomes sizeable with Pit aging, reaching a reactivity change of Delta(p)similar or equal to+700 pcm for integral measurements carried out with MOX fuel containing a large amount of americium. This bias was observed for various critical configurations performed in the zero power reactor EOLE of the Commissariat a l'energie atomique et aux energies alternatives (CEA), Cadarache, France. The present work focusesmore » on the improvements achieved with the new 239PU and 241Am evaluated nuclear data files available in the latest version of the JEFF library (JEFF-3.2). The resolved resonance range of the plutonium evaluation was reevaluated at Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee, with the Ski/NH code in collaboration with CEA Cadarache. The resonance parameters of the americium evaluation were obtained with the REFIT code in collaboration with the research institutes Institute for Reference Materials and Measurements aRmm, Geel, Belgium, and Institut de recherche sur les lois fondamentales de l'Univers ofio, Saclay, France.« less

  18. Evaluation of Internal Criticality of the Plutonium Dispostion MOX SNF Waste Form

    SciTech Connect (OSTI)

    A.A. Alsaed

    1999-09-28

    The purpose of this calculation is to perform a parametric study to determine the effects of fission product leaching, assembly collapse, and iron oxide loss ({Delta}Fe{sub 2}O{sub 3}) on the reactivity of a waste package (WP) containing mixed oxide (MOX) spent nuclear fuel (SNF). Previous calculations (CRWMS M&O 1998a) have shown that the criticality control features of the WP are adequate to prevent criticality of a flooded WP for all the enrichment/burnup pairs expected for the MOX SNF. Therefore, the objective of this calculation is to determine the increase in reactivity that might result from possible degradation of the WP criticality control features. Specifically, this calculation tests the sensitivity of effective neutron multiplication factor (k{sub eff}) to loss (from the WP) of the following: (1) fission product neutron absorbers, or (2) moderator displacement material (principally, the iron oxide that results from the corrosion of carbon steel).

  19. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Ross, Kyle W.; Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  20. Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

    SciTech Connect (OSTI)

    Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

    2002-02-26

    In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is

  1. A Validation Study of Pin Heat Transfer for MOX Fuel Based on the IFA-597 Experiments

    SciTech Connect (OSTI)

    Phillippe, Aaron M; Clarno, Kevin T; Banfield, James E; Ott, Larry J; Philip, Bobby; Berrill, Mark A; Sampath, Rahul S; Allu, Srikanth; Hamilton, Steven P

    2014-01-01

    Abstract The IFA-597 (Integrated Fuel Assessment) experiments from the International Fuel Performance Experiments (IFPE) database were designed to study the thermal behavior of mixed oxide (MOX) fuel and the effects of an annulus on fission gas release in light-water-reactor fuel. An evaluation of nuclear fuel pin heat transfer in the FRAPCON-3.4 and Exnihilo codes for MOX fuel systems was performed, with a focus on the first 20 time steps ( 6 GWd/MT(iHM)) for explicit comparison between the codes. In addition, sensitivity studies were performed to evaluate the effect of the radial power shape and approximations to the geometry to account for the thermocouple hole, dish, and chamfer. The analysis demonstrated relative agreement for both solid (rod 1) and annular (rod 2) fuel in the experiment, demonstrating the accuracy of the codes and their underlying material models for MOX fuel, while also revealing a small energy loss artifact in how gap conductance is currently handled in Exnihilo for chamfered fuel pellets. The within-pellet power shape was shown to significantly impact the predicted centerline temperatures. This has provided an initial benchmarking of the pin heat transfer capability of Exnihilo for MOX fuel with respect to a well-validated nuclear fuel performance code.

  2. Investigation of Mixed Oxide Catalysts for NO Oxidation

    SciTech Connect (OSTI)

    Szanyi, Janos; Karim, Ayman M.; Pederson, Larry R.; Kwak, Ja Hun; Mei, Donghai; Tran, Diana N.; Herling, Darrell R.; Muntean, George G.; Peden, Charles HF; Howden, Ken; Qi, Gongshin; Li, Wei

    2014-12-09

    The oxidation of engine-generated NO to NO2 is an important step in the reduction of NOx in lean engine exhaust because NO2 is required for the performance of the LNT technology [2], and it enhances the activities of ammonia selective catalytic reduction (SCR) catalysts [1]. In particular, for SCR catalysts an NO:NO2 ratio of 1:1 is most effective for NOx reduction, whereas for LNT catalysts, NO must be oxidized to NO2 before adsorption on the storage components. However, NO2 typically constitutes less than 10% of NOx in lean exhaust, so catalytic oxidation of NO is essential. Platinum has been found to be especially active for NO oxidation, and is widely used in DOC and LNT catalysts. However, because of the high cost and poor thermal durability of Pt-based catalysts, there is substantial interest in the development of alternatives. The objective of this project, in collaboration with partner General Motors, is to develop mixed metal oxide catalysts for NO oxidation, enabling lower precious metal usage in emission control systems. [1] M. Koebel, G. Madia, and M. Elsener, Catalysis Today 73, 239 (2002). [2] C. H. Kim, G. S. Qi, K. Dahlberg, and W. Li, Science 327, 1624 (2010).

  3. MOX Fabrication Isolation Considerations

    SciTech Connect (OSTI)

    Eric L. Shaber; Bradley J Schrader

    2005-08-01

    This document provides a technical position on the preferred level of isolation to fabricate demonstration quantities of mixed oxide transmutation fuels. The Advanced Fuel Cycle Initiative should design and construct automated glovebox fabrication lines for this purpose. This level of isolation adequately protects the health and safety of workers and the general public for all mixed oxide (and other transmutation fuel) manufacturing efforts while retaining flexibility, allowing parallel development and setup, and minimizing capital expense. The basis regulations, issues, and advantages/disadvantages of five potential forms of isolation are summarized here as justification for selection of the preferred technical position.

  4. Weapons-Grade MOX Fuel Burnup Characteristics in Advanced Test Reactor Irradiation

    SciTech Connect (OSTI)

    G. S. Chang

    2006-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory (LANL) by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, 40, and 50 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energys Fissile Materials Disposition Program (FMDP). A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2(MCWO). MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. The fuel burnup analyses presented in this study were performed using MCWO. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements and the irradiated WG-MOX post irradiation examination (PIE) data.

  5. Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

    SciTech Connect (OSTI)

    Kudinov, K.G.; Tretyakov, A.A.; Sorokin, Y.P.; Bondin, V.V.; Manakova, L.F.; Jardine, L.J.

    2001-12-01

    In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is incineration

  6. Thermochemical cycle of a mixed metal oxide for augmentation...

    Office of Scientific and Technical Information (OSTI)

    Thermochemical cycle of a mixed metal oxide for augmentation of thermal energy storage in solid particles. Citation Details In-Document Search Title: Thermochemical cycle of a ...

  7. Hanford MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect (OSTI)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site (SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. Hanford has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 1 facility. In all, a total of three LA MOX fuel fabrication options were identified by Hanford that could accommodate the program. In every case, only minor modification would be required to ready any of the facilities to accept the equipment necessary to accomplish the LA program.

  8. ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect (OSTI)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1997-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement.

  9. Strong Support for MOX Continues

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    When completed, the MOX facility will convert weapons- grade plutonium to nuclear reactor fuel assemblies. These assemblies will then be used as fuel in commercial nuclear power ...

  10. Evaluation of codisposal viability of MOX (FFTF) DOE-owned fuel: Phase 1 -- Intact mode calculations

    SciTech Connect (OSTI)

    Goluoglu, S.; Davis, J.W.; Montierth, L.M.

    1999-07-01

    The authors provide the intact criticality information that supports the disposal of spent nuclear fuel (SNF) from the US Department of Energy's (DOE's) Fast Flux Test Facility (FFTF) in the potential Monitored Geologic Repository at Yucca Mountain. FFTF is one of more than 250 forms of DOE-owned SNF. Because of the variety of the DOE SNF, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The FFTF fuel is representative of the mixed-oxide fuel (MOX) group.

  11. Neutronics and safety characteristics of a 100% MOX fueled PWR using weapons grade plutonium

    SciTech Connect (OSTI)

    Biswas, D.; Rathbun, R.; Lee, Si Young; Rosenthal, P.

    1993-12-31

    Preliminary neutronics and safety studies, pertaining to the feasibility of using 100% weapons grade mixed-oxide (MOX) fuel in an advanced PWR Westinghouse design are presented in this paper. The preliminary results include information on boron concentration, power distribution, reactivity coefficients and xenon and control rode worth for the initial and the equilibrium cycle. Important safety issues related to rod ejection and steam line break accidents and shutdown margin requirements are also discussed. No significant change from the commercial design is needed to denature weapons-grade plutonium under the current safety and licensing criteria.

  12. LAB-SCALE DEMONSTRATION OF PLUTONIUM PURIFICATION BY ANION EXCHANGE, PLUTONIUM (IV) OXALATE PRECIPITATION, AND CALCINATION TO PLUTONIUM OXIDE TO SUPPORT THE MOX FEED MISSION

    SciTech Connect (OSTI)

    Crowder, M.; Pierce, R.

    2012-08-22

    H-Canyon and HB-Line are tasked with the production of PuO{sub 2} from a feed of plutonium metal. The PuO{sub 2} will provide feed material for the MOX Fuel Fabrication Facility. After dissolution of the Pu metal in H-Canyon, the solution will be transferred to HB-Line for purification by anion exchange. Subsequent unit operations include Pu(IV) oxalate precipitation, filtration and calcination to form PuO{sub 2}. This report details the results from SRNL anion exchange, precipitation, filtration, calcination, and characterization tests, as requested by HB-Line1 and described in the task plan. This study involved an 80-g batch of Pu and employed test conditions prototypical of HB-Line conditions, wherever feasible. In addition, this study integrated lessons learned from earlier anion exchange and precipitation and calcination studies. H-Area Engineering selected direct strike Pu(IV) oxalate precipitation to produce a more dense PuO{sub 2} product than expected from Pu(III) oxalate precipitation. One benefit of the Pu(IV) approach is that it eliminates the need for reduction by ascorbic acid. The proposed HB-Line precipitation process involves a digestion time of 5 minutes after the time (44 min) required for oxalic acid addition. These were the conditions during HB-line production of neptunium oxide (NpO{sub 2}). In addition, a series of small Pu(IV) oxalate precipitation tests with different digestion times were conducted to better understand the effect of digestion time on particle size, filtration efficiency and other factors. To test the recommended process conditions, researchers performed two nearly-identical larger-scale precipitation and calcination tests. The calcined batches of PuO{sub 2} were characterized for density, specific surface area (SSA), particle size, moisture content, and impurities. Because the 3013 Standard requires that the calcination (or stabilization) process eliminate organics, characterization of PuO{sub 2} batches monitored the

  13. O/M RATIO MEASUREMENT IN PURE AND MIXED OXIDE FULES - WHERE ARE WE NOW?

    SciTech Connect (OSTI)

    J. RUBIN; ET AL

    2000-12-01

    The oxygen-to-metal (O/M) ratio is one of the most critical parameters of nuclear fuel fabrication, and its measurement is closely monitored for manufacturing process control and to ensure the service behavior of the final product. Thermogravimetry is the most widely used method, the procedure for which has remained largely unchanged since its development some thirty years ago. It was not clear to us, however, that this method is still the optimum one in light of advances in instrumentation, and in the current regulatory environment, particularly with regard to waste management and disposal. As part of the MOX fuel fabrication program at Los Alamos, we conducted a comprehensive review of methods for O/M measurements in UO{sub 2}, PuO{sub 2} and mixed oxide fuels for thermal reactors. A concerted effort was made to access information not available in the open literature. We identified approximately thirty five experimental methods that (a) have been developed with the intent of measuring O/M, (b) provided O/M indirectly by suitable reduction of the measured data, or (c) could provide O/M data with suitable data reduction or when combined with other methods. We will discuss the relative strengths and weaknesses of these methods in their application to current routine and small-lot production environment.

  14. SMALL-SCALE TESTING OF PLUTONIUM (IV) OXALATE PRECIPITATION AND CALCINATION TO PLUTONIUM OXIDE TO SUPPORT THE MOX FEED MISSION

    SciTech Connect (OSTI)

    Crowder, M.; Pierce, R.; Scogin, J.; Daniel, G.; King, W.

    2012-06-25

    The H-Canyon facility will be used to dissolve Pu metal for subsequent purification and conversion to plutonium dioxide (PuO{sub 2}) using Phase II of HB-Line. To support the new mission, SRNL conducted a series of experiments to produce calcined plutonium (Pu) oxide and measure the physical properties and water adsorption of that material. This data will help define the process operating conditions and material handling steps for HB-Line. An anion exchange column experiment produced 1.4 L of a purified 52.6 g/L Pu solution. Over the next nine weeks, seven Pu(IV) oxalate precipitations were performed using the same stock Pu solution, with precipitator feed acidities ranging from 0.77 M to 3.0 M nitric acid and digestion times ranging from 5 to 30 minutes. Analysis of precipitator filtrate solutions showed Pu losses below 1% for all precipitations. The four larger precipitation batches matched the target oxalic acid addition time of 44 minutes within 4 minutes. The three smaller precipitation batches focused on evaluation of digestion time and the oxalic acid addition step ranged from 25-34 minutes because of pump limitations in the low flow range. Following the precipitations, 22 calcinations were performed in the range of 610-690 C, with the largest number of samples calcined at either 650 or 635 C. Characterization of the resulting PuO{sub 2} batches showed specific surface areas in the range of 5-14 m{sup 2}/g, with 16 of the 22 samples in the range of 5-10 m2/g. For samples analyzed with typical handling (exposed to ambient air for 15-45 minutes with relative humidities of 20-55%), the moisture content as measured by Mass Spectrometry ranged from 0.15 to 0.45 wt % and the total mass loss at 1000 C, as measured by TGA, ranged from 0.21 to 0.58 wt %. For the samples calcined between 635 and 650 C, the moisture content without extended exposure ranged from 0.20 to 0.38 wt %, and the TGA mass loss ranged from 0.26 to 0.46 wt %. Of these latter samples, the samples

  15. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect (OSTI)

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule.

  16. Chemical and Radiochemical Composition of Thermally Stabilized Plutonium Oxide from the Plutonium Finishing Plant Considered as Alternate Feedstock for the Mixed Oxide Fuel Fabrication Facility

    SciTech Connect (OSTI)

    Tingey, Joel M.; Jones, Susan A.

    2005-07-01

    PFP. Samples varied in appearance depending on the original source of material. Rocky Flats items were mostly dark olive green with clumps that crushed easily with a mortar and pestle. PRF/RMC items showed more variability. These items were mostly rust colored. One sample contained white particles that were difficult to crush, and another sample was a dark grey with a mixture of fines and large, hard fragments. The appearance and feel of the fragments indicated they might be an alloy. The color of the solution samples was indicative of the impurities in the sample. The double-pass filtrate solution was a brown color indicative of the iron impurities in the sample. The other solution sample was light gray in color. Radiochemical analyses, including thermal ionization mass spectrometry (TIMS), alpha and gamma energy analysis (AEA and GEA), and kinetic phosphorescence analysis (KPA), indicate that these materials are all weapons-grade plutonium with consistent plutonium isotopics. A small amount of uranium (<0.14 wt%) is also present in these samples. The isotopic composition of the uranium varied widely but was consistent among each category of material. The primary water-soluble anions in these samples were Cl-, NO3-, SO42-, and PO43-. The only major anion observed in the Rocky Flats materials was Cl-, but the PRF/RMC samples had significant quantities of all of the primary anions observed. Prompt gamma measurements provide a representative analysis of the Cl- concentration in the bulk material. The primary anions observed in the solution samples were NO3-, and PO43-. The concentration of these anions did not exceed the mixed oxide (MOX) specification limits. Cations that exceeded the MOX specification limits included Cr, Fe, Ni, Al, Cu, and Si. All of the samples exceeded at least the 75% specification limit in one element.

  17. PRIVACY IMPACT ASSESSMENT: Shaw Areva MOX Services, LLC MOX

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    . ,-) ')7 73?¥i5": )~"'f"YC-:;'~dt?f(~"'f9'FrrZ , . PRIVACY IMPACT ASSESSMENT: Shaw Areva MOX Services, LLC MOX Services Unclassified Information System Template - January 30, 2009, Version 2 Department of Energy Privacy Impact Assessment (pIA) Guidance is provided in the template. See DOE Order 206.1, Department of Energy Privacy Program, Appendix A, Privacy Impact Assessments, for requirements and additional guidance for conducting a PIA:

  18. Composite mixed oxide ionic and electronic conductors for hydrogen separation

    DOE Patents [OSTI]

    Gopalan, Srikanth; Pal, Uday B.; Karthikeyan, Annamalai; Hengdong, Cui

    2009-09-15

    A mixed ionic and electronic conducting membrane includes a two-phase solid state ceramic composite, wherein the first phase comprises an oxygen ion conductor and the second phase comprises an n-type electronically conductive oxide, wherein the electronically conductive oxide is stable at an oxygen partial pressure as low as 10.sup.-20 atm and has an electronic conductivity of at least 1 S/cm. A hydrogen separation system and related methods using the mixed ionic and electronic conducting membrane are described.

  19. Design and synthesis of mixed oxides nanoparticles for biofuel applications

    SciTech Connect (OSTI)

    Chen, Senniang

    2010-05-15

    The work in this dissertation presents the synthesis of two mixed metal oxides for biofuel applications and NMR characterization of silica materials. In the chapter 2, high catalytic efficiency of calcium silicate is synthesized for transesterfication of soybean oil to biodisels. Chapter 3 describes the synthesis of a new Rh based catalyst on mesoporous manganese oxides. The new catalyst is found to have higher activity and selectivity towards ethanol. Chapter 4 demonstrates the applications of solid-state Si NMR in the silica materials.

  20. Progress on Acidic Zirconia Mixed Oxides for Efficient NH3-SCR...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Zirconia Mixed Oxides for Efficient NH3-SCR Catalysis Progress on Acidic Zirconia Mixed Oxides for Efficient NH3-SCR Catalysis Details progress on non-zeolitic zirconia-based ...

  1. Mixed-conducting oxides for gas separation applications.

    SciTech Connect (OSTI)

    Balachandran, U.

    1999-04-20

    Mixed-conducting oxides are attracting increased attention because of their potential uses in high-temperature electrochemical applications such as solid-oxide fuel cells, batteries, sensors, and gas-permeable membranes. We are developing mixed-conducting, dense ceramic membranes to selectively transport oxygen and hydrogen. Ceramic membranes made of Sr-Fe-Co oxide (SFC), which exhibits high combined electronic and oxygen ionic conductivities, can be used to selectively transport oxygen during the partial oxidation of methane to synthesis gas (syngas, a mixture of CO and H{sub 2}). Steady-state oxygen permeability of SrFeCo{sub 0.5}O{sub x} has been measured as a function of oxygen-partial-pressure gradient and temperature. At 900 C, oxygen permeability was {approx}2.5 scc{center_dot}cm{sup {minus}2}-min{sup {minus}1} for a 2.9-mm-thick membrane, and this value increases as membrane thickness decreases. We have fabricated tubular SrFeCo{sub 0.5}O{sub x} membranes and operated them at 900 C for >1000 h during conversion of methane into syngas. Yttria-doped BaCeO{sub 3} (BCY) is a good protonic conductor; however, its lack of electronic conductivity can potentially limit its hydrogen permeability. To enhance the electronic conductivity and thus improve hydrogen permeation, a membrane composite material was developed. Nongalvanic permeation of hydrogen through the composite membrane was characterized as a function of thickness.

  2. Development of mixed-conducting oxides for gas separation

    SciTech Connect (OSTI)

    Balachandran, U.; Ma, B.; Maiya, P.S.

    1997-08-01

    Mixed-conducting oxides have been used in many applications, including fuel cells, gas separation membranes, sensors, and electrocatalysis. The authors are developing a mixed-conducting, dense ceramic membrane for selectively transporting oxygen and hydrogen. Ceramic membranes made of Sr-Fe-Co oxide, which has high combined electronic and oxygen ionic conductions, can be used to selectively transport oxygen during the partial oxidation of methane to synthesis gas (syngas, CO + H{sub 2}). The authors have measured the steady-state oxygen permeability of SrFeCo{sub 0.5}O{sub x} as a function of oxygen-partial-pressure gradient and temperature. At 900{degrees}C, oxygen permeability was {approx}2.5 scc{center_dot}cm{sup {minus}2}{center_dot}min{sup {minus}1} for a 2.9-mm-thick membrane and this value increases as membrane thickness decreases. The authors have fabricated tubular SrFeCo{sub 0.5}O{sub x} membranes and operated them at 900{degrees}C for >1000 h during conversion of methane into syngas. The hydrogen ion (proton) transport properties of yttria-doped BaCeO{sub 3} were investigated by impedance spectroscopy and open-cell voltage measurements. High proton conductivity and a high protonic transference number make yttria-doped BaCeO{sub 3} a potential membrane for hydrogen separation.

  3. SRS MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    SciTech Connect (OSTI)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R.

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site(SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. SRS has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 2 or 3 facility with storage of bulk PuO{sub 2} and assembly, storage, and shipping of fuel bundles in an S and S Category 1 facility. The total Category 1 approach, which is the recommended option, would be done in the 221-H Canyon Building. A facility that was never in service will be removed from one area, and a hardened wall will be constructed in another area to accommodate execution of the LA fuel fabrication. The non-Category 1 approach would require removal of process equipment in the FB-Line metal production and packaging glove boxes, which requires work in a contamination area. The Immobilization Hot Demonstration Program

  4. Reduction of spalling in mixed metal oxide desulfurization sorbents by addition of a large promoter metal oxide

    DOE Patents [OSTI]

    Poston, James A.

    1997-01-01

    Mixed metal oxide pellets for removing hydrogen sulfide from fuel gas mixes derived from coal are stabilized for operation over repeated cycles of desulfurization and regeneration reactions by addition of a large promoter metal oxide such as lanthanum trioxide. The pellets, which may be principally made up of a mixed metal oxide such as zinc titanate, exhibit physical stability and lack of spalling or decrepitation over repeated cycles without loss of reactivity. The lanthanum oxide is mixed with pellet-forming components in an amount of 1 to 10 weight percent.

  5. Reduction of spalling in mixed metal oxide desulfurization sorbents by addition of a large promoter metal oxide

    DOE Patents [OSTI]

    Poston, J.A.

    1997-12-02

    Mixed metal oxide pellets for removing hydrogen sulfide from fuel gas mixes derived from coal are stabilized for operation over repeated cycles of desulfurization and regeneration reactions by addition of a large promoter metal oxide such as lanthanum trioxide. The pellets, which may be principally made up of a mixed metal oxide such as zinc titanate, exhibit physical stability and lack of spalling or decrepitation over repeated cycles without loss of reactivity. The lanthanum oxide is mixed with pellet-forming components in an amount of 1 to 10 weight percent.

  6. Preparation of uniform nanoparticles of ultra-high purity metal oxides, mixed metal oxides, metals, and metal alloys

    DOE Patents [OSTI]

    Woodfield, Brian F.; Liu, Shengfeng; Boerio-Goates, Juliana; Liu, Qingyuan; Smith, Stacey Janel

    2012-07-03

    In preferred embodiments, metal nanoparticles, mixed-metal (alloy) nanoparticles, metal oxide nanoparticles and mixed-metal oxide nanoparticles are provided. According to embodiments, the nanoparticles may possess narrow size distributions and high purities. In certain preferred embodiments, methods of preparing metal nanoparticles, mixed-metal nanoparticles, metal oxide nanoparticles and mixed-metal nanoparticles are provided. These methods may provide tight control of particle size, size distribution, and oxidation state. Other preferred embodiments relate to a precursor material that may be used to form nanoparticles. In addition, products prepared from such nanoparticles are disclosed.

  7. Diffusion-controlled creep in mixed-conducting oxides

    SciTech Connect (OSTI)

    Routbort, J.L.; Goretta, K.C.; Cook, R.E.; Wolfenstine, J.; Armstrong, T.R.; Clauss, C.; Dominguez-Rodriguez, A.

    1996-06-01

    Steady-state creep rate of the mixed conducting oxides La{sub 1-x}Sr{sub x}MnO{sub 3} (x=0.1, 0.15, 0.25) and La{sub 0.7}Ca{sub 0.3}MnO{sub 3} has been investigated between 1150 and 1300 C. Creep parameters and TEM indicate that deformation is controlled by lattice diffusion of one of the cations. Dependence of creep rate on Sr concentration, combined with a point-defect model, confirms this hypothesis; however the oxygen partial pressure dependence of creep (from 10{sup -1} to 2x10{sup 4} Pa) cannot be accounted for within the framework of a simple point-defect model.

  8. Progress on Acidic Zirconia Mixed Oxides for Efficient NH3-SCR Catalysis |

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Department of Energy Acidic Zirconia Mixed Oxides for Efficient NH3-SCR Catalysis Progress on Acidic Zirconia Mixed Oxides for Efficient NH3-SCR Catalysis Details progress on non-zeolitic zirconia-based mixed oxides as promising new SCR catalyst materials and results of engine bench testing of full-size SCR prototype confirms Details progress on non-zeolitic zirconia-based mixed oxides as promising new SCR catalyst materials and results of engine bench testing of full-size SCR prototype

  9. MOX Reprocessing at Tokai Reprocessing Plant

    SciTech Connect (OSTI)

    Taguchi, Katsuya; Nagaoka, Shinichi; Yamanaka, Atsushi; Nakamura, Yoshinobu; Omori, Eiichi; SATO, Takehiko; MIURA, Nobuyuki

    2007-07-01

    In March 2007, the first reprocessing of the 'Type B' MOX spent fuels of the Prototype Advanced Thermal Reactor FUGEN was initiated at Tokai Reprocessing Plant as a plant-scale demonstration of MOX fuel reprocessing. The operation was advanced satisfactorily and it has been confirmed that the MOX fuels as well as UO{sub 2} fuels can be reprocessed safely. Some characteristics of MOX fuels on reprocessing, such as properties of undissolved residue affecting the clarification process, are becoming visible. Reprocessing of the 'Type B' MOX fuels will be continued for several more years from now on, further investigations on solubility of fuels, characteristics of undissolved residues, progress of solvent degradation and so on will be continued. (authors)

  10. Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Goldmann, Andrew S.; Wagner, Kenneth C.; Powers, Dana Auburn; Ashbaugh, Scott G.; Longmire, Pamela

    2010-04-01

    As part of a Nuclear Regulatory Commission (NRC) research program to evaluate the impact of using mixed-oxide (MOX) fuel in commercial nuclear power plants, a study was undertaken to evaluate the impact of the usage of MOX fuel on the consequences of postulated severe accidents. A series of 23 severe accident calculations was performed using MELCOR 1.8.5 for a four-loop Westinghouse reactor with an ice condenser containment. The calculations covered five basic accident classes that were identified as the risk- and consequence-dominant accident sequences in plant-specific probabilistic risk assessments for the McGuire and Catawba nuclear plants, including station blackouts and loss-of-coolant accidents of various sizes, with both early and late containment failures. Ultimately, the results of these MELCOR simulations will be used to provide a supplement to the NRC's alternative source term described in NUREG-1465. Source term magnitude and timing results are presented consistent with the NUREG-1465 format. For each of the severe accident release phases (coolant release, gap release, in-vessel release, ex-vessel release, and late in-vessel release), source term timing information (onset of release and duration) is presented. For all release phases except for the coolant release phase, magnitudes are presented for each of the NUREG-1465 radionuclide groups. MELCOR results showed variation of noble metal releases between those typical of ruthenium (Ru) and those typical of molybdenum (Mo); therefore, results for the noble metals were presented for Ru and Mo separately. The collection of the source term results can be used as the basis to develop a representative source term (across all accident types) that will be the MOX supplement to NUREG-1465.

  11. MOX Services Unclassified Information System PIA, National Nuclear Services

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Administration | Department of Energy MOX Services Unclassified Information System PIA, National Nuclear Services Administration MOX Services Unclassified Information System PIA, National Nuclear Services Administration MOX Services Unclassified Information System PIA, National Nuclear Services Administration MOX Services Unclassified Information System PIA, National Nuclear Services Administration (378.48 KB) More Documents & Publications TRAIN-PIA.pdf Occupational Medicine - Assistant

  12. Performance of Thorium-Based Mixed Oxide Fuels for the Consumption of Plutonium and Minor Actinides in Current and Advanced Reactors

    SciTech Connect (OSTI)

    Weaver, Kevan Dean; Herring, James Stephen

    2002-06-01

    A renewed interest in thorium-based fuels has arisen lately based on the need for proliferation resistance, longer fuel cycles, higher burnup and improved wasteform characteristics. Recent studies have been directed toward homogeneously mixed, heterogeneously mixed, and seed-and-blanket thorium-uranium fuel cycles that rely on "in situ" use of the bred-in U-233. However, due to the higher initial enrichment required to achieve acceptable burnups, these fuels are encountering economic constraints. Thorium can nevertheless play a large role in the nuclear fuel cycle; particularly in the reduction of plutonium. While uranium-based mixedoxide (MOX) fuel will decrease the amount of plutonium, the reduction is limited due to the breeding of more plutonium (and higher actinides) from the U-238. Here we present calculational results and a comparison of the potential burnup of a thorium-based and uranium-based mixed oxide fuel in a light water reactor (LWR). Although the uranium-based fuels outperformed the thorium-based fuels in achievable burnup, a depletion comparison of the initially charged plutonium (both reactor and weapons grade) showed that the thorium-based fuels outperformed the uranium-based fuels by more that a factor of 2; where more than 70% of the total plutonium in the thorium-based fuel is consumed during the cycle. This is significant considering that the achievable burnup of the thorium-based fuels were 1.4 to 4.6 times less than the uranium-based fuels. Furthermore, use of a thorium-based fuel could also be used as a strategy for reducing the amount of long-lived nuclides (including the minor actinides), and thus the radiotoxicity in spent nuclear fuel. Although the breeding of U-233 is a concern, the presence of U-232 and its daughter products can aid in making this fuel self-protecting, and/or enough U-238 can be added to denature the fissile uranium. From these calculations, it appears that thorium-based fuel for plutonium incineration is superior as

  13. Complex catalytic behaviors of CuTiOx mixed-oxide during CO oxidation

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Kim, Hyun You; Liu, Ping

    2015-09-21

    Mixed metal oxides have attracted considerable attention in heterogeneous catalysis due to the unique stability, reactivity, and selectivity. Here, the activity and stability of the CuTiOx monolayer film supported on Cu(111), CuTiOx/Cu(111), during CO oxidation was explored using density functional theory (DFT). The unique structural frame of CuTiOx is able to stabilize and isolate a single Cu+ site on the terrace, which is previously proposed active for CO oxidation. Furthermore, it is not the case, where the reaction via both the Langmuir–Hinshelwood (LH) and the Mars-van Krevelen (M-vK) mechanisms are hindered on such single Cu+ site. Upon the formation ofmore » step-edges, the synergy among Cuδ+ sites, TiOx matrix, and Cu(111) is able to catalyze the reaction well. Depending on temperatures and partial pressure of CO and O2, the surface structure varies, which determines the dominant mechanism. In accordance with our results, the Cuδ+ ion alone does not work well for CO oxidation in the form of single sites, while the synergy among multiple active sites is necessary to facilitate the reaction.« less

  14. Development of an integrated, unattended assay system for LWR-MOX fuel pellet trays

    SciTech Connect (OSTI)

    Stewart, J.E.; Hatcher, C.R.; Pollat, L.L.

    1994-08-01

    Four identical unattended plutonium assay systems have been developed for use at the new light-water-reactor mixed oxide (LWR-MOX) fuel fabrication facility at Hanau, Germany. The systems provide quantitative plutonium verification for all MOX pellet trays entering or leaving a large, intermediate store. Pellet-tray transport and storage systems are highly automated. Data from the ``I-Point`` (information point) assay systems will be shared by the Euratom and International Atomic Energy Agency (IAEA) Inspectorates. The I-Point system integrates, for the first time, passive neutron coincidence counting (NCC) with electro-mechanical sensing (EMS) in unattended mode. Also, provisions have been made for adding high-resolution gamma spectroscopy. The system accumulates data for every tray entering or leaving the store between inspector visits. During an inspection, data are analyzed and compared with operator declarations for the previous inspection period, nominally one month. Specification of the I-point system resulted from a collaboration between the IAEA, Euratom, Siemens, and Los Alamos. Hardware was developed by Siemens and Los Alamos through a bilateral agreement between the German Federal Ministry of Research and Technology (BMFT) and the US DOE. Siemens also provided the EMS subsystem, including software. Through the USSupport Program to the IAEA, Los Alamos developed the NCC software (NCC COLLECT) and also the software for merging and reviewing the EMS and NCC data (MERGE/REVIEW). This paper describes the overall I-Point system, but emphasizes the NCC subsystem, along with the NCC COLLECT and MERGE/REVIEW codes. We also summarize comprehensive testing results that define the quality of assay performance.

  15. Research and development of americium-containing mixed oxide fuel for fast reactors

    SciTech Connect (OSTI)

    Tanaka, Kosuke; Osaka, Masahiko; Sato, Isamu; Miwa, Shuhei; Koyama, Shin-ichi; Ishi, Yohei; Hirosawa, Takashi; Obayashi, Hiroshi; Yoshimochi, Hiroshi; Tanaka, Kenya

    2007-07-01

    The present status of the R and D program for americium-containing MOX fuel is reported. Successful achievements for development of fabrication technology with remote handling and evaluation of irradiation behavior together with evaluation of thermo-chemical properties based on the out-of-pile experiments are mentioned with emphasis on effects of Am addition on the MOX fuel properties. (authors)

  16. DOSE RATES FOR WESTINGHOUSE 17X17 MOX PWR SNF IN A WASTE PACKAGE (SCPB: N/A)

    SciTech Connect (OSTI)

    T.L. Lotz

    1997-01-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to estimate the dose rate on and near the surface a Multi-Purpose Canister (MPC) PWR waste package (WP) which is loaded with Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel. The 21 PWR MPC WP is used to provide an upper bound for waste package designs since the 12 PWR MPC WP will have a smaller source term and an equivalent amount of shielding. the objectives of this evaluation are to calculate the requested dose rate(s) and document the calculation in a fashion to allow comparisons to other waste forms and WP designs at a future time.

  17. Thermodynamic calculations of oxygen self-diffusion in mixed-oxide nuclear fuels

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Parfitt, David C.; Cooper, Michael William; Rushton, Michael J.D.; Christopoulos, S. R.; Fitzpatrick, M. E.; Chroneos, A.

    2016-07-29

    Mixed-oxide fuels containing uranium with thorium and/or plutonium may play an important part in future nuclear fuel cycles. There are, however, significantly less data available for these materials than conventional uranium dioxide fuel. In the present study, we employ molecular dynamics calculations to simulate the elastic properties and thermal expansivity of a range of mixed oxide compositions. These are then used to support equations of state and oxygen self-diffusion models to provide a self-consistent prediction of the behaviour of these mixed oxide fuels at arbitrary compositions.

  18. Operation of mixed conducting metal oxide membrane systems under transient conditions

    DOE Patents [OSTI]

    Carolan, Michael Francis

    2008-12-23

    Method of operating an oxygen-permeable mixed conducting membrane having an oxidant feed side, an oxidant feed surface, a permeate side, and a permeate surface, which method comprises controlling the differential strain between the permeate surface and the oxidant feed surface at a value below a selected maximum value by varying the oxygen partial pressure on either or both of the oxidant feed side and the permeate side of the membrane.

  19. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    SciTech Connect (OSTI)

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W.

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program.

  20. Evaluation of codisposal viability of MOX (FFTF) DOE-owned fuel: Phase 2 -- Degraded mode calculations

    SciTech Connect (OSTI)

    Goluoglu, S.; Angers, L.; Davis, J.W.; Stockman, H.; Gottlieb, P.; Montierth, L.M.

    1999-07-01

    The authors provide the degraded criticality information that supports the disposal of spent nuclear fuel (SNF) from the US Department of Energy's (DOE's) Fast Flux Test facility (FFTF) in the potential Monitored Geologic Repository (MGR) at Yucca Mountain. FFTF is one of more than 250 forms of DOE-owned SNF. Because of the variety of the SNF, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The FFTF fuel is a mixture of uranium and plutonium oxides and is representative of the mixed-oxide fuel (MOX) group. The analyses were performed according to the disposal criticality analysis methodology that was documented in the topical report submitted to the US nuclear Regulatory Commission (YMP/TR-004Q). The methodology includes analyzing the geochemical and physical processes that can breach the waste package and degrade the waste forms. This paper summarizes the results of geochemistry degradation analysis and the criticality calculations using the degradation products.

  1. PLUTONIUM LOADING CAPACITY OF REILLEX HPQ ANION EXCHANGE COLUMN - AFS-2 PLUTONIUM FLOWSHEET FOR MOX

    SciTech Connect (OSTI)

    Kyser, E.; King, W.; O'Rourke, P.

    2012-07-26

    Radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the dependence of column loading performance on the feed composition in the H-Canyon dissolution process for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). These loading experiments show that a representative feed solution containing {approx}5 g Pu/L can be loaded onto Reillex{trademark} HPQ resin from solutions containing 8 M total nitrate and 0.1 M KF provided that the F is complexed with Al to an [Al]/[F] molar ratio range of 1.5-2.0. Lower concentrations of total nitrate and [Al]/[F] molar ratios may still have acceptable performance but were not tested in this study. Loading and washing Pu losses should be relatively low (<1%) for resin loading of up to 60 g Pu/L. Loading above 60 g Pu/L resin is possible, but Pu wash losses will increase such that 10-20% of the additional Pu fed may not be retained by the resin as the resin loading approaches 80 g Pu/L resin.

  2. Molten carbonate fuel cell cathode with mixed oxide coating

    DOE Patents [OSTI]

    Hilmi, Abdelkader; Yuh, Chao-Yi

    2013-05-07

    A molten carbonate fuel cell cathode having a cathode body and a coating of a mixed oxygen ion conductor materials. The mixed oxygen ion conductor materials are formed from ceria or doped ceria, such as gadolinium doped ceria or yttrium doped ceria. The coating is deposited on the cathode body using a sol-gel process, which utilizes as precursors organometallic compounds, organic and inorganic salts, hydroxides or alkoxides and which uses as the solvent water, organic solvent or a mixture of same.

  3. Synergetic effects of mixed copper-iron oxides oxygen carriers in chemical looping combustion

    SciTech Connect (OSTI)

    Siriwardane, Ranjani; Tian, Hanjing; Simonyi, Thomas; Poston, James

    2013-06-01

    Chemical looping combustion (CLC) is an emerging technology for clean energy production from fuels. CLC produces sequestration-ready CO{sub 2}-streams without a significant energy penalty. Development of efficient oxygen carriers is essential to successfully operate a CLC system. Copper and iron oxides are promising candidates for CLC. Copper oxide possesses high reactivity but it has issues with particle agglomeration due to its low melting point. Even though iron oxide is an inexpensive oxygen carrier it has a slower reactivity. In this study, mixed metal oxide carriers containing iron and copper oxides were evaluated for coal and methane CLC. The components of CuO and Fe{sub 2}O{sub 3} were optimized to obtain good reactivity while maintaining physical and chemical stability during cyclic reactions for methane-CLC and solid-fuel CLC. Compared with single metal oxygen carriers, the optimized Cu–Fe mixed oxide oxygen carriers demonstrated high reaction rate, better combustion conversion, greater oxygen usage and improved physical stability. Thermodynamic calculations, XRD, TGA, flow reactor studies and TPR experiments suggested that there is a strong interaction between CuO and Fe{sub 2}O{sub 3} contributing to a synergistic effect during CLC reactions. The amount of oxygen release of the mixed oxide carrier in the absence of a fuel was similar to that of the single metal oxides. However, in the presence of fuels, the oxygen consumption and the reaction profiles of the mixed oxide carriers were significantly better than that of the single metal oxides. The nature of the fuel not only influenced the reactivity, but also the final reduction status of the oxygen carriers during chemical looping combustion. Cu oxide of the mixed oxide was fully reduced metallic copper with both coal and methane. Fe oxide of the mixed oxide was fully reduced Fe metal with methane but it was reduced to only FeO with coal. Possible mechanisms of how the presence of CuO enhances the

  4. How to stabilize highly active Cu+ cations in a mixed-oxide catalyst

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Mudiyanselage, Kumudu; Luo, Si; Kim, Hyun You; Yang, Xiaofang; Baber, Ashleigh E.; Hoffmann, Friedrich M.; Senanayake, Sananayake; Rodriguez, Jose A.; Chen, Jingguang G.; Liu, Ping; et al

    2015-09-12

    Mixed-metal oxides exhibit novel properties that are not present in their isolated constituent metal oxides and play a significant role in heterogeneous catalysis. In this study, a titanium-copper mixed-oxide (TiCuOx) film has been synthesized on Cu(111) and characterized by complementary experimental and theoretical methods. At sub-monolayer coverages of titanium, a Cu2O-like phase coexists with TiCuOx and TiOx domains. When the mixed-oxide surface is exposed at elevated temperatures (600–650 K) to oxygen, the formation of a well-ordered TiCuOx film occurs. Stepwise oxidation of TiCuOx shows that the formation of the mixed-oxide is faster than that of pure Cu2O. As the Timore » coverage increases, Ti-rich islands (TiOx) form. The adsorption of CO has been used to probe the exposed surface sites on the TiOx–CuOx system, indicating the existence of a new Cu+ adsorption site that is not present on Cu2O/Cu(111). Adsorption of CO on Cu+ sites of TiCuOx is thermally more stable than on Cu(111), Cu2O/Cu(111) or TiO2(110). The Cu+ sites in TiCuOx domains are stable under both reducing and oxidizing conditions whereas the Cu2O domains present on sub-monolayer loads of Ti can be reduced or oxidized under mild conditions. Furthermore, the results presented here demonstrate novel properties of TiCuOx films, which are not present on Cu(111), Cu2O/Cu(111), or TiO2(110), and highlight the importance of the preparation and characterization of well-defined mixed-metal oxides in order to understand fundamental processes that could guide the design of new materials.« less

  5. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    SciTech Connect (OSTI)

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

    2013-07-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

  6. Preparation of low oxygen-to-metal mixed oxide fuels for the advanced fast reactor

    SciTech Connect (OSTI)

    Kato, Masato; Nakamichi, Shinya; Takano, Tatsuo

    2007-07-01

    The preparation process for homogeneous mixed oxide pellets with a precise O/M ratio was established. The process was used to prepare pellets for heat treatments in two stages which consisted of the sintering process at high oxygen potential and the annealing process done in the atmosphere of controlled oxygen partial pressure. In the annealing process, it was found that abnormal growth of pores and occurrence of cracks were caused inside the pellet, and it was necessary for prevention of the microstructure change to control the oxygen potential of the atmosphere. Mixed oxide pellets with minor actinides were fabricated by the process and were provided to irradiation tests. (authors)

  7. HB-LINE ANION EXCHANGE PURIFICATION OF AFS-2 PLUTONIUM FOR MOX

    SciTech Connect (OSTI)

    Kyser, E. A.; King, W. D.

    2012-07-31

    Non-radioactive cerium (Ce) and radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the feasibility of using either gadolinium nitrate (Gd) or boric acid (B as H{sub 3}BO{sub 3}) as a neutron poison in the H-Canyon dissolution process. Expected typical concentrations of probable impurities were tested and the removal of these impurities by a decontamination wash was measured. Impurity concentrations are compared to two specifications - designated as Column A or Column B (most restrictive) - proposed for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). Use of Gd as a neutron poison requires a larger volume of wash for the proposed Column A specification. Since boron (B) has a higher proposed specification and is more easily removed by washing, it appears to be the better candidate for use in the H-Canyon dissolution process. Some difficulty was observed in achieving the Column A specification due to the limited effectiveness that the wash step has in removing the residual B after ~4 BV's wash. However a combination of the experimental 10 BV's wash results and a calculated DF from the oxalate precipitation process yields an overall DF sufficient to meet the Column A specification. For those impurities (other than B) not removed by 10 BV's of wash, the impurity is either not expected to be present in the feedstock or process, or recommendations have been provided for improvement in the analytical detection/method or validation of calculated results. In summary, boron is recommended as the appropriate neutron poison for H-Canyon dissolution and impurities are expected to meet the Column A specification limits for oxide production in HB-Line.

  8. HB-LINE ANION EXCHANGE PURIFICATION OF AFS-2 PLUTONIUM FOR MOX

    SciTech Connect (OSTI)

    Kyser, E.; King, W.

    2012-04-25

    Non-radioactive cerium (Ce) and radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the feasibility of using either gadolinium nitrate (Gd) or boric acid (B as H{sub 3}BO{sub 3}) as a neutron poison in the H-Canyon dissolution process. Expected typical concentrations of probable impurities were tested and the removal of these impurities by a decontamination wash was measured. Impurity concentrations are compared to two specifications - designated as Column A or Column B (most restrictive) - proposed for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). Use of Gd as a neutron poison requires a larger volume of wash for the proposed Column A specification. Since boron (B) has a higher proposed specification and is more easily removed by washing, it appears to be the better candidate for use in the H-Canyon dissolution process. Some difficulty was observed in achieving the Column A specification due to the limited effectiveness that the wash step has in removing the residual B after {approx}4 BV's wash. However a combination of the experimental 10 BV's wash results and a calculated DF from the oxalate precipitation process yields an overall DF sufficient to meet the Column A specification. For those impurities (other than B) not removed by 10 BV's of wash, the impurity is either not expected to be present in the feedstock or process, or recommendations have been provided for improvement in the analytical detection/method or validation of calculated results. In summary, boron is recommended as the appropriate neutron poison for H-Canyon dissolution and impurities are expected to meet the Column A specification limits for oxide production in HB-Line.

  9. LANL disassembles "pits," makes mixed-oxide fuel

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    disassembles "pits" LANL disassembles "pits," makes mixed-oxide fuel LANL has successfully disassembled nuclear weapons "pits" and converted them into more than 240 kilograms of plutonium oxide. October 7, 2011 Los Alamos National Laboratory sits on top of a once-remote mesa in northern New Mexico with the Jemez mountains as a backdrop to research and innovation covering multi-disciplines from bioscience, sustainable energy sources, to plasma physics and new

  10. ENDF/B-VII.0, ENDF/B-VI, JEFF-3.1, AND JENDL-3.3 RESULTS FOR UNREFLECTED PLUTONIUM SOLUTIONS AND MOX LATTICES (U)

    SciTech Connect (OSTI)

    MOSTELLER, RUSSELL D.

    2007-02-09

    Previous studies have indicated that ENDF/B-VII preliminary releases {beta}-2 and {beta}-3, predecessors to the recent initial release of ENDF/B-VII.0, produce significantly better overall agreement with criticality benchmarks than does ENDF/B-VI. However, one of those studies also suggests that improvements still may be needed for thermal plutonium cross sections. The current study substantiates that concern by examining criticality benchmarks for unreflected spheres of plutonium-nitrate solutions and for slightly and heavily borated mixed-oxide (MOX) lattices. Results are presented for the JEFF-3.1 and JENDL-3.3 nuclear data libraries as well as ENDF/B-VII.0 and ENDF/B-VI. It is shown that ENDF/B-VII.0 tends to overpredict reactivity for thermal plutonium benchmarks over at least a portion of the thermal range. In addition, it is found that additional benchmark data are needed for the deep thermal range.

  11. Screening study of mixed transition-metal oxides for use as cathodes in thermal batteries

    SciTech Connect (OSTI)

    Guidotti, R.A.; Reinhardt, F.W.

    1996-05-01

    Over 100 candidates were examined, including commercial materials and many that were synthesized in house. The mixed oxides were based on Ti, V, Nb, Cr, Mo, W, Mn, Fe, Co, Ni, and Cu doped with other transition metals. A number of individual (single-metal) oxides were included for comparison. The candidates were tested in single cells with Li(Si) anodes and separators based on LiCl-KCl eutectic. Screening was done under constant-current conditions at current densities of 125 me/cm{sup 2} and, to a lesser extent, 50 me/cm{sup 2} at 500 C. Relative performance and limitations of the oxide cathodes are discussed.

  12. Effect of Co/Ni ratios in cobalt nickel mixed oxide catalysts on methane combustion

    SciTech Connect (OSTI)

    Lim, Tae Hwan; Cho, Sung June; Yang, Hee Sung; Engelhard, Mark H.; Kim, Do Heui

    2015-07-31

    A series of cobalt nickel mixed oxide catalysts with the varying ratios of Co to Ni, prepared by co-precipitation method, were applied to methane combustion. Among the various ratios, cobalt nickel mixed oxides having the ratios of Co to Ni of (50:50) and (67:33) demonstrate the highest activity for methane combustion. Structural analysis obtained from X-ray diffraction (XRD) and extended X-ray absorption fine structure (EXAFS) evidently demonstrates that CoNi (50:50) and (67:33) samples consist of NiCo2O4and NiO phase and, more importantly, NiCo2O4spinel structure is largely distorted, which is attributed to the insertion of Ni2+ions into octahedral sites in Co3O4spinel structure. Such structural dis-order results in the enhanced portion of surface oxygen species, thus leading to the improved reducibility of the catalysts in the low temperature region as evidenced by temperature programmed reduction by hydrogen (H2TPR) and X-ray photoelectron spectroscopy (XPS) O 1s results. They prove that structural disorder in cobalt nickel mixed oxides enhances the catalytic performance for methane combustion. Thus, it is concluded that a strong relationship between structural property and activity in cobalt nickel mixed oxide for methane combustion exists and, more importantly, distorted NiCo2O4spinel structure is found to be an active site for methane combustion.

  13. Catalyst support of mixed cerium zirconium titanium oxide, including use and method of making

    DOE Patents [OSTI]

    Willigan, Rhonda R.; Vanderspurt, Thomas Henry; Tulyani, Sonia; Radhakrishnan, Rakesh; Opalka, Susanne Marie; Emerson, Sean C.

    2011-01-18

    A durable catalyst support/catalyst is capable of extended water gas shift operation under conditions of high temperature, pressure, and sulfur levels. The support is a homogeneous, nanocrystalline, mixed metal oxide of at least three metals, the first being cerium, the second being Zr, and/or Hf, and the third importantly being Ti, the three metals comprising at least 80% of the metal constituents of the mixed metal oxide and the Ti being present in a range of 5% to 45% by metals-only atomic percent of the mixed metal oxide. The mixed metal oxide has an average crystallite size less than 6 nm and forms a skeletal structure with pores whose diameters are in the range of 4-9 nm and normally greater than the average crystallite size. The surface area of the skeletal structure per volume of the material of the structure is greater than about 240 m.sup.2/cm.sup.3. The method of making and use are also described.

  14. Microsoft PowerPoint - MOX Adventure_Reactor Subcommittee_Tamara...

    National Nuclear Security Administration (NNSA)

    3 MOX Fuel - General MOX fuel pellets from former weapons plutonium Blend of 5% PuO 2 with 95% depleted UO 2 Like LEU fuel pellets, MOX fuel pellets are primarily ...

  15. How to stabilize highly active Cu+ cations in a mixed-oxide catalyst

    SciTech Connect (OSTI)

    Mudiyanselage, Kumudu; Luo, Si; Kim, Hyun You; Yang, Xiaofang; Baber, Ashleigh E.; Hoffmann, Friedrich M.; Senanayake, Sananayake; Rodriguez, Jose A.; Chen, Jingguang G.; Liu, Ping; Stacchiola, Dario J.

    2015-09-12

    Mixed-metal oxides exhibit novel properties that are not present in their isolated constituent metal oxides and play a significant role in heterogeneous catalysis. In this study, a titanium-copper mixed-oxide (TiCuOx) film has been synthesized on Cu(111) and characterized by complementary experimental and theoretical methods. At sub-monolayer coverages of titanium, a Cu2O-like phase coexists with TiCuOx and TiOx domains. When the mixed-oxide surface is exposed at elevated temperatures (600–650 K) to oxygen, the formation of a well-ordered TiCuOx film occurs. Stepwise oxidation of TiCuOx shows that the formation of the mixed-oxide is faster than that of pure Cu2O. As the Ti coverage increases, Ti-rich islands (TiOx) form. The adsorption of CO has been used to probe the exposed surface sites on the TiOx–CuOx system, indicating the existence of a new Cu+ adsorption site that is not present on Cu2O/Cu(111). Adsorption of CO on Cu+ sites of TiCuOx is thermally more stable than on Cu(111), Cu2O/Cu(111) or TiO2(110). The Cu+ sites in TiCuOx domains are stable under both reducing and oxidizing conditions whereas the Cu2O domains present on sub-monolayer loads of Ti can be reduced or oxidized under mild conditions. Furthermore, the results presented here demonstrate novel properties of TiCuOx films, which are not present on Cu(111), Cu2O/Cu(111), or TiO2(110), and highlight the importance of the preparation and characterization of well-defined mixed-metal oxides in order to understand fundamental processes that could guide the design of new materials.

  16. Method of producing homogeneous mixed metal oxides and metal-metal oxide mixtures

    DOE Patents [OSTI]

    Quinby, Thomas C.

    1978-01-01

    Metal powders, metal oxide powders, and mixtures thereof of controlled particle size are provided by reacting an aqueous solution containing dissolved metal values with excess urea. Upon heating, urea reacts with water from the solution leaving a molten urea solution containing the metal values. The molten urea solution is heated to above about 180.degree. C. whereupon metal values precipitate homogeneously as a powder. The powder is reduced to metal or calcined to form oxide particles. One or more metal oxides in a mixture can be selectively reduced to produce metal particles or a mixture of metal and metal oxide particles.

  17. Microsoft PowerPoint - MOX Adventure_Reactor Subcommittee_Tamara Reavis

    National Nuclear Security Administration (NNSA)

    MOX Adventure Tamara Reavis May 2015 Page 2 Overview of Presentation > Characteristics of MOX Fuel  MOX Fuel at Duke Energy  MOX Fuel and NMMSS Page 3 MOX Fuel - General  MOX fuel pellets from former weapons plutonium  Blend of ~5% PuO 2 with ~95% depleted UO 2  Like LEU fuel pellets, MOX fuel pellets are primarily uranium  Fission power comes primarily from plutonium (Pu 239 ) instead of uranium (U 235 )  Other than the material of the fuel pellets, MOX and uranium fuel

  18. Bi–Mn mixed metal organic oxide: A novel 3d-6p mixed metal coordination network

    SciTech Connect (OSTI)

    Shi, Fa-Nian; Rosa Silva, Ana; Bian, Liang

    2015-05-15

    A new terminology of metal organic oxide (MOO) was given a definition as a type of coordination polymers which possess the feature of inorganic connectivity between metals and the direct bonded atoms and show 1D, 2D or 3D inorganic sub-networks. One such compound was shown as an example. A 3d-6p (Mn–Bi. Named MOOMnBi) mixed metals coordination network has been synthesized via hydrothermal method. The new compound with the molecular formula of [MnBi{sub 2}O(1,3,5-BTC){sub 2}]{sub n} (1,3,5-BTC stands for benzene-1,3,5-tricarboxylate) was characterized via single crystal X-ray diffraction technique that revealed a very interesting 3-dimensional (3D) framework with Bi{sub 4}O{sub 2}(COO){sub 12} clusters which are further connected to Mn(COO){sub 6} fragments into a 2D MOO. The topology study indicates an unprecedented topological type with the net point group of (4{sup 13}.6{sup 2})(4{sup 13}.6{sup 8})(4{sup 16}.6{sup 5})(4{sup 18}.6{sup 10})(4{sup 22}.6{sup 14})(4{sup 3}) corresponding to 3,6,7,7,8,9-c hexa-nodal net. MOOMnBi shows catalytic activity in the synthesis of (E)-α,β-unsaturated ketones. - Graphical abstract: This metal organic framework (MOF) is the essence of a 2D metal organic oxide (MOO). - Highlights: • New concept of metal organic oxide (MOO) was defined and made difference from metal organic framework. • New MOO of MOOMnBi was synthesized by hydrothermal method. • Crystal structure of MOOMnBi was determined by single crystal X-ray analysis. • The catalytic activity of MOOMnBi was studied showing reusable after 2 cycles.

  19. Feasibility Study of MOX Fuel Online Burnup Analysis

    SciTech Connect (OSTI)

    Dennis, M.L.; Usman, S.

    2006-07-01

    This research is an extension of well established Non-Destructive Analysis of UO fuel using gamma spectroscopy of Cs-137 and other related isotopes. Given the performance similarities between UO fuel and MOX fuel, investigations are underway to develop similar correlation for MOX. MOX fuel burnup and decay simulations are being performed using ORIGEN-ARP (Oak Ridge Isotope Generation and Depletion Code - Automatic Rapid Processing). Simulation results are being analyzed and will be used to determine performance specifications of a detection system for field applications. Analysis of isotopic activity from irradiated fuel will be used to develop correlations to determine burn-up and Plutonium content of MOX fuel. These results will be particularly useful in view of the recent interest in MOX fuel. (authors)

  20. Investigation of some new hydro(solvo)thermal synthesis routes to nanostructured mixed-metal oxides

    SciTech Connect (OSTI)

    Burnett, David L.; Harunsani, Mohammad H.; Kashtiban, Reza J.; Playford, Helen Y.; Sloan, Jeremy; Hannon, Alex C.; Walton, Richard I.

    2014-06-01

    We present a study of two new solvothermal synthesis approaches to mixed-metal oxide materials and structural characterisation of the products formed. The solvothermal oxidation of metallic gallium by a diethanolamine solution of iron(II) chloride at 240 °C produces a crystalline sample of a spinel-structured material, made up of nano-scale particles typically 20 nm in dimension. XANES spectroscopy at the K-edge shows that the material contains predominantly Fe{sup 2+} in an octahedral environment, but that a small amount of Fe{sup 3+} is also present. Careful analysis using transmission electron microscopy and powder neutron diffraction shows that the sample is actually a mixture of two spinel materials: predominantly (>97%) an Fe{sup 2+} phase Ga{sub 1.8}Fe{sub 1.2}O{sub 3.9}, but with a minor impurity phase that is iron-rich. In contrast, the hydrothermal reaction of titanium bis(ammonium lactato)dihydroxide in water with increasing amounts of Sn(IV) acetate allows nanocrystalline samples of the SnO{sub 2}–TiO{sub 2} solid solution to be prepared directly, as proved by powder XRD and Raman spectroscopy. - Graphical abstract: New solvothermal synthesis approaches to spinel and rutile mixed-metal oxides are reported. - Highlights: • Solvothermal oxidation of gallium metal in organic iron(II) solution gives a novel iron gallate spinel. • Hydrothermal reaction of titanium(IV) complex and tin(IV) acetate produces the complete SnO{sub 2}–TiO{sub 2} solid solution. • Nanostructured mixed-metal oxide phases are produced directly from solution.

  1. Complex catalytic behaviors of CuTiOx mixed-oxide during CO oxidation

    SciTech Connect (OSTI)

    Kim, Hyun You; Liu, Ping

    2015-09-21

    Mixed metal oxides have attracted considerable attention in heterogeneous catalysis due to the unique stability, reactivity, and selectivity. Here, the activity and stability of the CuTiOx monolayer film supported on Cu(111), CuTiOx/Cu(111), during CO oxidation was explored using density functional theory (DFT). The unique structural frame of CuTiOx is able to stabilize and isolate a single Cu+ site on the terrace, which is previously proposed active for CO oxidation. Furthermore, it is not the case, where the reaction via both the LangmuirHinshelwood (LH) and the Mars-van Krevelen (M-vK) mechanisms are hindered on such single Cu+ site. Upon the formation of step-edges, the synergy among Cu?+ sites, TiOx matrix, and Cu(111) is able to catalyze the reaction well. Depending on temperatures and partial pressure of CO and O2, the surface structure varies, which determines the dominant mechanism. In accordance with our results, the Cu?+ ion alone does not work well for CO oxidation in the form of single sites, while the synergy among multiple active sites is necessary to facilitate the reaction.

  2. Kinetic study of the oxidation of n-butane on vanadium oxide supported on Al/Mg mixed oxide

    SciTech Connect (OSTI)

    Dejoz, A.; Vazquez, I.; Nieto, J.M.L.; Melo, F.

    1997-07-01

    The reaction kinetics of the oxidative dehydrogenation (ODH) of n-butane over vanadia supported on a heat-treated Mg/Al hydrotalcite (37.3 wt % of V{sub 2}O{sub 5}) was investigated by both linear and nonlinear regression techniques. A reaction network including the formation of butenes (1-, 2-cis-, and 2-trans-butene), butadiene, and carbon oxides by parallel and consecutive reactions, at low and high n-butane conversions, has been proposed. Langmuir-Hinshelwood (LH) models can be used as suitable models which allows reproduction of the global kinetic behavior, although differences between oxydehydrogenation and deep oxidation reactions have been observed. Thus, the formation of oxydehydrogenation products can be described by a LH equation considering a dissociative adsorption of oxygen while the formation of carbon oxides is described by a LH equation with a nondissociative adsorption of oxygen. Two different mechanisms operate on the catalyst: (i) a redox mechanism responsible of the formation of olefins and diolefins and associated to vanadium species, which is initiated by a hydrogen abstraction; (ii) a radical mechanism responsible of the formation of carbon oxides from n-butane and butenes and associated to vanadium-free sites of the support. On the other hand, the selectivity to oxydehydrogenation products increases with the reaction temperature. This catalytic performance can be explained taking into account the low reducibility of V{sup 5+}-sites and the higher apparent activation energies of the oxydehydrogenation reactions with respect to deep oxidation reactions.

  3. Savannah River Field Office | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Home fieldoffices Savannah River Field Office Learn More Mixed Oxide (MOX) Fuel Fabrication Facility

  4. A National Tracking Center for Monitoring Shipments of HEU, MOX, and Spent Nuclear Fuel: How do we implement?

    SciTech Connect (OSTI)

    Mark Schanfein

    2009-07-01

    Nuclear material safeguards specialists and instrument developers at US Department of Energy (USDOE) National Laboratories in the United States, sponsored by the National Nuclear Security Administration (NNSA) Office of NA-24, have been developing devices to monitor shipments of UF6 cylinders and other radioactive materials , . Tracking devices are being developed that are capable of monitoring shipments of valuable radioactive materials in real time, using the Global Positioning System (GPS). We envision that such devices will be extremely useful, if not essential, for monitoring the shipment of these important cargoes of nuclear material, including highly-enriched uranium (HEU), mixed plutonium/uranium oxide (MOX), spent nuclear fuel, and, potentially, other large radioactive sources. To ensure nuclear material security and safeguards, it is extremely important to track these materials because they contain so-called “direct-use material” which is material that if diverted and processed could potentially be used to develop clandestine nuclear weapons . Large sources could be used for a dirty bomb also known as a radioactive dispersal device (RDD). For that matter, any interdiction by an adversary regardless of intent demands a rapid response. To make the fullest use of such tracking devices, we propose a National Tracking Center. This paper describes what the attributes of such a center would be and how it could ultimately be the prototype for an International Tracking Center, possibly to be based in Vienna, at the International Atomic Energy Agency (IAEA).

  5. An Assessment of the Attractiveness of Material Associated with a MOX Fuel Cycle from a Safeguards Perspective

    SciTech Connect (OSTI)

    Bathke, Charles G; Wallace, Richard K; Ireland, John R; Johnson, M W; Hase, Kevin R; Jarvinen, Gordon D; Ebbinghaus, Bartley B; Sleaford, Brad W; Collins, Brian A; Robel, Martin; Bradley, Keith S; Prichard, Andrew W; Smith, Brian W

    2009-01-01

    This paper is an extension to earlier studies that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) and alternate nuclear materials (ANM) associated with the PUREX, UREX, coextraction, THOREX, and PYROX reprocessing schemes. This study extends the figure of merit (FOM) for evaluating attractiveness to cover a broad range of proliferant State and sub-national group capabilities. This study also considers those materials that will be recycled and burned, possibly multiple times, in LWRs [e.g., plutonium in the form of mixed oxide (MOX) fuel]. The primary conclusion of this study is that all fissile material needs to be rigorously safeguarded to detect diversion by a State and provided the highest levels of physical protection to prevent theft by sub-national groups; no 'silver bullet' has been found that will permit the relaxation of current international safeguards or national physical security protection levels. This series of studies has been performed at the request of the United States Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for nuclear materials in DOE nuclear facilities. The expanded methodology and updated findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security are discussed.

  6. Wet oxidation of an azo dye: Lumped kinetics in batch and mixed flow reactors

    SciTech Connect (OSTI)

    Donlagic, J.; Levec, J.

    1999-12-01

    Oxidation of a dilute aqueous solution of a model azo dye pollutant (Orange II) was studied in batch and continuous well-mixed (CSTR) reactors. Both reactors operate at 200--250 C, and total pressures up to 50 bar and at oxygen partial pressure from 10 to 30 bar. The model pollutant concentrations were in a range between 100 and 1,000 mg/L, which may be found in industrial wastewaters. The dye oxidation undergoes parallel-consecutive reaction pathways, in which it first decomposes thermally and oxidatively to aromatic intermediates and via organic acids to the final product carbon dioxide. To develop a kinetic equation capable of predicting organic carbon reduction, all organic species present in solution were lumped by total organic carbon (TOC). The lumped oxidation rate in batch reactor exhibited second-order behavior, whereas in the CSTR is was found linearly proportional to its TOC concentration. The lump behavior in batch reactor was dominated by the refractory low molecular mass aliphatic acids formed during the oxidation.

  7. Preparation of extrusions of bulk mixed oxide compounds with high macroporosity and mechanical strength

    DOE Patents [OSTI]

    Flytzani-Stephanopoulos, Maria; Jothimurugesan, Kandaswami

    1990-01-01

    A simple and effective method for producing bulk single and mixed oxide absorbents and catalysts is disclosed. The method yields bulk single oxide and mixed oxide absorbent and catalyst materials which combine a high macroporosity with relatively high surface area and good mechanical strength. The materials are prepared in a pellet form using as starting compounds, calcined powders of the desired composition and physical properties these powders are crushed to broad particle size distribution, and, optionally may be combined with an inorganic clay binder. The necessary amount of water is added to form a paste which is extruded, dried and heat treated to yield and desired extrudate strength. The physical properties of the extruded materials (density, macroporosity and surface area) are substantially the same as the constituent powder is the temperature of the heat treatment of the extrudates is approximately the same as the calcination temperature of the powder. If the former is substantially higher than the latter, the surface area decreases, but the macroporosity of the extrusions remains essentially constant.

  8. Method for acid oxidation of radioactive, hazardous, and mixed organic waste materials

    DOE Patents [OSTI]

    Pierce, Robert A.; Smith, James R.; Ramsey, William G.; Cicero-Herman, Connie A.; Bickford, Dennis F.

    1999-01-01

    The present invention is directed to a process for reducing the volume of low level radioactive and mixed waste to enable the waste to be more economically stored in a suitable repository, and for placing the waste into a form suitable for permanent disposal. The invention involves a process for preparing radioactive, hazardous, or mixed waste for storage by contacting the waste starting material containing at least one organic carbon-containing compound and at least one radioactive or hazardous waste component with nitric acid and phosphoric acid simultaneously at a contacting temperature in the range of about 140.degree. C. to about 210 .degree. C. for a period of time sufficient to oxidize at least a portion of the organic carbon-containing compound to gaseous products, thereby producing a residual concentrated waste product containing substantially all of said radioactive or inorganic hazardous waste component; and immobilizing the residual concentrated waste product in a solid phosphate-based ceramic or glass form.

  9. Reliability of fast reactor mixed-oxide fuel during operational transients

    SciTech Connect (OSTI)

    Boltax, A.; Neimark, L.A.; Tsai, Hanchung ); Katsuragawa, M.; Shikakura, S. . Oarai Engineering Center)

    1991-07-01

    Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 figs., 4 tabs.

  10. Tunable catalytic properties of bi-functional mixed oxides in ethanol conversion to high value compounds

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Ramasamy, Karthikeyan K.; Gray, Michel; Job, Heather; Smith, Colin; Wang, Yong

    2016-02-03

    Here, a highly versatile ethanol conversion process to selectively generate high value compounds is presented here. By changing the reaction temperature, ethanol can be selectively converted to >C2 alcohols/oxygenates or phenolic compounds over hydrotalcite derived bi-functional MgO–Al2O3 catalyst via complex cascade mechanism. Reaction temperature plays a role in whether aldol condensation or the acetone formation is the path taken in changing the product composition. This article contains the catalytic activity comparison between the mono-functional and physical mixture counterpart to the hydrotalcite derived mixed oxides and the detailed discussion on the reaction mechanisms.

  11. MOX Fuel Presentation to Duke Board of Directors

    National Nuclear Security Administration (NNSA)

    PuO 2 with 95% depleted UO 2 - Like LEU fuel pellets, MOX fuel pellets are primarily uranium * Fission power comes primarily from plutonium (Pu 239 ) instead of uranium (U 235 )...

  12. Co-Al mixed metal oxides/carbon nanotubes nanocomposite prepared via a precursor route and enhanced catalytic property

    SciTech Connect (OSTI)

    Fan Guoli; Wang Hui; Xiang Xu; Li Feng

    2013-01-15

    The present work reported the synthesis of Co-Al mixed metal oxides/carbon nanotubes (CoAl-MMO/CNT) nanocomposite from Co-Al layered double hydroxide/CNTs composite precursor (CoAl-LDH/CNT). The materials were characterized by powder X-ray diffraction (XRD), transmission electron microscopy (TEM), low temperature nitrogen adsorption-desorption experiments, thermogravimetric and differential thermal analyses (TG-DTA), Raman spectra and X-ray photoelectron spectroscopy (XPS). The results revealed that in CoAl-MMO/CNT nanocomposite, the nanoparticles of cobalt oxide (CoO) and Co-containing spinel-type complex metal oxides could be well-dispersed on the surface of CNTs, thus forming the heterostructure of CoAl-MMO and CNTs. Furthermore, as-synthesized CoAl-MMO/CNT nanocomposite was utilized as additives for catalytic thermal decomposition of ammonium perchlorate (AP). Compared to those for pure AP and CoAl-MMO, the peak temperature of AP decomposition for CoAl-MMO/CNT was significantly decreased, which is attributed to the novel heterostructure and synergistic effect of multi-component metal oxides of nanocomposite. - Graphical abstract: Hybrid Co-Al mixed metal oxides/carbon nanotubes nanocomposite showed the enhanced catalytic activity in the thermal decomposition of ammonium perchlorate, as compared to carbon nanotubes and pure Co-Al mixed metal oxides. Highlights: Black-Right-Pointing-Pointer Co-Al mixed metal oxides/carbon nanotubes nanocomposite was synthesized. Black-Right-Pointing-Pointer Co-Al mixed metal oxides consisted of cobalt oxide and Co-containing spinels. Black-Right-Pointing-Pointer Nanocomposite exhibited excellent catalytic activity for the decomposition of AP. Black-Right-Pointing-Pointer The superior catalytic property is related to novel heterostructure and composition.

  13. Microstructural Characterization of High Burn-up Mixed Oxide Fast Reactor Fuel

    SciTech Connect (OSTI)

    Melissa C. Teague; Brian P. Gorman; Steven L. Hayes; Douglas L. Porter; Jeffrey King

    2013-10-01

    High burn-up mixed oxide fuel with local burn-ups of 3.423.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 79% FIMA. Samples with burn-ups in excess of 79% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column were observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 35 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.

  14. Structural and spectroscopic properties of high temperature prepared ZrO₂–TiO₂ mixed oxides

    SciTech Connect (OSTI)

    Gionco, Chiara; Battiato, Alfio; Vittone, Ettore; Paganini, Maria Cristina; Giamello, Elio

    2013-05-01

    ZrO₂-TiO₂ mixed oxides of various composition, with the molar fraction of TiO₂ ranging from 0.1% to 15%, have been prepared via sol-gel synthesis and then calcined at 1273 K to check both their thermal stability and physicochemical properties. These solids are usually employed in photocatalytic processes and as active phase supports in heterogeneous catalysis. As indicated by X-ray diffraction and Raman spectroscopy, solid solutions based on Ti ions diluted in the ZrO₂ matrix are formed in the whole range of Ti molar fraction examined. Materials with low Ti loading (0.1%–1%) are basically constituted by the monoclinic phase of ZrO₂ while the tetragonal phase becomes prevalent at 15% of TiO₂ molar fraction. The presence of Ti ions modify the electronic structure of the solid as revealed by investigation of the optical properties. The typical band gap transition of ZrO₂ undergoes, in fact, a red shift roughly proportional to the Ti loading which reach the remarkable value of 1.6 eV for the sample with 10% of molar Ti concentration. Comparing chemical analysis of the solids with XPS data it has been put into evidence that the titanium ions distribution into the solid is not uniform and the concentration of Ti⁴⁺ tend to be higher in subsurface layers than in the crystal bulk. The introduction of titanium ions in the structure increases the reducibility of the solid. Annealing under vacuum at various temperatures causes oxygen depletion with consequent reduction of the solid which shows up mainly in terms of formation of Ti³⁺ reduced centres which are characterized by a typical EPR signal. Ti³⁺ defects forms, as also forecast by theoretical modelling of the solid, as their energy is lower than that of other possible reduced defective centers. The reduced solids are able to transfer electrons to adsorbed oxygen molecules in mild condition resulting in the formation of surface superoxide anions (O₂⁻) which are stabilized on surface Zr

  15. Strength Loss in MA-MOX Green Pellets from Radiation Damage to Binders

    SciTech Connect (OSTI)

    Paul A. Lessing; W.R. Cannon; Gerald W. Egeland; Larry D. Zuck; James K. Jewell; Douglas W. Akers; Gary S. Groenewold

    2013-06-01

    The fracture strength of green Minor Actinides (MA)-MOX pellets containing 75 wt.% DUO2, 20 wt. % PuO2, 3 wt. % AmO2 and 2 wt. % NpO2 was studied as a function of storage time, after mixing in the binder and before sintering, to test the effect of radiation damage on binders. Fracture strength degraded continuously over the 10 days of the study for all three binders studied: PEG binder (Carbowax 8000), microcrystalline wax (Mobilcer X) and Styrene-acrylic copolymer (Duramax B1022) but the fracture strength of Duramax B1022 degraded the least. For instance, for several hours after mixing Carbowax 8000 with MA MOX, the fracture strength of a pellet was reasonably high and pellets were easily handled without breaking but the pellets were too weak to handle after 10 days. Strength measured using diametral compression test showed strength degradation was more rapid in pellets containing 1.0 wt. % Carbowax PEG 8000 compared to those containing only 0.2 wt. %, suggesting that irradiation not only left the binder less effective but also reduced the pellet strength. In contrast the strength of pellets containing Duramax B1022 degraded very little over the 10 day period. It was suggested that the styrene portion of the Duramax B1022 copolymer provided the radiation resistance.

  16. Low temperature photochemical vapor deposition of alloy and mixed metal oxide films

    DOE Patents [OSTI]

    Liu, David K.

    1992-01-01

    Method and apparatus for formation of an alloy thin film, or a mixed metal oxide thin film, on a substrate at relatively low temperatures. Precursor vapor(s) containing the desired thin film constituents is positioned adjacent to the substrate and irradiated by light having wavelengths in a selected wavelength range, to dissociate the gas(es) and provide atoms or molecules containing only the desired constituents. These gases then deposit at relatively low temperatures as a thin film on the substrate. The precursor vapor(s) is formed by vaporization of one or more precursor materials, where the vaporization temperature(s) is selected to control the ratio of concentration of metals present in the precursor vapor(s) and/or the total precursor vapor pressure.

  17. Low temperature photochemical vapor deposition of alloy and mixed metal oxide films

    DOE Patents [OSTI]

    Liu, D.K.

    1992-12-15

    Method and apparatus are described for formation of an alloy thin film, or a mixed metal oxide thin film, on a substrate at relatively low temperatures. Precursor vapor(s) containing the desired thin film constituents is positioned adjacent to the substrate and irradiated by light having wavelengths in a selected wavelength range, to dissociate the gas(es) and provide atoms or molecules containing only the desired constituents. These gases then deposit at relatively low temperatures as a thin film on the substrate. The precursor vapor(s) is formed by vaporization of one or more precursor materials, where the vaporization temperature(s) is selected to control the ratio of concentration of metals present in the precursor vapor(s) and/or the total precursor vapor pressure. 7 figs.

  18. Evaluation of tubular reactor designs for supercritical water oxidation of U.S. Department of Energy mixed waste

    SciTech Connect (OSTI)

    Barnes, C.M.

    1994-12-01

    Supercritical water oxidation (SCWO) is an emerging technology for industrial waste treatment and is being developed for treatment of the US Department of Energy (DOE) mixed hazardous and radioactive wastes. In the SCWO process, wastes containing organic material are oxidized in the presence of water at conditions of temperature and pressure above the critical point of water, 374 C and 22.1 MPa. DOE mixed wastes consist of a broad spectrum of liquids, sludges, and solids containing a wide variety of organic components plus inorganic components including radionuclides. This report is a review and evaluation of tubular reactor designs for supercritical water oxidation of US Department of Energy mixed waste. Tubular reactors are evaluated against requirements for treatment of US Department of Energy mixed waste. Requirements that play major roles in the evaluation include achieving acceptable corrosion, deposition, and heat removal rates. A general evaluation is made of tubular reactors and specific reactors are discussed. Based on the evaluations, recommendations are made regarding continued development of supercritical water oxidation reactors for US Department of Energy mixed waste.

  19. Removal of Hazardous Pollutants from Wastewaters: Applications of TiO 2 -SiO 2 Mixed Oxide Materials

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Rasalingam, Shivatharsiny; Peng, Rui; Koodali, Ranjit T.

    2014-01-01

    The direct release of untreated wastewaters from various industries and households results in the release of toxic pollutants to the aquatic environment. Advanced oxidation processes (AOP) have gained wide attention owing to the prospect of complete mineralization of nonbiodegradable organic substances to environmentally innocuous products by chemical oxidation. In particular, heterogeneous photocatalysis has been demonstrated to have tremendous promise in water purification and treatment of several pollutant materials that include naturally occurring toxins, pesticides, and other deleterious contaminants. In this work, we have reviewed the different removal techniques that have been employed for water purification. In particular, the applicationmore » of TiO 2 -SiO 2 binary mixed oxide materials for wastewater treatment is explained herein, and it is evident from the literature survey that these mixed oxide materials have enhanced abilities to remove a wide variety of pollutants.« less

  20. Mediated electrochemical oxidation treatment for Rocky Flats combustible low-level mixed waste. Final report, FY 1993 and 1994

    SciTech Connect (OSTI)

    Chiba, Z.; Lewis, P.R.; Murguia, L.C.

    1994-09-01

    Mediated Electrochemical Oxidation (MEO) is an aqueous process which destroys hazardous organics by oxidizing a mediator at the anode of an electrochemical cell; the mediator in turn oxidizes the organics within the bulk of the electrolyte. With this process organics can be nearly completely destroyed, that is, the carbon and hydrogen present in the hydrocarbon are almost entirely mineralized to carbon dioxide and water. The MEO process is also capable of dissolving radioactive materials, including difficult-to-dissolve compounds such as plutonium oxide. Hence, this process can treat mixed wastes, by destroying the hazardous organic components of the waste, and dissolving the radioactive components. The radioactive material can be recovered if desired, or disposed of as non-mixed radioactive waste. The process is inherently safe, since the hazardous and radioactive materials are completely contained in the aqueous phase, and the system operates at low temperatures (below 80{degree}C) and at ambient pressures.

  1. LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

    SciTech Connect (OSTI)

    Bronson, M.C.

    1997-10-01

    The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium.

  2. Catalytic propane dehydrogenation over In₂O₃–Ga₂O₃ mixed oxides

    SciTech Connect (OSTI)

    Tan, Shuai; Gil, Laura Briones; Subramanian, Nachal; Sholl, David S.; Nair, Sankar; Jones, Christopher W.; Moore, Jason S.; Liu, Yujun; Dixit, Ravindra S.; Pendergast, John G.

    2015-08-26

    We have investigated the catalytic performance of novel In₂O₃–Ga₂O₃ mixed oxides synthesized by the alcoholic-coprecipitation method for propane dehydrogenation (PDH). Reactivity measurements reveal that the activities of In₂O₃–Ga₂O₃ catalysts are 1–3-fold (on an active metal basis) and 12–28-fold (on a surface area basis) higher than an In₂O₃–Al₂O₃ catalyst in terms of C₃H₈ conversion. The structure, composition, and surface properties of the In₂O₃–Ga₂O₃ catalysts are thoroughly characterized. NH₃-TPD shows that the binary oxide system generates more acid sites than the corresponding single-component catalysts. Raman spectroscopy suggests that catalysts that produce coke of a more graphitic nature suppress cracking reactions, leading to higher C₃H₆ selectivity. Lower reaction temperature also leads to higher C₃H₆ selectivity by slowing down the rate of side reactions. XRD, XPS, and XANES measurements, strongly suggest that metallic indium and In₂O₃ clusters are formed on the catalyst surface during the reaction. The agglomeration of In₂O₃ domains and formation of a metallic indium phase are found to be irreversible under O₂ or H₂ treatment conditions used here, and may be responsible for loss of activity with increasing time on stream.

  3. Plutonium Disposition Program | National Nuclear Security Administrati...

    National Nuclear Security Administration (NNSA)

    to fabricate it into Mixed Oxide (MOX) fuel and irradiate it in existing light water reactors. This approach requires construction of new facilities including the MOX Fuel...

  4. Method of CO and/or CO.sub.2 hydrogenation using doped mixed-metal oxides

    DOE Patents [OSTI]

    Shekhawat, Dushyant; Berry, David A.; Haynes, Daniel J.; Abdelsayed, Victor; Smith, Mark W.; Spivey, James J.

    2015-10-06

    A method of hydrogenation utilizing a reactant gas mixture comprising a carbon oxide and a hydrogen agent, and a hydrogenation catalyst comprising a mixed-metal oxide containing metal sites supported and/or incorporated into the lattice. The mixed-metal oxide comprises a perovskite, a pyrochlore, a fluorite, a brownmillerite, or mixtures thereof doped at the A-site or the B-site. The metal site may comprise a deposited metal, where the deposited metal is a transition metal, an alkali metal, an alkaline earth metal, or mixtures thereof. Contact between the carbon oxide, hydrogen agent, and hydrogenation catalyst under appropriate conditions of temperature, pressure and gas flow rate generate a hydrogenation reaction and produce a hydrogenated product made up of carbon from the carbon oxide and some portion of the hydrogen agent. The carbon oxide may be CO, CO.sub.2, or mixtures thereof and the hydrogen agent may be H.sub.2. In a particular embodiment, the hydrogenated product comprises an alcohol, an olefin, an aldehyde, a ketone, an ester, an oxo-product, or mixtures thereof.

  5. Integrated demonstration of molten salt oxidation with salt recycle for mixed waste treatment

    SciTech Connect (OSTI)

    Hsu, P.C.

    1997-11-01

    Molten Salt Oxidation (MSO) is a thermal, nonflame process that has the inherent capability of completely destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. Lawrence Livermore National Laboratory (LLNL) has prepared a facility and constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are performed under carefully controlled (experimental) conditions. The system consists of a MSO processor with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. This integrated system was designed and engineered based on laboratory experience with a smaller engineering-scale reactor unit and extensive laboratory development on salt recycle and final forms preparation. In this paper we present design and engineering details of the system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is identification of the most suitable waste streams and waste types for MSO treatment.

  6. Improved layered mixed transition metal oxides for Li-ion batteries

    SciTech Connect (OSTI)

    Doeff, Marca M.; Conry, Thomas; Wilcox, James

    2010-03-05

    Recent work in our laboratory has been directed towards development of mixed layered transition metal oxides with general composition Li[Ni, Co, M, Mn]O2 (M=Al, Ti) for Li ion battery cathodes. Compounds such as Li[Ni1/3Co1/3Mn1/3]O2 (often called NMCs) are currently being commercialized for use in consumer electronic batteries, but the high cobalt content makes them too expensive for vehicular applications such as electric vehicles (EV), plug-in hybrid electric vehicles (PHEVs), or hybrid electric vehicles (HEVs). To reduce materials costs, we have explored partial or full substitution of Co with Al, Ti, and Fe. Fe substitution generally decreases capacity and results in poorer rate and cycling behavior. Interestingly, low levels of substitution with Al or Ti improve aspects of performance with minimal impact on energy densities, for some formulations. High levels of Al substitution compromise specific capacity, however, so further improvements require that the Ni and Mn content be increased and Co correspondingly decreased. Low levels of Al or Ti substitution can then be used offset negative effects induced by the higher Ni content. The structural and electrochemical characterization of substituted NMCs is presented in this paper.

  7. Synthesis of metastable rare-earth-iron mixed oxide with the hexagonal crystal structure

    SciTech Connect (OSTI)

    Nishimura, Tatsuya; Hosokawa, Saburo; Masuda, Yuichi; Wada, Kenji; Inoue, Masashi

    2013-01-15

    Rare-earth-iron mixed oxides with the rare earth/iron ratio=1 have either orthorhombic (o-REFeO{sub 3}) or hexagonal (h-REFeO{sub 3}) structure. h-REFeO{sub 3} is a metastable phase and the synthesis of h-REFeO{sub 3} is usually difficult. In this work, the crystallization process of the precursors obtained by co-precipitation and Pechini methods was investigated in detail to synthesize h-REFeO{sub 3}. It was found that the crystallization from amorphous to hexagonal phase and the phase transition from hexagonal to orthorhombic phase occurred at a similar temperature range for rare earth elements with small ionic radii (Er-Lu, Y). For both co-precipitation and Pechini methods, single-phase h-REFeO{sub 3} was obtained by shortening the heating time during calcination process. The hexagonal-to-orthorhombic phase transition took place by a nucleation growth mechanism and vermicular morphology of the thus-formed orthorhombic phase was observed. The hexagonal YbFeO{sub 3} had higher catalytic activity for C{sub 3}H{sub 8} combustion than orthorhombic YbFeO{sub 3}. - Graphical abstract: Although the synthesis of metastable hexagonal REFeO{sub 3} by the conventional method is difficult, we found that this phase is obtained by shortening the heating time of the precursor prepared by co-precipitation method. Highlights: Black-Right-Pointing-Pointer Synthesis of metastable REFeO{sub 3} with hexagonal structure by the co-precipitation method. Black-Right-Pointing-Pointer Hexagonal REFeO{sub 3} is obtained for the rare earth elements with small ionic radii. Black-Right-Pointing-Pointer Hexagonal-to-orthorhombic transformation of REFeO{sub 3}. Black-Right-Pointing-Pointer Catalytic activity of hexagonal REFeO{sub 3} for C{sub 3}H{sub 8} combustion.

  8. Performance of fast reactor mixed-oxide fuels pins during extended overpower transients

    SciTech Connect (OSTI)

    Tsai, H.; Neimark, L.A. ); Asaga, T.; Shikakura, S. )

    1991-02-01

    The Operational Reliability Testing (ORT) program, a collaborative effort between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan, was initiated in 1982 to investigate the behavior of mixed-oxide fuel pin under various slow-ramp transient and duty-cycle conditions. In the first phase of the program, a series of four extended overpower transient tests, with severity sufficient to challenge the pin cladding integrity, was conducted. The objectives of the designated TOPI-1A through -1D tests were to establish the cladding breaching threshold and mechanisms, and investigate the thermal and mechanical effects of the transient on pin behavior. The tests were conducted in EBR-2, a normally steady-state reactor. The modes of transient operation in EBR-2 were described in a previous paper. Two ramp rates, 0.1%/s and 10%/s, were selected to provide a comparison of ramp-rate effects on fuel behavior. The test pins chosen for the series covered a range of design and pre-test irradiation parameters. In the first test (1A), all pins maintained their cladding integrity during the 0.1%/s ramp to 60% peak overpower. Fuel pins with aggressive designs, i.e., high fuel- smear density and/or thin cladding, were, therefore, included in the follow-up 1B and 1C tests to enhance the likelihood of achieving cladding breaching. In the meantime, a higher pin overpower capability, to greater than 100%, was established by increasing the reactor power limit from 62.5 to 75 MWt. In this paper, the significant results of the 1B and 1C tests are presented. 4 refs., 5 figs., 1 tab.

  9. The thermal conductivity of mixed fuel UxPu1-xO2: molecular dynamics simulations

    SciTech Connect (OSTI)

    Liu, Xiang-Yang; Cooper, Michael William Donald; Stanek, Christopher Richard; Andersson, Anders David Ragnar

    2015-10-16

    Mixed oxides (MOX), in the context of nuclear fuels, are a mixture of the oxides of heavy actinide elements such as uranium, plutonium and thorium. The interest in the UO2-PuO2 system arises from the fact that these oxides are used both in fast breeder reactors (FBRs) as well as in pressurized water reactors (PWRs). The thermal conductivity of UO2 fuel is an important material property that affects fuel performance since it is the key parameter determining the temperature distribution in the fuel, thus governing, e.g., dimensional changes due to thermal expansion, fission gas release rates, etc. For this reason it is important to understand the thermal conductivity of MOX fuel and how it differs from UO2. Here, molecular dynamics (MD) simulations are carried out to determine quantitatively, the effect of mixing on the thermal conductivity of UxPu1-xO2, as a function of PuO2 concentrations, for a range of temperatures, 300 – 1500 K. The results will be used to develop enhanced continuum thermal conductivity models for MARMOT and BISON by INL. These models express the thermal conductivity as a function of microstructure state-variables, thus enabling thermal conductivity models with closer connection to the physical state of the fuel.

  10. Nature of Catalytic Active Sites Present on the Surface of Advanced Bulk Tantalum Mixed Oxide Photocatalysts

    SciTech Connect (OSTI)

    Phivilay, Somphonh; Puretzky, Alexander A; Domen, Kazunari Domen; Wachs, Israel

    2013-01-01

    The most active photocatalyst system for water splitting under UV irradiation (270 nm) is the promoted 0.2%NiO/NaTaO3:2%La photocatalyst with optimized photonic efficiency (P.E.) of 56%, but fundamental issues about the nature of the surface catalytic active sites and their involvement in the photocatalytic process still need to be clarified. This is the first study to apply cutting edge surface spectroscopic analyses to determine the surface nature of tantalum mixed oxide photocatalysts. Surface analysis with HR-XPS (1-3nm) and HS-LEIS (0.3nm) spectroscopy indicates that the NiO and La2O3 promoters are concentrated in the surface region of the bulk NaTaO3 phase. The La2O3 is concentrated on the NaTaO3 outermost surface layers while NiO is distributed throughout the NaTaO3 surface region (1-3nm). Raman and UV-vis spectroscopy revealed that the bulk molecular and electronic structures, respectively, of NaTaO3 were not modified by the addition of the La2O3 and NiO promoters, with La2O3 resulting in a slightly more ordered structure. Photoluminescence (PL) spectroscopy reveals that the addition of La2O3 and NiO produces a greater number of electron traps resulting in the suppression of the recombination of excited electrons/holes. In contrast to earlier reports, the La2O3 is only a textural promoter (increasing the BET surface area ~7x by stabilizing smaller NaTaO3 particles), but causes a ~3x decrease in the specific photocatalytic TORs ( mol H2/m2/h) rate because surface La2O3 blocks exposed catalytic active NaTaO3 sites. The NiO promoter was found to be a potent electronic promoter that enhances the NaTaO3 surface normalized TORs by a factor of ~10-50 and TOF by a factor of ~10. The level of NiO promotion is the same in the absence and presence of La2O3 demonstrating that there is no promotional synergistic interaction between the NiO and La2O3 promoters. This study demonstrates the important contributions of the photocatalyst surface properties to the fundamental

  11. Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

    SciTech Connect (OSTI)

    Ellis, Ronald James

    2015-04-01

    The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) during cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.

  12. Note: Application of CR-39 plastic nuclear track detectors for quality assurance of mixed oxide fuel pellets

    SciTech Connect (OSTI)

    Kodaira, S. Kurano, M.; Hosogane, T.; Ishikawa, F.; Kageyama, T.; Sato, M.; Kayano, M.; Yasuda, N.

    2015-05-15

    A CR-39 plastic nuclear track detector was used for quality assurance of mixed oxide fuel pellets for next-generation nuclear power plants. Plutonium (Pu) spot sizes and concentrations in the pellets are significant parameters for safe use in the plants. We developed an automatic Pu detection system based on dense α-radiation tracks in the CR-39 detectors. This system would greatly improve image processing time and measurement accuracy, and will be a powerful tool for rapid pellet quality assurance screening.

  13. Final assessment of MOX fuel performance experiment with Japanese PWR specification fuel in the HBWR

    SciTech Connect (OSTI)

    Fujii, Hajime; Teshima, Hideyuki; Kanasugi, Katsumasa; Kosaka, Yuji; Arakawa, Yasushi

    2007-07-01

    In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pu-fissile enrichment of about 6 wt% had been irradiated in the HBWR from 1995 to 2006. The peak burn-up of MOX pellet achieved 72 GWd/tM. In this test, fuel centerline temperature, rod internal pressure, stack length and cladding length were measured for MOX fuel and UO{sub 2} fuel as reference. MOX fuel temperature is confirmed to have no significant difference in comparison with UO{sub 2}, taking into account of adequate thermal conductivity degradation due to PuO{sub 2} addition and burn-up development. And the measured fuel temperature agrees well with FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly greater than UO{sub 2} based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behavior. MOX fuel swelling rate agrees well with solid swelling rate in the literature. Cladding elongation data shows onset of PCMI in high power region. (authors)

  14. A Neutronic Analysis of TRU Recycling in PWRs Loaded with MOX-UE Fuel (MOX with U-235 Enriched U Support)

    SciTech Connect (OSTI)

    G. Youinou; S. Bays

    2009-05-01

    This report presents the results of a study dealing with the homogeneous recycling of either Pu or Pu+Np or Pu+Np+Am or Pu+Np+Am+Cm in PWRs using MOX-UE fuel, i.e. standard MOX fuel with a U235 enriched uranium support instead of the standard tail uranium (0.25%) for standard MOX fuel. This approach allows to multirecycle Pu or TRU (Pu+MA) as long as U235 is available, by keeping the Pu or TRU content in the fuel constant and at a value ensuring a negative moderator void coefficient (i.e. the loss of the coolant brings imperatively the reactor to a subcritical state). Once this value is determined, the U235 enrichment of the MOX-UE fuel is adjusted in order to reach the target burnup (51 GWd/t in this study).

  15. Final Report: Fiscal Year 1997 demonstration of omnivorous non-thermal mixed waste treatment: Direct chemical oxidation of organic solids and liquids using peroxydisulfate

    SciTech Connect (OSTI)

    Cooper, J.F.

    1998-01-01

    Direct Chemical Oxidation (DCO) is a non-thermal, ambient pressure, aqueous-based technology for the oxidative destruction of the organic components of hazardous or mixed waste streams. The process has been developed for applications in waste treatment, chemical demilitarization and decontamination at LLNL since 1992. The process uses solutions of the peroxydisulfate ion (typically sodium or ammonium salts) to completely mineralize the organics to carbon dioxide and water. The expended oxidant may be electrolytically regenerated to minimize secondary waste. The paper briefly describes: free radical and secondary oxidant formation; electrochemical regeneration; offgas stream; and throughput.

  16. ANALYTICAL RESULTS FOR MOX COLEMANITE SAMPLES RECEIVED ON JULY 22, 2013

    SciTech Connect (OSTI)

    Reigel, M.; Best, D.

    2014-05-19

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory (SRNL) is tasked with measuring the boron oxide content of the colemanite raw aggregate material prior to it being mixed into the concrete. SRNL received ten samples of colemanite for analysis on July 22, 2013. The elemental boron content of each sample was measured according to ASTM C 1301. The boron oxide content was calculated using the oxide conversion factor for boron.

  17. ANALYTICAL RESULTS FOR MOX COLEMANITE SAMPLES RECEIVED ON JULY 22, 2013

    SciTech Connect (OSTI)

    Reigel, M.; Best, D.

    2013-08-13

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory (SRNL) is tasked with measuring the boron oxide content of the colemanite raw aggregate material prior to it being mixed into the concrete. SRNL received ten samples of colemanite for analysis on July 22, 2013. The elemental boron content of each sample was measured according to ASTM C 1301. The boron oxide content was calculated using the oxide conversion factor for boron.

  18. Melting temperatures of the ZrO{sub 2}-MOX system

    SciTech Connect (OSTI)

    Uchida, T.; Hirooka, S.; Kato, M.; Morimoto, K.; Sugata, H.; Shibata, K.; Sato, D.

    2013-07-01

    Severe accidents occurred at the Fukushima Daiichi Nuclear Power Plant Units 1-3 on March 11, 2011. MOX fuels were loaded in the Unit 3. For the thermal analysis of the severe accident, melting temperature and phase state of MOX corium were investigated. The simulated coriums were prepared from 4%Pu-containing MOX, 8%Pu-containing MOX and ZrO{sub 2}. Then X-ray diffraction, density and melting temperature measurements were carried out as a function of zirconium and plutonium contents. The cubic phase was observed in the 25%Zr-containing corium and the tetragonal phase was observed in the 50% and 75%Zr-containing coria. The lattice parameter and density monotonically changed with Pu content. Melting temperature increased with increasing Pu content; melting temperature were estimated to be 2932 K for 4%Pu MOX corium and 3012 K for 8%Pu MOX corium in the 25%ZrO{sub 2}-MOX system. The lowest melting temperature was observed for 50%Zr-containing corium. (authors)

  19. Use of impure inert gases in the controlled heating and cooling of mixed conducting metal oxide materials

    DOE Patents [OSTI]

    Carolan, Michael Francis; Bernhart, John Charles

    2012-08-21

    Method for processing an article comprising mixed conducting metal oxide material. The method comprises contacting the article with an oxygen-containing gas and either reducing the temperature of the oxygen-containing gas during a cooling period or increasing the temperature of the oxygen-containing gas during a heating period; during the cooling period, reducing the oxygen activity in the oxygen-containing gas during at least a portion of the cooling period and increasing the rate at which the temperature of the oxygen-containing gas is reduced during at least a portion of the cooling period; and during the heating period, increasing the oxygen activity in the oxygen-containing gas during at least a portion of the heating period and decreasing the rate at which the temperature of the oxygen-containing gas is increased during at least a portion of the heating period.

  20. LITERATURE REVIEW FOR OXALATE OXIDATION PROCESSES AND PLUTONIUM OXALATE SOLUBILITY

    SciTech Connect (OSTI)

    Nash, C.

    2012-02-03

    A literature review of oxalate oxidation processes finds that manganese(II)-catalyzed nitric acid oxidation of oxalate in precipitate filtrate is a viable and well-documented process. The process has been operated on the large scale at Savannah River in the past, including oxidation of 20 tons of oxalic acid in F-Canyon. Research data under a variety of conditions show the process to be robust. This process is recommended for oxalate destruction in H-Canyon in the upcoming program to produce feed for the MOX facility. Prevention of plutonium oxalate precipitation in filtrate can be achieved by concentrated nitric acid/ferric nitrate sequestration of oxalate. Organic complexants do not appear practical to sequester plutonium. Testing is proposed to confirm the literature and calculation findings of this review at projected operating conditions for the upcoming campaign. H Canyon plans to commence conversion of plutonium metal to low-fired plutonium oxide in 2012 for eventual use in the Mixed Oxide Fuel (MOX) Facility. The flowsheet includes sequential operations of metal dissolution, ion exchange, elution, oxalate precipitation, filtration, and calcination. All processes beyond dissolution will occur in HB-Line. The filtration step produces an aqueous filtrate that may have as much as 4 M nitric acid and 0.15 M oxalate. The oxalate needs to be removed from the stream to prevent possible downstream precipitation of residual plutonium when the solution is processed in H Canyon. In addition, sending the oxalate to the waste tank farm is undesirable. This report addresses the processing options for destroying the oxalate in existing H Canyon equipment.

  1. Cladding inner surface wastage for mixed-oxide liquid metal reactor fuel pins

    SciTech Connect (OSTI)

    Lawrence, L.A.; Bard, F.E.; Cannon, N.S.

    1990-11-01

    Cladding inner surface wastage was measured on reference fuel pins with stainless steel and D9 cladding irradiated beyond goal burnup in the Fast Flux Test Facility. Measurements were compared to the Experimental Breeder Reactor No. 2 based fuel-cladding chemical interaction correlation developed for uranium-plutonium oxide fuels with 20% cold-worked stainless steel cladding. The fuel-cladding chemical interaction was also measured in fuel pins irradiated with HT9 cladding. Comparison of the measurements with the design correlation showed the correlation adequately accounted for the extent of interaction in the Fast Flux Test Facility fuel pins with cold-worked stainless steel D9, and HT9 cladding. 9 refs., 6 figs.

  2. Molybdenum-based additives to mixed-metal oxides for use in hot gas cleanup sorbents for the catalytic decomposition of ammonia in coal gases

    DOE Patents [OSTI]

    Ayala, Raul E.

    1993-01-01

    This invention relates to additives to mixed-metal oxides that act simultaneously as sorbents and catalysts in cleanup systems for hot coal gases. Such additives of this type, generally, act as a sorbent to remove sulfur from the coal gases while substantially simultaneously, catalytically decomposing appreciable amounts of ammonia from the coal gases.

  3. Options for converting excess plutonium to feed for the MOX fuel fabrication facility

    SciTech Connect (OSTI)

    Watts, Joe A; Smith, Paul H; Psaras, John D; Jarvinen, Gordon D; Costa, David A; Joyce, Jr., Edward L

    2009-01-01

    The storage and safekeeping of excess plutonium in the United States represents a multibillion-dollar lifecycle cost to the taxpayers and poses challenges to National Security and Nuclear Non-Proliferation. Los Alamos National Laboratory is considering options for converting some portion of the 13 metric tons of excess plutonium that was previously destined for long-term waste disposition into feed for the MOX Fuel Fabrication Facility (MFFF). This approach could reduce storage costs and security ri sks, and produce fuel for nuclear energy at the same time. Over the course of 30 years of weapons related plutonium production, Los Alamos has developed a number of flow sheets aimed at separation and purification of plutonium. Flow sheets for converting metal to oxide and for removing chloride and fluoride from plutonium residues have been developed and withstood the test oftime. This presentation will address some potential options for utilizing processes and infrastructure developed by Defense Programs to transform a large variety of highly impure plutonium into feedstock for the MFFF.

  4. High Catalytic Activity of Au/CeOx/TiO2(110) Controlled by the Nature of the Mixed Metal Oxide at the Nanometer Level

    SciTech Connect (OSTI)

    Park, J.; Graciani, J; Evans, J; Stacchiola, D; Ma, S; Liu, P; Nambu, A; Sanz, J; Hrbek, J; et. al.

    2009-01-01

    Mixed-metal oxides play a very important role in many areas of chemistry, physics, materials science, and geochemistry. Recently, there has been a strong interest in understanding phenomena associated with the deposition of oxide nanoparticles on the surface of a second (host) oxide. Here, scanning tunneling microscopy, photoemission, and density-functional calculations are used to study the behavior of ceria nanoparticles deposited on a TiO2(110) surface. The titania substrate imposes nontypical coordination modes on the ceria nanoparticles. In the CeOx/TiO2(110) systems, the Ce cations adopt an structural geometry and an oxidation state (+3) that are quite different from those seen in bulk ceria or for ceria nanoparticles deposited on metal substrates. The increase in the stability of the Ce3+ oxidation state leads to an enhancement in the chemical and catalytic activity of the ceria nanoparticles. The codeposition of ceria and gold nanoparticles on a TiO2(110) substrate generates catalysts with an extremely high activity for the production of hydrogen through the water-gas shift reaction (H2O + CO ? H2 + CO2) or for the oxidation of carbon monoxide (2CO + O2 ? 2CO2). The enhanced stability of the Ce3+ state is an example of structural promotion in catalysis described here on the atomic level. The exploration of mixed-metal oxides at the nanometer level may open avenues for optimizing catalysts through stabilization of unconventional surface structures with special chemical activity.

  5. Programmatic and technical requirements for the FMDP fresh MOX fuel transport package

    SciTech Connect (OSTI)

    Ludwig, S. B.; Michelhaugh, R. D.; Pope, R. B.; Shappert, L. B.; Singletary, B. H.; Chae, S. M.; Parks, C. V.; Broadhead, B. L.; Schmid, S. P.; Cowart, C. G.

    1997-12-01

    This document is intended to guide the designers of the package to all pertinent regulatory and other design requirements to help ensure the safe and efficient transport of the weapons-grade (WG) fresh MOX fuel under the Fissile Materials Disposition Program. To accomplish the disposition mission using MOX fuel, the unirradiated MOX fuel must be transported from the MOX fabrication facility to one or more commercial reactors. Because the unirradiated fuel contains large quantities of plutonium and is not sufficient radioactive to create a self-protecting barrier to deter the material from theft, DOE intends to use its fleet of safe secure trailers (SSTs) to provide the necessary safeguards and security for the material in transit. In addition to these requirements, transport of radioactive materials must comply with regulations of the Department of Transportation and the Nuclear Regulatory Commission (NRC). In particular, NRC requires that the packages must meet strict performance requirements. The requirements for shipment of MOX fuel (i.e., radioactive fissile materials) specify that the package design is certified by NRC to ensure the materials contained in the packages are not released and remain subcritical after undergoing a series of hypothetical accident condition tests. Packages that pass these tests are certified by NRC as a Type B fissile (BF) package. This document specifies the programmatic and technical design requirements a package must satisfy to transport the fresh MOX fuel assemblies.

  6. Atomic-Resolution Visualization of Distinctive Chemical Mixing Behavior of Ni, Co and Mn with Li in Layered Lithium Transition-Metal Oxide Cathode Materials

    SciTech Connect (OSTI)

    Yan, Pengfei; Zheng, Jianming; Lv, Dongping; Wei, Yi; Zheng, Jiaxin; Wang, Zhiguo; Kuppan, Saravanan; Yu, Jianguo; Luo, Langli; Edwards, Danny J.; Olszta, Matthew J.; Amine, Khalil; Liu, Jun; Xiao, Jie; Pan, Feng; Chen, Guoying; Zhang, Jiguang; Wang, Chong M.

    2015-07-06

    Capacity and voltage fading of layer structured cathode based on lithium transition metal oxide is closely related to the lattice position and migration behavior of the transition metal ions. However, it is scarcely clear about the behavior of each of these transition metal ions. We report direct atomic resolution visualization of interatomic layer mixing of transition metal (Ni, Co, Mn) and lithium ions in layer structured oxide cathodes for lithium ion batteries. Using chemical imaging with aberration corrected scanning transmission electron microscope (STEM) and DFT calculations, we discovered that in the layered cathodes, Mn and Co tend to reside almost exclusively at the lattice site of transition metal (TM) layer in the structure or little interlayer mixing with Li. In contrast, Ni shows high degree of interlayer mixing with Li. The fraction of Ni ions reside in the Li layer followed a near linear dependence on total Ni concentration before reaching saturation. The observed distinctively different behavior of Ni with respect to Co and Mn provides new insights on both capacity and voltage fade in this class of cathode materials based on lithium and TM oxides, therefore providing scientific basis for selective tailoring of oxide cathode materials for enhanced performance.

  7. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint US/Russian Progress Report for Fiscal 1997. Volume 3 - Calculations Performed in the Russian Federation

    SciTech Connect (OSTI)

    1998-06-01

    This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the Russian Federation during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the contaminated benchmarks that the United States and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors.

  8. Mixed N-Heterocyclic Carbene-Bis(oxazolinyl)borato Rhodium and Iridium Complexes in Photochemical and Thermal Oxidative Addition Reactions

    SciTech Connect (OSTI)

    Xu, Songchen; Manna, Kuntal; Ellern, Arkady; Sadow, Aaron D

    2014-12-08

    In order to facilitate oxidative addition chemistry of fac-coordinated rhodium(I) and iridium(I) compounds, carbenebis(oxazolinyl)phenylborate proligands have been synthesized and reacted with organometallic precursors. Two proligands, PhB(OxMe2)2(ImtBuH) (H[1]; OxMe2 = 4,4-dimethyl-2-oxazoline; ImtBuH = 1-tert-butylimidazole) and PhB(OxMe2)2(ImMesH) (H[2]; ImMesH = 1-mesitylimidazole), are deprotonated with potassium benzyl to generate K[1] and K[2], and these potassium compounds serve as reagents for the synthesis of a series of rhodium and iridium complexes. Cyclooctadiene and dicarbonyl compounds {PhB(OxMe2)2ImtBu}Rh(?4-C8H12) (3), {PhB(OxMe2)2ImMes}Rh(?4-C8H12) (4), {PhB(OxMe2)2ImMes}Rh(CO)2 (5), {PhB(OxMe2)2ImMes}Ir(?4-C8H12) (6), and {PhB(OxMe2)2ImMes}Ir(CO)2 (7) are synthesized along with ToMM(?4-C8H12) (M = Rh (8); M = Ir (9); ToM = tris(4,4-dimethyl-2-oxazolinyl)phenylborate). The spectroscopic and structural properties and reactivity of this series of compounds show electronic and steric effects of substituents on the imidazole (tert-butyl vs mesityl), effects of replacing an oxazoline in ToM with a carbene donor, and the influence of the donor ligand (CO vs C8H12). The reactions of K[2] and [M(?-Cl)(?2-C8H14)2]2 (M = Rh, Ir) provide {?4-PhB(OxMe2)2ImMes?CH2}Rh(?-H)(?-Cl)Rh(?2-C8H14)2 (10) and {PhB(OxMe2)2ImMes}IrH(?3-C8H13) (11). In the former compound, a spontaneous oxidative addition of a mesityl ortho-methyl to give a mixed-valent dirhodium species is observed, while the iridium compound forms a monometallic allyl hydride. Photochemical reactions of dicarbonyl compounds 5 and 7 result in CH bond oxidative addition providing the compounds {?4-PhB(OxMe2)2ImMes?CH2}RhH(CO) (12) and {PhB(OxMe2)2ImMes}IrH(Ph)CO (13). In 12, oxidative addition results in cyclometalation of the mesityl ortho-methyl similar to 10, whereas the iridium compound reacts with the benzene solvent to give a rare crystallographically characterized cis-[Ir](H)(Ph) complex

  9. ZIRCONIA-BASED MIXED POTENTIAL CARBON MONOXIDE/HYDROCARBON SENSORS WITH LANTHANUM MAGNESIUM OXIDE, AND TERBIUM-DOPED YTTRIUM STABILIZED ZIRCONIA ELECTRODES

    SciTech Connect (OSTI)

    E. L. BROSHA; R. MUKUNDAN; ET AL

    2000-10-01

    We have investigated the performance of dual metal oxide electrode mixed potential sensors in an engine-out, dynamometer environment. Sensors were fabricated by sputtering thin films of LaMnO{sub 3} and Tb-doped YSZ onto YSZ electrolyte. Au gauze held onto the metal oxide thin films with Au ink was used for current collection. The exhaust gas from a 4.8L, V8 engine operated in open loop, steady-state mode around stoichiometry at 1500 RPM and 50 Nm. The sensor showed a stable EMF response (with no hysteresis) to varying concentrations of total exhaust gas HC content. The sensor response was measured at 620 and 670 C and shows temperature behavior characteristic of mixed potential-type sensors. The results of these engine-dynamometer tests are encouraging; however, the limitations associated with Au current collection present the biggest impediment to automotive use.

  10. Critical Experiments that Simulated Damp MOX Powders - Do They Meet the Need?

    SciTech Connect (OSTI)

    J. Blair Briggs; Dr. Ali Nouri; Dr. Claes Nordborg

    2005-09-01

    The OECD Nuclear Energy Agency (NEA) Working Party on Nuclear Criticality Safety (WPNCS) identified the MOX fuel manufacturing process as an area in which there is a need for additional integral benchmark data. The specific need focused on damp MOX powders. The WPNCS was ultimately asked by the NEA Nuclear Science Committee (NSC) to provide the framework for the selection and performance of new experiments that fill the identified need. A set of criteria was established to enable uniform comparison of experimental proposals with generic MOX application data. Criteria were established for five general characteristics: (1) neutronic parameters, (2) type of experiments, (3) financial aspects, (4) schedule, and (5) other considerations. Proposals were judged most importantly on their ability to match the neutronic parameters of predetermined MOX applications. The neutronic parameters that formed the basis for comparison included core average values (not local values) for flux, fission and capture rate; detailed balance data (fission and capture) for the main isotopes (Actinides, H and O); sensitivity coefficients to important nuclear reactions (fission, capture, elastic and inelastic scatter, nu-bar, mu-bar) for all uranium and plutonium isotopes, hydrogen, and oxygen; sensitivity profiles to the main nuclear reactions for uranium and plutonium isotopes; energy of average lethargy causing fission; and the average fission group energy. The focus of this paper is on the definition of the need; the neutronics criteria established to assess which, if any, of three proposed MOX experimental programs best meet the need; and the actual assessment of the proposed experimental programs.

  11. IDENTIFYING IMPURITIES IN SURPLUS NON PIT PLUTONIUM FEEDS FOR MOX OR ALTERNATIVE DISPOSITION

    SciTech Connect (OSTI)

    Allender, J; Moore, E

    2010-07-14

    This report provides a technical basis for estimating the level of corrosion products in materials stored in DOE-STD-3013 containers based on extrapolating available chemical sample results. The primary focus is to estimate the levels of nickel, iron, and chromium impurities in plutonium-bearing materials identified for disposition in the United States Mixed Oxide fuel process.

  12. Adsorption of propane, isopropyl, and hydrogen on cluster models of the M1 phase of Mo-V-Te-Nb-O mixed metal oxide catalyst

    SciTech Connect (OSTI)

    Govindasamy, Agalya; Muthukumar, Kaliappan; Yu, Junjun; Xu, Ye; Guliants, Vadim V.

    2010-01-01

    The Mo-V-Te-Nb-O mixed metal oxide catalyst possessing the M1 phase structure is uniquely capable of directly converting propane into acrylonitrile. However, the mechanism of this complex eight-electron transformation, which includes a series of oxidative H-abstraction and N-insertion steps, remains poorly understood. We have conducted a density functional theory study of cluster models of the proposed active and selective site for propane ammoxidation, including the adsorption of propane, isopropyl (CH{sub 3}CHCH{sub 3}), and H which are involved in the first step of this transformation, that is, the methylene C-H bond scission in propane, on these active site models. Among the surface oxygen species, the telluryl oxo (Te=O) is found to be the most nucleophilic. Whereas the adsorption of propane is weak regardless of the MO{sub x} species involved, isopropyl and H adsorption exhibits strong preference in the order of Te=O > V=O > bridging oxygens > empty Mo apical site, suggesting the importance of TeO{sub x} species for H abstraction. The adsorption energies of isopropyl and H and consequently the reaction energy of the initial dehydrogenation of propane are strongly dependent on the number of ab planes included in the cluster, which points to the need to employ multilayer cluster models to correctly capture the energetics of surface chemistry on this mixed metal oxide catalyst.

  13. Slide 1

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    MOX Update Kelly Trice, President & COO Shaw AREVA MOX Services, LLC. 2 What is MOX? * Mission - Convert at least 34 metric tons of U.S. weapons-grade plutonium to mixed oxide (MOX) fuel for use in commercial power reactors - Implements international agreement with Russia where they will also dispose of 34 metric tons of surplus weapons-grade plutonium 3 MOX Safety Performance * Safety performance remains excellent - Over 16 million hours worked since start of construction - Currently have

  14. Amperometric Biosensors Based on Carbon Paste Electrodes Modified with Nanostructured Mixed-valence Manganese Oxides and Glucose Oxidase

    SciTech Connect (OSTI)

    Cui, Xiaoli; Liu, Guodong; Lin, Yuehe

    2005-06-01

    Nanostructured multivalent manganese oxides octahedral molecular sieve (OMS), including cryptomelane-type manganese oxides and todorokite-type manganese oxides, were synthesized and evaluated for chemical sensing and biosensing at low operating potential. Both cryptomelane-type manganese oxides and todorokite-type manganese oxides are nanofibrous crystals with sub-nanometer open tunnels that provide a unique property for sensing applications. The electrochemical and electrocatalytic performance of OMS for the oxidation of H2O2 have been compared. Both cryptomelane-type manganese oxides and todorokite-type manganese oxides can be used to fabricate sensitive H2O2 sensors. Amperometric glucose biosensors are constructed by bulk modification of carbon paste electrodes (CPEs) with glucose oxidase as a biocomponent and nanostructured OMS as a mediator. A Nafion thin film was applied as an immobilization/encapsulation and protective layer. The biosensors were evaluated as an amperometric glucose detector at phosphate buffer solution with a pH 7.4 at an operating potential of 0.3 V (vs. Ag/AgCl). The biosensor is characterized by a well-reproducible amperometric response, linear signal-to-glucose concentration range up to 3.5 mM and 1.75 mM, and detection limits (S/N = 3) of 0.1 mM and 0.05 mM for todorokite-type manganese oxide and cryptomelane-type manganese oxide modified electrodes, respectively. The biosensors based on OMS exhibit considerable good reproducibility and stability, and the construction and renewal are simple and inexpensive.

  15. Multirecycling of Plutonium from LMFBR Blanket in Standard PWRs Loaded with MOX Fuel

    SciTech Connect (OSTI)

    Sonat Sen; Gilles Youinou

    2013-02-01

    It is now well-known that, from a physics standpoint, Pu, or even TRU (i.e. Pu+M.A.), originating from LEU fuel irradiated in PWRs can be multirecycled also in PWRs using MOX fuel. However, the degradation of the isotopic composition during irradiation necessitates using enriched U in conjunction with the MOX fuel either homogeneously or heterogeneously to maintain the Pu (or TRU) content at a level allowing safe operation of the reactor, i.e. below about 10%. The study is related to another possible utilization of the excess Pu produced in the blanket of a LMFBR, namely in a PWR(MOX). In this case the more Pu is bred in the LMFBR, the more PWR(MOX) it can sustain. The important difference between the Pu coming from the blanket of a LMFBR and that coming from a PWR(LEU) is its isotopic composition. The first one contains about 95% of fissile isotopes whereas the second one contains only about 65% of fissile isotopes. As it will be shown later, this difference allows the PWR fed by Pu from the LMFBR blanket to operate with natural U instead of enriched U when it is fed by Pu from PWR(LEU)

  16. Oxide

    SciTech Connect (OSTI)

    2014-07-15

    Oxide is a modular framework for feature extraction and analysis of executable files. Oxide is useful in a variety of reverse engineering and categorization tasks relating to executable content.

  17. Method for making fine and ultrafine spherical particles of zirconium titanate and other mixed metal oxide systems

    DOE Patents [OSTI]

    Hu, Michael Z.

    2006-05-23

    Disclosed is a method for making amorphous spherical particles of zirconium titanate and crystalline spherical particles of zirconium titanate comprising the steps of mixing an aqueous solution of zirconium salt and an aqueous solution of titanium salt into a mixed solution having equal moles of zirconium and titanium and having a total salt concentration in the range from 0.01 M to about 0.5 M. A stearic dispersant and an organic solvent is added to the mixed salt solution, subjecting the zirconium salt and the titanium salt in the mixed solution to a coprecipitation reaction forming a solution containing amorphous spherical particles of zirconium titanate wherein the volume ratio of the organic solvent to aqueous part is in the range from 1 to 5. The solution of amorphous spherical particles is incubated in an oven at a temperature .ltoreq.100.degree. C. for a period of time .ltoreq.24 hours converting the amorphous particles to fine or ultrafine crystalline spherical particles of zirconium titanate.

  18. Comparative XRPD and XAS study of the impact of the synthesis process on the electronic and structural environments of uranium–americium mixed oxides

    SciTech Connect (OSTI)

    Prieur, D.; Lebreton, F.; Somers, J.; Delahaye, T.

    2015-10-15

    Uranium–americium mixed oxides are potential compounds to reduce americium inventory in nuclear waste via a partitioning and transmutation strategy. A thorough assessment of the oxygen-to-metal ratio is paramount in such materials as it determines the important underlying electronic structure and phase relations, affecting both thermal conductivity of the material and its interaction with the cladding and coolant. In 2011, various XAS experiments on U{sub 1−x}Am{sub x}O{sub 2±δ} samples prepared by different synthesis methods have reported contradictory results on the charge distribution of U and Am. This work alleviates this discrepancy. The XAS results confirm that, independently of the synthesis process, the reductive sintering of U{sub 1−x}Am{sub x}O{sub 2±δ} leads to the formation of similar fluorite solid solution indicating the presence of Am{sup +III} and U{sup +V} in equimolar proportions. - Graphical abstract: Formation of (U{sup IV/V},Am{sup III})O{sup 2} solid solution by sol–gel and by powder metallurgy. - Highlights: • Uranium–americium mixed oxides were synthesized by sol–gel and powder metallurgy. • Fluorite solid solutions with similar local environment have been obtained. • U{sup V} and Am{sup III} are formed in equimolar proportions.

  19. Synthesis and characterization of Th{sub 1−x}Ln{sub x}O{sub 2−x/2} mixed-oxides

    SciTech Connect (OSTI)

    Horlait, D.; Clavier, N.

    2012-12-15

    Graphical abstract: Display Omitted Highlights: ► Th{sub 1−x}Ln{sub x}O{sub 2−x/2} (Ln = Nd, Sm, Gd, Dy, Yb) were produced from oxalates precursors. ► The isomorphic thermal conversion of oxalates to oxides were followed by HT-ESEM. ► The stabilization of the ThO{sub 2} fluorite structure up to x = 0.4 was evidenced by XRD. ► Neodymium incorporation limits in solid solutions were determined by XRD and μ-Raman. ► Lanthanide incorporation mechanism in ThO{sub 2} structure was formally determined by μ-Raman. -- Abstract: Several Th{sub 1−x}Ln{sub x}O{sub 2−x/2} mixed-oxides (Ln = Nd, Sm, Gd, Dy, Er or Yb) were prepared from oxalate precursors. Structural and microstructural investigations of both initial precursors and resulting oxides were undertaken by XRD, μ-Raman spectroscopy and SEM, with a particular attention paid to the Th{sub 1−x}Nd{sub x}O{sub 2−x/2} series. For oxides, XRD and μ-Raman agreed well with the stabilization of the Fm3{sup ¯}m structure up to x{sub Nd} = 0.4, thanks to the concomitant creation of oxygen vacancies, as also confirmed by μ-Raman. Then, a structural transition to the Ia3{sup ¯} superstructure occurred. For x{sub Nd} ≥ 0.49, mixed-oxides with an additional hexagonal Nd{sub 2}O{sub 3} phase were prepared. Besides, the unit cell parameter of the Th{sub 1−x}Ln{sub x}O{sub 2−x/2} series followed a quadratic relation versus the x substitution rate as a result of the combination between modifications of cationic radius and cation coordination, and the decrease of O-O repulsion linked to the presence of oxygen vacancies.

  20. QUALIFICATION OF THE SAVANNAH RIVER SITE 252CF SHUFFLER FOR RECEIPT VERIFICATION MEASUREMENTS OF MIXED U-PU OXIDES STORED IN 9975 SHIPPING CONTAINERS

    SciTech Connect (OSTI)

    Dubose, F.

    2011-05-26

    To extend their ability to perform accountability and verification measurements of {sup 235}U in a U-Pu oxide matrix, the K-Area Material Storage facility commissioned the development and construction of a Passive/Active {sup 252}Cf Shuffler. A series of {sup 252}Cf, PuO{sub 2}, and U-Pu oxide standards, in addition to a single U{sub 3}O{sub 8} standard, were measured to characterize and calibrate the shuffler. Accompanying these measurements were simulations using MCNP5/MCNPX, aimed at isolating the neutron countrate contributions for each of the isotopes present. Two calibration methods for determining the {sup 235}U content in mixed UPu oxide were then developed, yielding comparable results. The first determines the {sup 235}U mass by estimating the {sup 239}Pu/{sup 235}U ratio-dependent contributions from the primary delayed neutron contributors. The second defines an average linear response based on the {sup 235}U and {sup 239}Pu mass contents. In each case, it was observed that self-shielding due to {sup 235}U mass has a large influence on the observed rates, requiring bounds on the applicable limits of each calibration method.

  1. Application of wavelet scaling function expansion continuous-energy resonance calculation method to MOX fuel problem

    SciTech Connect (OSTI)

    Yang, W.; Wu, H.; Cao, L.

    2012-07-01

    More and more MOX fuels are used in all over the world in the past several decades. Compared with UO{sub 2} fuel, it contains some new features. For example, the neutron spectrum is harder and more resonance interference effects within the resonance energy range are introduced because of more resonant nuclides contained in the MOX fuel. In this paper, the wavelets scaling function expansion method is applied to study the resonance behavior of plutonium isotopes within MOX fuel. Wavelets scaling function expansion continuous-energy self-shielding method is developed recently. It has been validated and verified by comparison to Monte Carlo calculations. In this method, the continuous-energy cross-sections are utilized within resonance energy, which means that it's capable to solve problems with serious resonance interference effects without iteration calculations. Therefore, this method adapts to treat the MOX fuel resonance calculation problem natively. Furthermore, plutonium isotopes have fierce oscillations of total cross-section within thermal energy range, especially for {sup 240}Pu and {sup 242}Pu. To take thermal resonance effect of plutonium isotopes into consideration the wavelet scaling function expansion continuous-energy resonance calculation code WAVERESON is enhanced by applying the free gas scattering kernel to obtain the continuous-energy scattering source within thermal energy range (2.1 eV to 4.0 eV) contrasting against the resonance energy range in which the elastic scattering kernel is utilized. Finally, all of the calculation results of WAVERESON are compared with MCNP calculation. (authors)

  2. Microsoft Word - Document in Microsoft Internet Explorer

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Status of the Mixed Oxide Fuel Fabrication Facility DOE/IG-0713 December 2005 STATUS OF THE MIXED OXIDE FUEL FABRICATION FACILITY TABLE OF CONTENTS MOX Facility Design Costs Details of Finding 1 Recommendations 5 Comments 6 Appendices Objective, Scope, and Methodology 9 Prior Audit Reports 11 Management Comments 12 MOX Facility Design Costs Page 1 Details of Finding Design and The audit disclosed that the cost of the Mixed Oxide Fuel Construction Budget Facility (MOX) will significantly exceed

  3. RELAP5/MOD3.2 Analysis of a VVER-1000 Reactor with UO{sub 2} Fuel and Mixed-Oxide Fuel

    SciTech Connect (OSTI)

    Hassan, Yassin A.; Fu Chun

    2004-12-15

    A RELAP5/MOD3.2 model of the VVER-1000/MODEL V320 nuclear power plant was modified and a large-break loss-of-coolant accident (LBLOCA) in the cold leg was simulated. In this analysis, the core consisted of 162 UO{sub 2} assemblies and 1 mixed-oxide assembly. The results from the simulation were compared with the results from a similar study performed with the Russian computer program TECH-M. An uncertainty analysis was performed on the peak cladding temperature following a similar methodology called code scaling, applicability, and uncertainty. Monte Carlo calculations were performed using the response surface inferred from 15 runs of RELAP5 calculations. The result of this study shows that the emergency core coolant system would be sufficient to keep the cladding temperature during the LBLOCA scenario well below the required maximum limit.

  4. Interaction of iron-copper mixed metal oxide oxygen carriers with simulated synthesis gas derived from steam gasification of coal

    SciTech Connect (OSTI)

    Siriwardane, Ranjani V.; Ksepko, Ewelina; Tian, Hanging

    2013-01-01

    The objective of this work was to prepare supported bimetallic Fe–Cu oxygen carriers and to evaluate their performance for the chemical-looping combustion (CLC) process with simulated synthesis gas derived from steam gasification of coal/air. Ten-cycle CLC tests were conducted with Fe–Cu oxygen carriers in an atmospheric thermogravimetric analyzer utilizing simulated synthesis gas derived from the steam gasification of Polish Janina coal and Illinois #6 coal as fuel. The effect of temperature on reaction rates, chemical stability, and oxygen transport capacity were determined. Fractional reduction, fractional oxidation, and global rates of reactions were calculated from the thermogravimetric analysis (TGA) data. The supports greatly affected reaction performance. Data showed that reaction rates and oxygen capacities were stable during the 10-cycle TGA tests for most Fe–Cu/support oxygen carriers. Bimetallic Fe–Cu/support oxygen carriers showed higher reduction rates than Fe-support oxygen carriers. The carriers containing higher Cu content showed better stabilities and better reduction rates. An increase in temperature from 800 °C to 900 °C did not have a significant effect on either the oxygen capacity or the reduction rates with synthesis gas derived from Janina coal. Oxidation reaction was significantly faster than reduction reaction for all supported Fe–Cu oxygen carriers. Carriers with higher Cu content had lower oxidation rates. Ten-cycle TGA data indicated that these oxygen carriers had stable performances at 800–900 °C and might be successfully used up to 900 °C for coal CLC reaction in the presence of steam.

  5. A.12a (Pre-SAS 115) Letter to communicate significant deficiencies...

    Energy Savers [EERE]

    ... General Atomics Hot Cell Facility from the ... estimation of the surplus plutonium liability, as follows: * The Mixed Oxide (MOX) Fuel ... recorded receiptcost entries manually ...

  6. Analysis on fuel breeding capability of FBR core region based...

    Office of Scientific and Technical Information (OSTI)

    contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel ... to absorp neutrons for reducing excess reactivity and additional contribution for fuel ...

  7. Lessons Learned Quarterly Report; March 3, 2003; Issue No. 34...

    National Nuclear Security Administration (NNSA)

    Mixed Oxide (MOX) Fuel Fabrication Facility at DOE's Savannah River Site in South Carolina. ... terrorism in generic environmental reviews for nuclear power plant license renewal." ...

  8. Table of Contents

    Energy Savers [EERE]

    ... The facility will remove impurities from surplus weapons-grade plutonium and mix it with depleted uranium oxide to form fuel pellets for commercial nuclear power reactors. MOX ...

  9. Reverse micelles directed synthesis of TiO{sub 2}-CeO{sub 2} mixed oxides and investigation of their crystal structure and morphology

    SciTech Connect (OSTI)

    Matejova, Lenka; Vales, Vaclav; Fajgar, Radek; Matej, Zdenek; Holy, Vaclav; Solcova, Olga

    2013-02-15

    The synthesis of TiO{sub 2}-CeO{sub 2} mixed oxides based on the sol-gel process controlled within reverse micelles of non-ionic surfactant Triton X-114 in cyclohexane is reported. The crystallization, phase composition, trends in nanoparticles growth and porous structure properties are studied as a function of Ti:Ce molar composition and annealing temperature by in-situ X-ray diffraction, Raman spectroscopy and physisorption. The brannerite-type CeTi{sub 2}O{sub 6} crystallizes as a single crystalline phase at Ti:Ce molar composition of 70:30 and in the mixture with cubic CeO{sub 2} and anatase TiO{sub 2} for composition 50:50. At Ti:Ce molar ratios 90:10 and 30:70 the mixtures of TiO{sub 2} anatase, rutile and cubic CeO{sub 2} appear. In these mixtures TiO{sub 2} rutile is formed at higher temperatures than conventionally. Additionally, the amount of a present amorphous phase in individual mixtures was estimated from diffraction data. The porous structure morphology depends both on molar composition and annealing temperature. This is correlated with the presence of carbon impurities of different character. - Graphical abstract: The phase composition of Ti90--Ce10 and Ti50--Ce50 oxide mixtures as a function of annealing temperature. The amount of the amorphous phase was estimated and attributed to TiO{sub 2}. Highlights: Black-Right-Pointing-Pointer Ti/Ce oxides were prepared using reverse micelles of Triton X-114. Black-Right-Pointing-Pointer Crystallization of TiO{sub 2}, CeO{sub 2} or CeTi{sub 2}O{sub 6} depends on Ti:Ce molar ratio. Black-Right-Pointing-Pointer Amorphous phase attributed to TiO{sub 2} was identified. Black-Right-Pointing-Pointer Metal oxides surface area is influenced by the character of present carbon impurities.

  10. TRU decontamination of high-level Purex waste by solvent extraction using a mixed octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide/TBP/NPH (TRUEX) solvent

    SciTech Connect (OSTI)

    Horwitz, E.P.; Kalina, D.G.; Diamond, H.; Kaplan, L.; Vandegrift, G.F.; Leonard, R.A.; Steindler, M.J.; Schulz, W.W.

    1984-01-01

    The TRUEX (transuranium extraction) process was tested on a simulated high-level dissolved sludge waste (DSW). A batch counter-current extraction mode was used for seven extraction and three scrub stages. One additional extraction stage and two scrub stages and all strip stages were performed by batch extraction. The TRUEX solvent consisted of 0.20 M octyl(phenyl)-N,N-diisobutylcarbamoyl-methylphosphine oxide-1.4 M TBP in Conoco (C/sub 12/-C/sub 14/). The feed solution was 1.0 M in HNO/sub 3/, 0.3 M in H/sub 2/C/sub 2/O/sub 4/ and contained mixed (stable) fission products, U, Np, Pu, and Am, and a number of inert constituents, e.g., Fe and Al. The test showed that the process is capable of reducing the TRU concentration in the DSW by a factor of 4 x 10/sup 4/ (to <100 nCi/g of disposed form) and reducing the quantity of TRU waste by two orders of magnitude.

  11. Study on the mechanism of diametral cladding strain and mixed-oxide fuel element breaching in slow-ramp extended overpower transients

    SciTech Connect (OSTI)

    Tomoyuki Uwaba; Seiichiro Maeda; Tomoyasu Mizuno; Melissa C. Teague

    2012-10-01

    Cladding strain caused by fuel/cladding mechanical interaction (FCMI) was evaluated for mixed-oxide fuel elements subjected to 7090% slow-ramp extended overpower transient tests in the experimental breeder reactor II. Calculated transient-induced cladding strains were correlated with cumulative damage fractions (CDFs) using cladding strength correlations. In a breached high-smeared density solid fuel element with low strength cladding, cladding thermal creep strain was significantly increased to approximately half the transient-induced cladding strain that was considered to be caused by the tertiary creep when the CDF was close to the breach criterion (=1.0), with the remaining strain due to instantaneous plastic deformation. In low-smeared density annular fuel elements, FCMI load was significantly mitigated and resulted in little cladding strain. The CDFs of the annular fuel elements were lower than 0.01 at the end of the overpower transient, indicating a substantial margin to breach. A substantial margin to breach was also maintained in a high-smeared density fuel element with high strength cladding.

  12. A study of ZnxZryOz mixed oxides for direct conversion of ethanol to isobutene

    SciTech Connect (OSTI)

    Liu, Changjun; Sun, Junming; Smith, Colin; Wang, Yong

    2013-07-15

    ZnxZryOz mixed oxides were studied for direct conversion of ethanol to isobutene. Reaction conditions (temperature, residence time, ethanol molar fraction, steam to carbon ratio), catalyst composition, and pretreatment conditions were investigated, aiming at high-yield production of isobutene under industrially relevant conditions. An isobutene yield of 79% was achieved with an ethanol molar fraction of 8.3% at 475 °C on fresh Zn1Zr8O17 catalysts. Further durability and regeneration tests revealed that the catalyst exhibited very slow deactivation via coking formation with isobutene yield maintained above 75% for more than 10 h time-on-stream. More importantly, the catalysts activity in terms of isobutene yield can be readily recovered after in situ calcination in air at 550 °C for 2.5 h. XRD, TPO, IR analysis of adsorbed pyridine (IR-Py), and nitrogen sorption have been used to characterize the surface physical/chemical properties to correlate the structure and performance of the catalysts.

  13. 100% MOX BWR experimental program design using multi-parameter representative

    SciTech Connect (OSTI)

    Blaise, P.; Fougeras, P.; Cathalau, S.

    2012-07-01

    A new multiparameter representative approach for the design of Advanced full MOX BWR core physics experimental programs is developed. The approach is based on sensitivity analysis of integral parameters to nuclear data, and correlations among different integral parameters. The representativeness method is here used to extract a quantitative relationship between a particular integral response of an experimental mock-up and the same response in a reference project to be designed. The study is applied to the design of the 100% MOX BASALA ABWR experimental program in the EOLE facility. The adopted scheme proposes an original approach to the problem, going from the initial 'microscopic' pin-cells integral parameters to the whole 'macroscopic' assembly integral parameters. This approach enables to collect complementary information necessary to optimize the initial design and to meet target accuracy on the integral parameters to be measured. The study has demonstrated the necessity of new fuel pins fabrication, fulfilling minimal costs requirements, to meet acceptable representativeness on local power distribution. (authors)

  14. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 Volume 2-Calculations Performed in the United States

    SciTech Connect (OSTI)

    Primm III, RT

    2002-05-29

    This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the US during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the computational benchmarks and for those experimental benchmarks that the US and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors.

  15. Fischer-Tropsch Synthesis: Assessment of the Ripening of Cobalt Clusters and Mixing Between Co and Ru Promoter via Oxidation-Reduction-Cycles over Lower Co-Loaded Ru-Co/A12O3 Catalysts

    SciTech Connect (OSTI)

    Jacobs,G.; Sarkar, A.; Ji, Y.; Luo, M.; Dozier, A.; Davis, B.

    2008-01-01

    A 2% Ru-promoted 15% Co/Al2O3 catalyst was tested after reduction and after being subjected to oxidation-reduction cycles. The catalysts were characterized over four oxidation-reduction cycles by XANES/EXAFS, TPR, HRTEM, and EDS elemental mapping. The oxidation-reduction treatments were found to assist in sintering the metallic clusters to a larger size, and to promote mixing on at least the order of the nanoscale. The larger crystallites in closer proximity to the Ru promoter led to a more facile reduction of the cobalt crystallites. In addition, a catalyst exposed to two oxidation-reduction cycles resulted in slightly higher conversion, higher a-value product, slightly lower methane selectivity, and greater stability over a reduced freshly calcined catalyst.

  16. Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

    SciTech Connect (OSTI)

    Tomoyuki Uwaba; Masahiro Ito; Kozo Katsuyama; Bruce J. Makenas; David W. Wootan; Jon Carmack

    2011-05-01

    The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39 × 1026 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.

  17. Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

    SciTech Connect (OSTI)

    Uwaba, Tomoyuki; Ito, Masahiro; Mizuno, Tomoyasu; Katsuyama, Kozo; Makenas, Bruce J.; Wootan, David W.; Carmack, Jon

    2011-06-16

    The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39E26 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.

  18. Audit Report: IG-0713 | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    3 Audit Report: IG-0713 December 21, 2005 Status of the Mixed Oxide Fuel Fabrication Facility The audit disclosed that the cost of the Mixed Oxide Fuel Facility (MOX) will ...

  19. Analytical Results For MOX Colemanite Concrete Samples Received On November, 2013

    SciTech Connect (OSTI)

    Reigel, Marissa M.

    2013-12-18

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory (SRNL) is tasked with measuring the total density, partial hydrogen density, and partial boron density of the colemanite concrete. SRNL received two samples of colemanite concrete for analysis on November 21, 2013. The average total density of each of the samples measured by the ASTM method C 642, the average partial hydrogen density was measured using method ASTM E 1131, and the average partial boron density of each sample was measured according to ASTM C 1301. For all the samples tested, the total density and the boron partial density met or exceeded the specified limit. None of the samples met the lower limit for hydrogen partial density.

  20. ANALYTICAL RESULTS FOR MOX COLEMANITE CONCRETE SAMPLES RECEIVED ON SEPTEMBER 4, 2013

    SciTech Connect (OSTI)

    Reigel, M.

    2014-05-19

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory (SRNL) is tasked with measuring the total density, partial hydrogen density, and partial boron density of the colemanite concrete. SRNL received three samples of colemanite concrete for analysis on September 4, 2013. The average total density of each of the samples measured by the ASTM method C 642, the average partial hydrogen density was measured using method ASTM E 1131, and the average partial boron density of each sample was measured according to ASTM C 1301. The lower limits and measured values for the total density, hydrogen partial density, and boron partial density are presented. For all the samples tested, the total density and the boron partial density met or exceeded the specified limit. None of the samples met the lower limit for hydrogen partial density.

  1. ANALYTICAL RESULTS FOR MOX COLEMANITE CONCRETE SAMPLES RECEIVED ON NOVEMBER 21, 2013

    SciTech Connect (OSTI)

    Reigel, M.

    2014-05-19

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory (SRNL) is tasked with measuring the total density, partial hydrogen density, and partial boron density of the colemanite concrete. SRNL received two samples of colemanite concrete for analysis on November 21, 2013. The average total density of each of the samples measured by the ASTM method C 642, the average partial hydrogen density was measured using method ASTM E 1131, and the average partial boron density of each sample was measured according to ASTM C 1301. The lower limits and measured values for the total density, hydrogen partial density, and boron partial density are presented. For all the samples tested, the total density and the boron partial density met or exceeded the specified limit. None of the samples met the lower limit for hydrogen partial density.

  2. Analytical Results For MOX Colemanite Concrete Samples Received On September 4, 2013

    SciTech Connect (OSTI)

    Reigel, Marissa M.

    2013-09-24

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory (SRNL) is tasked with measuring the total density, partial hydrogen density, and partial boron density of the colemanite concrete. SRNL received three samples of colemanite concrete for analysis on September 4, 2013. The average total density of each of the samples measured by the ASTM method C 642, the average partial hydrogen density was measured using method ASTM E 1131, and the average partial boron density of each sample was measured according to ASTM C 1301. The lower limits and measured values for the total density, hydrogen partial density, and boron partial density are presented. For all the samples tested, the total density and the boron partial density met or exceeded the specified limit. None of the samples met the lower limit for hydrogen partial density.

  3. THERMAL EVALUATION OF THE USE OF BWR MOX SNF IN THE WASTE PACKAGE DESIGN (SCPB: N/A)

    SciTech Connect (OSTI)

    H. Wang

    1997-01-23

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8,4-11,4-24, 5-1, and 5-13; Ref. 5.10) and Waste Package Plan (pp. 3-15,3-17, and 3-24; Ref. 5.9). The design data request addressed herein is: (1) Characterize the conceptual 40 BWR and 24 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. (2) Characterize the conceptual 44 BWR and 24 BWR Uncanistered Fuel (UCF) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the WP Design, if used with SNF designed for a MOX fuel cycle, do not preclude WP compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual WP design with disposal container which is loaded with BWR MOX SNF under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation, and to provide the required guidance to determining the major design issues for future design efforts, and to show that the BWR MOX SNF loaded WP performance is similar to an WP loaded with commercial BWR SNF.

  4. Audit Report: IG-0887 | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    7 Audit Report: IG-0887 May 15, 2013 The Use of Staff Augmentation Subcontracts at the National Nuclear Security Administration's Mixed Oxide Fuel Fabrication Facility Shaw AREVA MOX Services, LLC (MOX Services) is responsible for the design and construction of the National Nuclear Security Administration's (NNSA) nearly $5 billion Mixed Oxide Fuel Fabrication Facility (MOX Project) at the Savannah River Site near Aiken, South Carolina. The facility will remove impurities from surplus

  5. IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

    SciTech Connect (OSTI)

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

    This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis, the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  6. ANALYTICAL RESULTS FOR MOX COLEMANITE CONCRETE SAMPLES RECEIVED ON JANUARY 15, 2013

    SciTech Connect (OSTI)

    Reigel, M.; Best, D.

    2013-02-13

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory (SRNL) is tasked with measuring the total density, partial hydrogen density, and partial boron density of the colemanite concrete. SRNL received twelve samples of colemanite concrete for analysis on January 15, 2013. The average total density of each of the samples measured by the ASTM method C 642, the average partial hydrogen density was measured using method ASTM E 1311, and the average partial boron density of each sample was measured according to ASTM C 1301. The lower limits and measured values for the total density, hydrogen partial density, and boron partial density are presented. For all the samples tested, the total density and the hydrogen partial density met or exceeded the specified limit. All of the samples met or exceeded the boron partial density lower bound with the exception of samples G3-M11-2000-H, G3-M11-3000-M, and G5-M1-3000-H which are below the limit of 1.65E-01 g/cm3.

  7. ANALYTICAL RESULTS OF MOX COLEMANITE CONCRETE SAMPLE PBC-44.2

    SciTech Connect (OSTI)

    Best, D.; Cozzi, A.; Reigel, M.

    2012-12-20

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory is tasked with measuring the total density, partial hydrogen density, and partial boron density of the colemanite concrete. Sample PBC-44.2 was received on 9/20/2012 and analyzed. The average total density measured by the ASTM method C 642 was 2.03 g/cm{sup 3}, within the lower bound of 1.88 g/cm3. The average partial hydrogen density was 6.64E-02 g/cm{sup 3} as measured using method ASTM E 1311 and met the lower bound of 6.04E-02 g/cm{sup 3}. The average measured partial boron density was 1.70E-01 g/cm{sup 3} which met the lower bound of 1.65E-01 g/cm{sup 3} measured by the ASTM C 1301 method.

  8. ANALYTICAL RESULTS OF MOX COLEMANITE CONCRETE SAMPLES POURED AUGUST 29, 2012

    SciTech Connect (OSTI)

    Cozzi, A.; Best, D.; Reigel, M.

    2012-10-30

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory is tasked with measuring the total density, partial hydrogen density, and partial boron density of the colemanite concrete. Samples poured 8/29/12 were received on 9/20/2012 and analyzed. The average total density of each of the samples measured by the ASTM method C 642 was within the lower bound of 1.88 g/cm{sup 3}. The average partial hydrogen density of samples 8.6.1, 8.7.1, and 8.5.3 as measured using method ASTM E 1311 met the lower bound of 6.04E-02 g/cm{sup 3}. The average measured partial boron density of each sample met the lower bound of 1.65E-01 g/cm{sup 3} measured by the ASTM C 1301 method. The average partial hydrogen density of samples 8.5.1, 8.6.3, and 8.7.3 did not meet the lower bound. The samples, as received, were not wrapped in a moist towel as previous samples and appeared to be somewhat drier. This may explain the lower hydrogen partial density with respect to previous samples.

  9. ANALYTICAL RESULTS OF MOX COLEMANITE CONCRETE SAMPLES POURED AUGUST 29, 2012

    SciTech Connect (OSTI)

    Best, D.; Cozzi, A.; Reigel, M.

    2012-12-20

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory is tasked with measuring the total density, partial hydrogen density, and partial boron density of the colemanite concrete. Samples poured 8/29/12 were received on 9/20/2012 and analyzed. The average total density of each of the samples measured by the ASTM method C 642 was within the lower bound of 1.88 g/cm{sup 3}. The average partial hydrogen density of samples 8.6.1, 8.7.1, and 8.5.3 as measured using method ASTM E 1311 met the lower bound of 6.04E-02 g/cm{sup 3}. The average measured partial boron density of each sample met the lower bound of 1.65E-01 g/cm{sup 3} measured by the ASTM C 1301 method. The average partial hydrogen density of samples 8.5.1, 8.6.3, and 8.7.3 did not meet the lower bound. The samples, as received, were not wrapped in a moist towel as previous samples and appeared to be somewhat drier. This may explain the lower hydrogen partial density with respect to previous samples.

  10. ANALYTICAL RESULTS OF MOX COLEMANITE CONCRETE SAMPLE POURED JULY 25, 2012 - CURED 28 DAYS

    SciTech Connect (OSTI)

    Cozzi, A. D.; Best, D. R.; Reigel, M. M.

    2012-09-18

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use Colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory is tasked with measuring the total density, partial hydrogen density, and partial boron density of the colemanite concrete. Samples 8.1.2, 8.2.2, 8.3.2, and 8.4.2 were received on 8/1/2012 and analyzed after curing for 28 days. The average total density measured by the ASTM method C 642 was 2.09 g/cm{sup 3}, within the lower bound of 1.88 g/cm{sup 3}. The average partial hydrogen density was 7.48E-02 g/cm{sup 3} as measured using method ASTM E 1311 and met the lower bound of 6.04E-02 g/cm{sup 3}. The average measured partial boron density was 1.71E-01 g/cm{sup 3} which met the lower bound of 1.65E-01 g/cm{sup 3} measured by the ASTM C 1301 method.

  11. ANALYTICAL RESULTS FOR MOX COLEMANITE CONCRETE SAMPLES RECEIVED ON JANUARY 15, 2013

    SciTech Connect (OSTI)

    Reigel, M.

    2014-05-19

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory (SRNL) is tasked with measuring the total density, partial hydrogen density, and partial boron density of the colemanite concrete. SRNL received twelve samples of colemanite concrete for analysis on January 15, 2013. The average total density of each of the samples measured by the ASTM method C 642, the average partial hydrogen density was measured using method ASTM E 1131, and the average partial boron density of each sample was measured according to ASTM C 1301. The lower limits and measured values for the total density, hydrogen partial density, and boron partial density are presented. For all the samples tested, the total density and the hydrogen partial density met or exceeded the specified limit. All of the samples met or exceeded the boron partial density lower bound with the exception of samples G3-M11-2000-H, G3-M11-3000-M, and G5-M1-3000-H which are below the limit of 1.65E-01 g/cm{sup 3}.

  12. Analytical Results Of MOX Colemanite Concrete Sample PBC-44.2

    SciTech Connect (OSTI)

    Cozzi, A. D.; Best, D. R.; Reigel, M. M.

    2012-10-18

    The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory is tasked with measuring the total density, partial hydrogen density, and partial boron density of the colemanite concrete. Sample PBC-44.2 was received on 9/20/2012 and analyzed. The average total density measured by the ASTM method C 642 was 2.03 g/cm{sup 3}, within the lower bound of 1.88 g/cm{sup 3}. The average partial hydrogen density was 6.64E-02 g/cm{sup 3} as measured using method ASTM E 1311 and met the lower bound of 6.04E-02 g/cm{sup 3}. The average measured partial boron density was 1.97E-01 g/cm{sup 3} which met the lower bound of 1.65E-01 g/cm{sup 3} measured by the ASTM C 1301 method.

  13. Impact of Pt loading methods over mesoporous-assembled TiO{sub 2}ZrO{sub 2} mixed oxide nanocrystal on photocatalytic dye-sensitized H{sub 2} production activity

    SciTech Connect (OSTI)

    Sreethawong, Thammanoon; Yoshikawa, Susumu

    2012-06-15

    Graphical abstract: The Pt loading on the synthesized mesoporous-assembled 0.95TiO{sub 2}0.05ZrO{sub 2} mixed oxide nanocrystal photocatalyst was comparatively investigated by two methods: single-step solgel (SSSG) and photochemical deposition (PCD). The Pt loading by the PCD method was found to be superior to that by the SSSG method in enhancing photocatalytic sensitized hydrogen production under visible light irradiation. The Pt loading amount and PCD conditions, i.e. light irradiation time and light intensity, also had a strong effect on the photocatalytic hydrogen production activity. Highlights: ? Pt-loaded mesoporous-assembled 0.95TiO{sub 2}0.05ZrO{sub 2} nanocrystals were synthesized. ? Pt loading was performed by single-step solgel and photochemical deposition. ? Pt loading by photochemical deposition more enhanced photocatalytic H{sub 2} production. ? Pt loading amount and photochemical deposition conditions were optimized. -- Abstract: In this work, the photocatalytic water splitting under visible light irradiation for hydrogen production was investigated by using Eosin Y-sensitized Pt-loaded mesoporous-assembled TiO{sub 2}ZrO{sub 2} mixed oxide nanocrystal photocatalysts. The mesoporous-assembled TiO{sub 2}ZrO{sub 2} mixed oxide with the TiO{sub 2}-to-ZrO{sub 2} molar ratio of 95:5 (i.e. 0.95TiO{sub 2}0.05ZrO{sub 2}) was synthesized by using a solgel process with the aid of a structure-directing surfactant. The Pt loading was comparatively performed via two different effective methods: single-step solgel (SSSG) and photochemical deposition (PCD). The synthesized photocatalysts were methodically characterized by N{sub 2} adsorptiondesorption, XRD, UVvisible spectroscopy, SEMEDX, TEMEDX, TPR, and H{sub 2} chemisorption analyses. The results revealed that the Pt loading by the PCD method greatly enhanced the photocatalytic hydrogen production activity of the synthesized mesoporous-assembled 0.95TiO{sub 2}0.05ZrO{sub 2} mixed oxide

  14. Release and disposal of materials during decommissioning of Siemens MOX fuel fabrication plant at Hanau, Germany

    SciTech Connect (OSTI)

    Koenig, Werner; Baumann, Roland

    2007-07-01

    In September 2006, decommissioning and dismantling of the Siemens MOX Fuel Fabrication Plant in Hanau were completed. The process equipment and the fabrication buildings were completely decommissioned and dismantled. The other buildings were emptied in whole or in part, although they were not demolished. Overall, the decommissioning process produced approximately 8500 Mg of radioactive waste (including inactive matrix material); clearance measurements were also performed for approximately 5400 Mg of material covering a wide range of types. All the equipment in which nuclear fuels had been handled was disposed of as radioactive waste. The radioactive waste was conditioned on the basis of the requirements specified for the projected German final disposal site 'Schachtanlage Konrad'. During the pre-conditioning, familiar processes such as incineration, compacting and melting were used. It has been shown that on account of consistently applied activity containment (barrier concept) during operation and dismantling, there has been no significant unexpected contamination of the plant. Therefore almost all the materials that were not a priori destined for radioactive waste were released without restriction on the basis of the applicable legal regulations (chap. 29 of the Radiation Protection Ordinance), along with the buildings and the plant site. (authors)

  15. Time cycle analysis and simulation of material flow in MOX process layout

    SciTech Connect (OSTI)

    Chakraborty, S.; Saraswat, A.; Danny, K.M.; Somayajulu, P.S.; Kumar, A.

    2013-07-01

    The (U,Pu)O{sub 2} MOX fuel is the driver fuel for the upcoming PFBR (Prototype Fast Breeder Reactor). The fuel has around 30% PuO{sub 2}. The presence of high percentages of reprocessed PuO{sub 2} necessitates the design of optimized fuel fabrication process line which will address both production need as well as meet regulatory norms regarding radiological safety criteria. The powder pellet route has highly unbalanced time cycle. This difficulty can be overcome by optimizing process layout in terms of equipment redundancy and scheduling of input powder batches. Different schemes are tested before implementing in the process line with the help of a software. This software simulates the material movement through the optimized process layout. The different material processing schemes have been devised and validity of the schemes are tested with the software. Schemes in which production batches are meeting at any glove box location are considered invalid. A valid scheme ensures adequate spacing between the production batches and at the same time it meets the production target. This software can be further improved by accurately calculating material movement time through glove box train. One important factor is considering material handling time with automation systems in place.

  16. Doped palladium containing oxidation catalysts

    DOE Patents [OSTI]

    Mohajeri, Nahid

    2014-02-18

    A supported oxidation catalyst includes a support having a metal oxide or metal salt, and mixed metal particles thereon. The mixed metal particles include first particles including a palladium compound, and second particles including a precious metal group (PMG) metal or PMG metal compound, wherein the PMG metal is not palladium. The oxidation catalyst may also be used as a gas sensor.

  17. Computation Results from a Parametric Study to Determine Bounding Critical Systems of Homogeneously Water-Moderated Mixed Plutonium--Uranium Oxides

    SciTech Connect (OSTI)

    Shimizu, Y.

    2001-01-11

    This report provides computational results of an extensive study to examine the following: (1) infinite media neutron-multiplication factors; (2) material bucklings; (3) bounding infinite media critical concentrations; (4) bounding finite critical dimensions of water-reflected and homogeneously water-moderated one-dimensional systems (i.e., spheres, cylinders of infinite length, and slabs that are infinite in two dimensions) that were comprised of various proportions and densities of plutonium oxides and uranium oxides, each having various isotopic compositions; and (5) sensitivity coefficients of delta k-eff with respect to critical geometry delta dimensions were determined for each of the three geometries that were studied. The study was undertaken to support the development of a standard that is sponsored by the International Standards Organization (ISO) under Technical Committee 85, Nuclear Energy (TC 85)--Subcommittee 5, Nuclear Fuel Technology (SC 5)--Working Group 8, Standardization of Calculations, Procedures and Practices Related to Criticality Safety (WG 8). The designation and title of the ISO TC 85/SC 5/WG 8 standard working draft is WD 14941, ''Nuclear energy--Fissile materials--Nuclear criticality control and safety of plutonium-uranium oxide fuel mixtures outside of reactors.'' Various ISO member participants performed similar computational studies using their indigenous computational codes to provide comparative results for analysis in the development of the standard.

  18. Roles of Fe2+, Fe3+, and Cr3+ Surface Sites in the Oxidation of NO on the (Fe,Cr)3O4(1 1 1) Surface Termination of an ?-(Fe,Cr)2O3(0 0 0 1) Mixed Oxide

    SciTech Connect (OSTI)

    Henderson, Michael A.

    2014-10-01

    The oxidation and photooxidation reactions of nitric oxide were explored on a mixed Fe and Cr mixed oxide surface using temperature programmed desorption (TPD). The mixed oxide surface examined initially had a corundum (0001) structure with a nominal cation composition of 75% Fe and 25% Cr, but after sputter/anneal cleaning was transformed into a magnetite-like (111) surface structure enriched with Cr (~40%). TPD studies of nitric oxide on the (Fe,Cr)3O4(111) surface revealed two main desorption states at 220 and 370 K, along with a third minor desorption state at ~310 K. Similarly, O2 TPD occurred in two main TPD states (100 and 230 K) and a minor state (155 K). The more strongly and weakly bound NO and O2 molecules were assigned to adsorption at Fe2+ and Fe3+ sites, respectively, with the minor desorption states assigned to Cr3+ sites. No thermal decomposition or surface chemistry was detected in TPD for adsorbed NO (e.g., no N2 or N2O formation), whereas ~10% of the adsorbed O2 irreversibly dissociated at Fe2+ sites. These dissociated oxygen species did not react with coadsorbed NO, but instead blocked NO adsorption at the Fe2+ sites, but had no effect on NO adsorption at Fe3+ sites. In contrast, NO reacted with preadsorbed O2 molecules to generate an adsorbed nitrate/nitrite species that decomposed in TPD to liberate NO at 425 K, leaving an O atom on the surface. Coadsorption of 15N18O with 16O2 suggests the oxidized species was a nitrate based on the detected level of oxygen scrambling. Preadsorption of O2 was required for nitrate formation as preadsorbed NO blocked both O2 adsorption and the oxidation reaction. Irradiation of adsorbed NO with 460 nm light at 40 K resulted in rapid photodesorption of NO without generation of any new surface species. Irradiation of the coadsorbed NO+O2 system did not promote additional NO oxidation, but limited the extent of thermal NO oxidation (in subsequent TPD) by photodepleting the surface of adsorbed NO. Preheating the NO

  19. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    SciTech Connect (OSTI)

    Salay, Michael; Gauntt, Randall O.; Lee, Richard Y.; Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  20. Prediction of an arc-tunable Weyl Fermion metallic state in MoxW1-xTe2

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Chang, Tay-Rong; Xu, Su-Yang; Chang, Guoqing; Lee, Chi-Cheng; Huang, Shin-Ming; Wang, BaoKai; Bian, Guang; Zheng, Hao; Sanchez, Daniel S.; Belopolski, Ilya; et al

    2016-02-15

    A Weyl semimetal is a new state of matter that hosts Weyl fermions as emergent quasiparticles. The Weyl fermions correspond to isolated points of bulk band degeneracy, Weyl nodes, which are connected only through the crystal’s boundary by exotic Fermi arcs. The length of the Fermi arc gives a measure of the topological strength, because the only way to destroy the Weyl nodes is to annihilate them in pairs in the reciprocal space. To date, Weyl semimetals are only realized in the TaAs class. Here, we propose a tunable Weyl state in MoxW1₋xTe2 where Weyl nodes are formed by touchingmore » points between metallic pockets. We show that the Fermi arc length can be changed as a function of Mo concentration, thus tuning the topological strength. Lastly,our results provide an experimentally feasible route to realizing Weyl physics in the layered compound MoxW1₋xTe2, where non-saturating magneto-resistance and pressure-driven superconductivity have been observed.« less

  1. Superior performance of Ni-W-Ce mixed-metal oxide catalysts for ethanol steam reforming: Synergistic effects of W- and Ni-dopants

    SciTech Connect (OSTI)

    Rodriguez, Jose A.; Liu, Zongyuan; Xu, Wenqian; Yao, Siyu; Johnson-Peck, Aaron C.; Zhao, Fuzhen; Michorczyk, Piotr; Kubacka, Anna; Stach, Eric A.; Fernandez-Garica, Marcos; Senanayake, Sanjaya D.

    2014-11-26

    The ethanol steam reforming (ESR) reaction was studied over a series of Ni-W-Ce oxide catalysts. The structures of the catalysts were characterized using in-situ techniques including X-ray diffraction, Pair Distribution Function, X-ray absorption fine structure and transmission electron microscopy; while possible surface intermediates for the ESR reaction were investigated by Diffuse Reflectance Infrared Fourier Transform Spectroscopy. In these materials, all the W and part of the Ni were incorporated into the CeO? lattice, with the remaining Ni forming highly dispersed nano NiO (< 2 nm) outside the Ni-W-Ce oxide structure. The nano NiO was reduced to Ni under ESR conditions. The Ni-W-Ce systeme exhibited a much larger lattice strain than those seen for Ni-Ce and W-Ce. Synergistic effects between Ni and W inside ceria produced a substantial amount of defects and O vacancies that led to high catalytic activity, selectivity and stability (i.e. resistance to coke formation) during ethanol steam reforming.

  2. Superior performance of Ni–W–Ce mixed-metal oxide catalysts for ethanol steam reforming: Synergistic effects of W- and Ni-dopants

    SciTech Connect (OSTI)

    Liu, Zongyuan; Xu, Wenqian; Yao, Siyu; Johnson-Peck, Aaron C.; Zhao, Fuzhen; Michorczyk, Piotr; Kubacka, Anna; Stach, Eric A.; Fernández-García, Marcos; Senanayake, Sanjaya D.; Rodriguez, José A.

    2014-11-26

    In this study, the ethanol steam reforming (ESR) reaction was examined over a series of Ni-W-Ce oxide catalysts. The structures of the catalysts were characterized using in-situ techniques including X-ray diffraction, Pair Distribution Function, X-ray absorption fine structure and transmission electron microscopy; while possible surface intermediates for the ESR reaction were investigated by Diffuse Reflectance Infrared Fourier Transform Spectroscopy. In these materials, all the W and part of the Ni were incorporated into the CeO₂ lattice, with the remaining Ni forming highly dispersed nano NiO (< 2 nm) outside the Ni-W-Ce oxide structure. The nano NiO was reduced to Ni under ESR conditions. The Ni-W-Ce systeme exhibited a much larger lattice strain than those seen for Ni-Ce and W-Ce. Synergistic effects between Ni and W inside ceria produced a substantial amount of defects and O vacancies that led to high catalytic activity, selectivity and stability (i.e. resistance to coke formation) during ethanol steam reforming.

  3. Superior performance of Ni–W–Ce mixed-metal oxide catalysts for ethanol steam reforming: Synergistic effects of W- and Ni-dopants

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Liu, Zongyuan; Xu, Wenqian; Yao, Siyu; Johnson-Peck, Aaron C.; Zhao, Fuzhen; Michorczyk, Piotr; Kubacka, Anna; Stach, Eric A.; Fernández-García, Marcos; Senanayake, Sanjaya D.; et al

    2014-11-26

    In this study, the ethanol steam reforming (ESR) reaction was examined over a series of Ni-W-Ce oxide catalysts. The structures of the catalysts were characterized using in-situ techniques including X-ray diffraction, Pair Distribution Function, X-ray absorption fine structure and transmission electron microscopy; while possible surface intermediates for the ESR reaction were investigated by Diffuse Reflectance Infrared Fourier Transform Spectroscopy. In these materials, all the W and part of the Ni were incorporated into the CeO₂ lattice, with the remaining Ni forming highly dispersed nano NiO (< 2 nm) outside the Ni-W-Ce oxide structure. The nano NiO was reduced to Nimore » under ESR conditions. The Ni-W-Ce systeme exhibited a much larger lattice strain than those seen for Ni-Ce and W-Ce. Synergistic effects between Ni and W inside ceria produced a substantial amount of defects and O vacancies that led to high catalytic activity, selectivity and stability (i.e. resistance to coke formation) during ethanol steam reforming.« less

  4. http://srsweb2.srs.gov/InSiteApp/assets/xml/articles/574.html

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    MOX Conservation Garden Features Federally Endangered Plant Linda Lee, botanist for the Savannah River Ecology Lab (from left), Clay Ramsey , federal project director of the Mixed Oxide Fuel Fabrication Facility for the National Nuclear Security Administration, and Kelly Trice, president and chief operating officer of Shaw AREVA MOX Services, place the final stones in the new MOX Conservation Garden. A conservation garden that features a rare, endangered plant native to SRS was dedicated April

  5. Superior performance of Ni-W-Ce mixed-metal oxide catalysts for ethanol steam reforming: Synergistic effects of W- and Ni-dopants

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Rodriguez, Jose A.; Liu, Zongyuan; Xu, Wenqian; Yao, Siyu; Johnson-Peck, Aaron C.; Zhao, Fuzhen; Michorczyk, Piotr; Kubacka, Anna; Stach, Eric A.; Fernandez-Garica, Marcos; et al

    2014-11-26

    The ethanol steam reforming (ESR) reaction was studied over a series of Ni-W-Ce oxide catalysts. The structures of the catalysts were characterized using in-situ techniques including X-ray diffraction, Pair Distribution Function, X-ray absorption fine structure and transmission electron microscopy; while possible surface intermediates for the ESR reaction were investigated by Diffuse Reflectance Infrared Fourier Transform Spectroscopy. In these materials, all the W and part of the Ni were incorporated into the CeO? lattice, with the remaining Ni forming highly dispersed nano NiO (moreThe Ni-W-Ce systeme exhibited a much larger lattice strain than those seen for Ni-Ce and W-Ce. Synergistic effects between Ni and W inside ceria produced a substantial amount of defects and O vacancies that led to high catalytic activity, selectivity and stability (i.e. resistance to coke formation) during ethanol steam reforming.less

  6. Application of modified direct denitration to support the ORNL coupled-end-to-end demonstration in production of mixed oxides suitable for pellet fabrication

    SciTech Connect (OSTI)

    Walker, E.A.; Vedder, R.J.; Felker, L.K.; Marschman, S.C.

    2007-07-01

    The current and future development of the Modified Direct Denitration (MDD) process is in support of Oak Ridge National Laboratory's (ORNL) Coupled End-to-End (CETE) research, development, and demonstration (R and D) of proposed advanced fuel reprocessing and fuel fabrication processes. This work will involve the co-conversion of the U/Pu/Np product streams from the UREX+3 separation flow sheet utilizing the existing MDD glove-box setup and the in-cell co-conversion of the U/Pu/Np/Am/Cm product streams from the UREX+1a flow sheet. Characterization equipment is being procured and installed. Oxide powder studies are being done on calcination/reduction variables, as well as pressing and sintering of pellets to permit metallographic examinations. (authors)

  7. Fuel Mix and Emissions Disclosure | Department of Energy

    Broader source: Energy.gov (indexed) [DOE]

    customers the fuel mix of its electricity production and the associated sulfur dioxide, nitrogen oxide, and carbon dioxide emissions emissions, expressed in pounds per 1000...

  8. Material unaccounted for at the Southwest Experimental Fast Oxide Reactor: The SEFOR MUF

    SciTech Connect (OSTI)

    Higinbotham, W.A.

    1994-11-07

    The U.S. Atomic Energy Commission contracted with the General Electric Company to design, construct, and operate the Southwest Experimental Fast Oxide Reactor (SEFOR) to measure the Doppler effect for fast neutron breeder reactors. It contracted with Nuclear Fuel Services to fabricate the fuel rods for the reactor. When the reactor went critical in May, 1969, it appeared that some of the mixed uranium-plutonium oxide (MOX) fuel rods did not contain the specified quantity of plutonium. The SEFOR operators soon found several fuel rods which appeared to be low in plutonium. The safeguards group at Brookhaven was asked to look into the problem and, if possible, determine how much plutonium was missing from the unirradiated rods and from the larger number which had been slightly irradiated in the reactor. It was decided that the plutonium content of the unirradiated and irradiated rods could be measured relative to a reference rod using a high resolution gamma-ray detector and also by neutron measurements using an auto-correlation circuit recently developed at the Naval Research Laboratory (NRL). During the next two years, Brookhaven personnel and C.V. Strain of NRL made several trips to the SEFOR reactor. About 250 of the 775 rods were measured by two or more methods, using a sodium-iodide detector, a high-resolution germanium detector, a neutron detector, or the reactor (to measure reactivity). The research team concluded that 4.6 {+-} 0.46 kg of plutonium was missing out of the 433 kg that the rods should have contained. This report describes the SEFOR experiment and the procedures used to determine the material unaccounted for, or MUF.

  9. ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

    SciTech Connect (OSTI)

    Lell, R. M.; Schaefer, R. W.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; Nuclear Engineering Division; Inst. of Physics and Power Engineering

    2007-10-01

    Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of demonstration

  10. Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-MOX, R2-UO2 and MORGANE/R configurations.

    SciTech Connect (OSTI)

    Zhong, Z.; Klann, R. T.; Nuclear Engineering Division

    2007-08-03

    An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R2-UO2 and MORGANE/R core configuration were completed. The calculation model was generated using the lattice physics code DRAGON. In addition, an initial comparison of calculated values to experimental measurements was performed based on preliminary results for the R1-MOX configuration.

  11. Overall Plan for Physics Outlining Steps Necessary for Insertion of the LTA and Operation Using a 1/3 MOX Loaded Core

    SciTech Connect (OSTI)

    Pavlovichev, A.M.

    2001-04-09

    Document issued according to Work Release KI-WR04RTP. P. 00-1 describes physics tasks that are included in the current version of ''Roadmap.Level 2'' concerning Reactor tasks of Weapon-grade plutonium disposition problem for VVER-1000. On this base the objective is to identify the physical tasks in FY2000 and in future as a part of global activities on weapon-grade MOX fuel introduction into VVER-1000.

  12. Changes of O/M, dimension and microstructure of MOX pellet during heat treatment

    SciTech Connect (OSTI)

    Watanabe, M.; Kato, M.; Sunaoshi, T.

    2013-07-01

    The oxidation and reduction behaviors of sintered (Pu{sub 0.3}U{sub 0.7})O{sub 2-x} pellets have been studied at 1873 K under a controlled oxygen partial pressure. From the results of oxygen-to-metal (O/M) ratio changes, dimensional and structural changes, it was concluded that the crack nucleation-propagation and the local density change of pores were caused by the tensile and compressive stresses due to the O/M ratio distribution in the direction of the pellet radius. (authors)

  13. A global approach of the representativity concept: Application on a high-conversion light water reactor MOX lattice case

    SciTech Connect (OSTI)

    Santos, N. D.; Blaise, P.; Santamarina, A.

    2013-07-01

    The development of new types of reactor and the increase in the safety specifications and requirements induce an enhancement in both nuclear data knowledge and a better understanding of the neutronic properties of the new systems. This enhancement is made possible using ad hoc critical mock-up experiments. The main difficulty is to design these experiments in order to obtain the most valuable information. Its quantification is usually made by using representativity and transposition concepts. These theories enable to extract some information about a quantity of interest (an integral parameter) on a configuration, but generally a posteriori. This paper presents a more global approach of this theory, with the idea of optimizing the representativity of a new experiment, and its transposition a priori, based on a multiparametric approach. Using a quadratic sum, we show the possibility to define a global representativity which permits to take into account several quantities of interest at the same time. The maximization of this factor gives information about all quantities of interest. An optimization method of this value in relation to technological parameters (over-clad diameter, atom concentration) is illustrated on a high-conversion light water reactor MOX lattice case. This example tackles the problematic of plutonium experiment for the plutonium aging and a solution through the optimization of both the over-clad and the plutonium content. (authors)

  14. Literature review for oxalate oxidation processes and plutonium oxalate solubility

    SciTech Connect (OSTI)

    Nash, C. A.

    2015-10-01

    A literature review of oxalate oxidation processes finds that manganese(II)-catalyzed nitric acid oxidation of oxalate in precipitate filtrate is a viable and well-documented process. The process has been operated on the large scale at Savannah River in the past, including oxidation of 20 tons of oxalic acid in F-Canyon. Research data under a variety of conditions show the process to be robust. This process is recommended for oxalate destruction in H-Canyon in the upcoming program to produce feed for the MOX facility. Prevention of plutonium oxalate precipitation in filtrate can be achieved by concentrated nitric acid/ferric nitrate sequestration of oxalate. Organic complexants do not appear practical to sequester plutonium. Testing is proposed to confirm the literature and calculation findings of this review at projected operating conditions for the upcoming campaign.

  15. Investigation of Mixed Oxide Catalysts for NO Oxidation

    Broader source: Energy.gov [DOE]

    2013 DOE Hydrogen and Fuel Cells Program and Vehicle Technologies Program Annual Merit Review and Peer Evaluation Meeting

  16. Investigation of Mixed Oxide Catalysts for NO Oxidation | Department of

    Broader source: Energy.gov (indexed) [DOE]

    (Presentation) | Department of Energy Presented at the 2007 Bio-Derived Liquids to Hydrogen Distributed Reforming Working Group held November 6, 2007 in Laurel, Maryland. 08_osu_bio-ethanol_steam_reforming.pdf (6.45 MB) More Documents & Publications Investigation of Reaction Networks and Active Sites In Bio-Ethanol Steam Reforming Over Co-Based Catalysts Bio-Derived Liquids to Hydrogen Distributed Reforming Working Group (BILIWG), Hydrogen Separation and Purification Working Group

  17. AP-XPS Measures MIEC Oxides in Action

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AP-XPS Measures MIEC Oxides in Action Print Oxide materials with mixed ionic-electronic conductivity (MIEC) can conduct both electrons and oxygen ions. MIEC oxides have broad...

  18. ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

    SciTech Connect (OSTI)

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

    2009-02-23

    ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide, U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium

  19. Sylgard Mixing Study

    SciTech Connect (OSTI)

    Bello, Mollie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Welch, Cynthia F. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Goodwin, Lynne Alese [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Keller, Jennie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-08-22

    Sylgard 184 and Sylgard 186 silicone elastomers form Dow Corning are used as potting agents across the Nuclear Weapons Complex. A standardized mixing procedure is required for filled versions of these products. The present study is a follow-up to a mixing study performed by MST-7 which established the best mixing procedure to use when adding filler to either 184 or 186 base resins. The most effective and consistent method of mixing resin and curing agent for three modified silicone elastomer recipes is outlined in this report. For each recipe, sample size, mixing type, and mixing time was varied over 10 separate runs. The results show that the THINKY Mixer gives reliable mixing over varying batch sizes and mixing times. Hand Mixing can give improved mixing, as indicated by reduced initial viscosity; however, this method is not consistent.

  20. Heterogeneous Reburning By Mixed Fuels

    SciTech Connect (OSTI)

    Anderson Hall

    2009-03-31

    Recent studies of heterogeneous reburning, i.e., reburning involving a coal-derived char, have elucidated its variables, kinetics and mechanisms that are valuable to the development of a highly efficient reburning process. Young lignite chars contain catalysts that not only reduce NO, but they also reduce HCN that is an important intermediate that recycles to NO in the burnout zone. Gaseous CO scavenges the surface oxides that are formed during NO reduction, regenerating the active sites on the char surface. Based on this mechanistic information, cost-effective mixed fuels containing these multiple features has been designed and tested in a simulated reburning apparatus. Remarkably high reduction of NO and HCN has been observed and it is anticipated that mixed fuel will remove 85% of NO in a three-stage reburning process.

  1. Mixing in astrophysics

    SciTech Connect (OSTI)

    Fryer, Christopher Lee

    2011-01-07

    Turbulent mixing plays a vital role in many fields in astronomy. Here I review a few of these sites, discuss the importance of this turbulent mixing and the techniques used by astrophysicists to solve these problems.

  2. Metal oxide films on metal

    DOE Patents [OSTI]

    Wu, Xin D. (Los Alamos, NM); Tiwari, Prabhat (Los Alamos, NM)

    1995-01-01

    A structure including a thin film of a conductive alkaline earth metal oxide selected from the group consisting of strontium ruthenium trioxide, calcium ruthenium trioxide, barium ruthenium trioxide, lanthanum-strontium cobalt oxide or mixed alkaline earth ruthenium trioxides thereof upon a thin film of a noble metal such as platinum is provided.

  3. Mixed Solvent Electrolyte Model

    Broader source: Energy.gov [DOE]

    With assistance from AMO, OLI Systems, Inc., developed the mixed-solvent electrolyte model, a comprehensive physical property package that can predict the properties of electrolyte systems ranging...

  4. Fuel Mix Disclosure

    Broader source: Energy.gov [DOE]

    In January 1999, the Colorado Public Utility Commission (PUC) adopted regulations requiring the state's utilities to disclose information regarding their fuel mix to retail customers. Utilities are...

  5. AP-XPS Measures MIEC Oxides in Action

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Measures MIEC Oxides in Action Print Wednesday, 25 May 2011 00:00 Oxide materials with mixed ionic-electronic conductivity (MIEC) can conduct both electrons and oxygen ions....

  6. SR0505

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    1, 2005 Media Contact: James Giusti (803) 952-7697 Subcontract Awarded for MOX Facility Site Preparation Work Flag Ribbon Art Aiken, S.C. - The National Nuclear Security Administration's (NNSA) plutonium disposition program has moved another step forward with the start of site preparation for its Mixed Oxide (MOX) Fuel Fabrication Facility at the Savannah River Site. NNSA announced today that the first subcontract for this work has been awarded to MDM Services Corporation, a small business

  7. ADVANCED MIXING MODELS

    SciTech Connect (OSTI)

    Lee, S.; Dimenna, R.; Tamburello, D.

    2011-02-14

    The process of recovering and processing High Level Waste (HLW) the waste in storage tanks at the Savannah River Site (SRS) typically requires mixing the contents of the tank with one to four mixers (pumps) located within the tank. The typical criteria to establish a mixed condition in a tank are based on the number of pumps in operation and the time duration of operation. To ensure that a mixed condition is achieved, operating times are typically set conservatively long. This approach results in high operational costs because of the long mixing times and high maintenance and repair costs for the same reason. A significant reduction in both of these costs might be realized by reducing the required mixing time based on calculating a reliable indicator of mixing with a suitably validated computer code. The focus of the present work is to establish mixing criteria applicable to miscible fluids, with an ultimate goal of addressing waste processing in HLW tanks at SRS and quantifying the mixing time required to suspend sludge particles with the submersible jet pump. A single-phase computational fluid dynamics (CFD) approach was taken for the analysis of jet flow patterns with an emphasis on the velocity decay and the turbulent flow evolution for the farfield region from the pump. Literature results for a turbulent jet flow are reviewed, since the decay of the axial jet velocity and the evolution of the jet flow patterns are important phenomena affecting sludge suspension and mixing operations. The work described in this report suggests a basis for further development of the theory leading to the identified mixing indicators, with benchmark analyses demonstrating their consistency with widely accepted correlations. Although the indicators are somewhat generic in nature, they are applied to Savannah River Site (SRS) waste tanks to provide a better, physically based estimate of the required mixing time. Waste storage tanks at SRS contain settled sludge which varies in

  8. Mixed waste characterization reference document

    SciTech Connect (OSTI)

    1997-09-01

    Waste characterization and monitoring are major activities in the management of waste from generation through storage and treatment to disposal. Adequate waste characterization is necessary to ensure safe storage, selection of appropriate and effective treatment, and adherence to disposal standards. For some wastes characterization objectives can be difficult and costly to achieve. The purpose of this document is to evaluate costs of characterizing one such waste type, mixed (hazardous and radioactive) waste. For the purpose of this document, waste characterization includes treatment system monitoring, where monitoring is a supplement or substitute for waste characterization. This document establishes a cost baseline for mixed waste characterization and treatment system monitoring requirements from which to evaluate alternatives. The cost baseline established as part of this work includes costs for a thermal treatment technology (i.e., a rotary kiln incinerator), a nonthermal treatment process (i.e., waste sorting, macronencapsulation, and catalytic wet oxidation), and no treatment (i.e., disposal of waste at the Waste Isolation Pilot Plant (WIPP)). The analysis of improvement over the baseline includes assessment of promising areas for technology development in front-end waste characterization, process equipment, off gas controls, and monitoring. Based on this assessment, an ideal characterization and monitoring configuration is described that minimizes costs and optimizes resources required for waste characterization.

  9. Mixing method and apparatus

    DOE Patents [OSTI]

    Green, Norman W.

    1982-06-15

    Method of mixing particulate materials comprising contacting a primary source and a secondary source thereof whereby resulting mixture ensues; preferably at least one of the two sources has enough motion to insure good mixing and the particulate materials may be heat treated if desired. Apparatus for such mixing comprising an inlet for a primary source, a reactor communicating therewith, a feeding means for supplying a secondary source to the reactor, and an inlet for the secondary source. Feeding means is preferably adapted to supply fluidized materials.

  10. Nearly discontinuous chaotic mixing

    SciTech Connect (OSTI)

    Sharp, David Howland [Los Alamos National Laboratory; Lim, Hyun K [STONYBROOK UNIV.; Yu, Yan [STONYBROOK UNIV.; Glimm, James G [STONYBROOK UNIV.

    2009-01-01

    A new scientific approach is presented for a broad class of chaotic problems involving a high degree of mixing over rapid time scales. Rayleigh-Taylor and Richtmyer-Meshkov unstable flows are typical of such problems. Microscopic mixing properties such as chemical reaction rates for turbulent mixtures can be obtained with feasible grid resolution. The essential dependence of (some) fluid mixing observables on transport phenomena is observed. This dependence includes numerical as well as physical transport and it includes laminar as well as turbulent transport. A new approach to the mathematical theory for the underlying equations is suggested.