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Sample records for cycle reprocessing light-water

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

    SciTech Connect

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

    2008-04-01

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

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

    SciTech Connect

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

    2013-11-05

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

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

    SciTech Connect

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

    2014-06-10

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

  4. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  5. Maintenance Cycle Extension in the IRIS Advanced Light Water Reactor Plant Design

    SciTech Connect

    Galvin, Mark R.; Todreas, Neil E.; Conway, Larry E.

    2003-09-15

    New nuclear power generation in the United States will be realized only if the economic performance can be made competitive with other methods of electrical power generation. The economic performance of a nuclear power plant can be significantly improved by increasing the time spent on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described that can be used to resolve, in the design phase, maintenance-related operating cycle length barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the International Reactor, Innovative and Secure (IRIS) design. IRIS is an advanced light water nuclear power plant that is being designed to maximize this on-line generating time by increasing the operating cycle length. This is consequently a maintenance strategy paper using the IRIS plant as the example.Potential IRIS operating cycle length maintenance-related barriers, determined by modification of an earlier operating pressurized water reactor (PWR) plant cycle length analysis to account for differences between the design of IRIS and this operating PWR, are presented. The proposed methodology to resolve these maintenance-related barriers by the design process is described. The results of applying the methodology to two potential IRIS cycle length barriers, relief valve testing and emergency heat removal system testing, are presented.

  6. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    SciTech Connect

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team indicates

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

    SciTech Connect

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

    2015-05-22

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

  8. Economic Analyiss of "Symbiotic" Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

    SciTech Connect

    Williams, Kent Alan; Shropshire, David E.

    2009-01-01

    A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle

  9. Technical aspects of fuel reprocessing

    SciTech Connect

    Groenier, W.S.

    1982-02-01

    The purpose of this paper is to present a brief description of fuel reprocessing and some present developments which show the reliability of nuclear energy as a long-term supply. The following topics are discussed: technical reasons for reprocessing; economic reasons for reprocessing; past experience; justification for advanced reprocessing R and D; technical aspects of current reprocessing development. The present developments are mainly directed at the reprocessing of breeder reactor fuels but there are also many applications to light-water reactor fuel reprocessing. These new developments involve totally remote operation, and maintenance. To demonstrate this advanced reprocessing concept, pilot-scale demonstration facilities are planned with commercial application occurring sometime after the year 2000. (ATT)

  10. Crystal chemistry of sodium zirconium phosphate based simulated ceramic waste forms of effluent cations (Ba(2+), Sn(4+), Fe(3+), Cr(3+), Ni(2+) and Si(4+)) from light water reactor fuel reprocessing plants.

    PubMed

    Shrivastava, O P; Chourasia, Rashmi

    2008-05-01

    A novel concept of immobilization of light water reactor (LWR) fuel reprocessing waste effluent through interaction with sodium zirconium phosphate (NZP) has been established. Such conversion utilizes waste materials like zirconium and nickel alloys, stainless steel, spent solvent tri-butyl phosphate and concentrated solution of NaNO(3). The resultant multi component NZP material is a physically and chemically stable single phase crystalline product having good mechanical strength. The NZP matrix can also incorporate all types of fission product cations in a stable crystalline lattice structure; therefore, the resultant solid solutions deserve quantification of crystallographic data. In this communication, crystal chemistry of the two types of simulated waste forms (type I-Na(1.49)Zr(1.56)Sn(0.02)Fe(0).(28)Cr(0.07)Ni(0.07)P(3)O(12) and type II-Na(1.35)Ba(0.14)Zr(1.56)Sn(0.02)Fe(0).(28)Cr(0.07)Ni(0.07)P(2.86)Si(0.14)O(12)) has been investigated using General Structure Analysis System (GSAS) programming of the X-ray powder diffraction data. About 4001 data points of each have been subjected to Rietveld analysis to arrive at a satisfactory structural convergence of Rietveld parameters; R-pattern (R(p))=0.0821, R-weighted pattern (R(wp))=0.1266 for type I and R(p)=0.0686, R(wp)=0.0910 for type II. The structure of type I and type II waste forms consist of ZrO(6) octahedra and PO(4) tetrahedra linked by the corners to form a three-dimensional network. Each phosphate group is on a two-fold rotation axis and is linked to four ZrO(6) octahedra while zirconium octahedra lies on a three-fold rotation axis and is connected to six PO(4) tetrahedra. Though the expansion along c-axis and shrinkage along a-axis with slight distortion of bond angles in the synthesized crystal indicate the flexibility of the structure, the waste forms are basically of NZP structure. Morphological examination by SEM reveals that the size of almost rectangular parallelepiped crystallites varies

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

    SciTech Connect

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

    2013-07-01

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

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

    SciTech Connect

    Parker, Frank L.

    2012-07-01

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

  13. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    SciTech Connect

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

    2012-04-01

    The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water

  14. Sustainability Considerations in Spent Light-water Nuclear Fuel Retrievability

    SciTech Connect

    Wood, Thomas W.; Rothwell, Geoffrey

    2012-01-10

    This paper examines long-term cost differences between two competing Light Water Reactor (LWR) fuels: Uranium Oxide (UOX) and Mixed Uranium Oxide-Plutonium Oxide (MOX). Since these costs are calculated on a life-cycle basis, expected savings from lower future MOX fuel prices can be used to value the option of substituting MOX for UOX, including the value of maintaining access to the used UOX fuel that could be reprocessed to make MOX. The two most influential cost drivers are the price of natural uranium and the cost of reprocessing. Significant and sustained reductions in reprocessing costs and/or sustained increases in uranium prices are required to give positive value to the retrievability of Spent Nuclear Fuel. While this option has positive economic value, it might not be exercised for 50 to 200 years. Therefore, there are many years for a program during which reprocessing technology can be researched, developed, demonstrated, and deployed. Further research is required to determine whether the cost of such a program would yield positive net present value and/or increases the sustainability of LWR energy systems.

  15. ASSESSMENT OF POSSIBLE CYCLE LENGTHS FOR FULLY-CERAMIC MICRO-ENCAPSULATED FUEL-BASED LIGHT WATER REACTOR CONCEPTS

    SciTech Connect

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal Pasamehmetoglu; Francesco Venneri

    2012-04-01

    The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO{sub 2} results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rates of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO{sub 2}-loaded TRISO particles-based fuel with enrichment at or below 20 w/o.

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

    SciTech Connect

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

    2010-09-22

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

  17. A study of a zone approach to IAEA (International Atomic Energy Agency) safeguards: The low-enriched-uranium zone of a light-water-reactor fuel cycle

    SciTech Connect

    Fishbone, L.G.; Higinbotham, W.A.

    1986-06-01

    At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the conclusions regarding the effectiveness of safeguards for the individual facilities within a state. In this study it was convenient to define three zones in a state with a closed light-water-reactor nuclear fuel cycle. Each zone contains those facilities or parts thereof which use or process nuclear materials of the same safeguards significance: low-enriched uranium, radioactive spent fuel, or recovered plutonium. The possibility that each zone might be treated as an extended material balance area for safeguards purposes is under investigation. The approach includes defining the relevant features of the facilities in the three zones and listing the safeguards activities which are now practiced. This study has focussed on the fresh-fuel zone, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. There are a number of possible safeguards approaches which fall between the two extremes. The intention is to develop a rational approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the approach involving the zone as a material balance area, and for some reasonable intermediate safeguards approaches.

  18. Light water reactor program

    SciTech Connect

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  19. Light water detritiation

    SciTech Connect

    Fedorchenko, O.A.; Aleksee, I.A.; Bondarenko, S.D.; Vasyanina, T.V.

    2015-03-15

    Hundreds of thousands of tons of tritiated light water have been accumulating from the enterprises of nuclear fuel cycles around the world. The Dual-Temperature Water-Hydrogen (DTWH) process looks like the only practical alternative to Combined Electrolysis and Catalytic Exchange (CECE). In DTWH power-consuming lower reflux device (electrolytic cell) is replaced by a so-called 'hot tower' (LPCE column operating at conditions which ensure relatively small value of elementary separation factor α(hot)). In the upper, cold tower, the tritium transfers from hydrogen to water while in the lower, hot tower - in the opposite direction - from water to hydrogen. The DTWH process is much more complicated compared to CECE; it must be thoroughly computed and strictly controlled by an automatic control system. The use of a simulation code for DTWH is absolutely important. The simulation code EVIO-5 deals with 3 flows inside a column (hydrogen gas, water vapour and liquid water) and 2 simultaneous isotope exchange sub-processes (counter-current phase exchange and co-current catalytic exchange). EVIO-5 takes into account the strong dependence of process performance on given conditions (temperature and pressure). It calculates steady-state isotope concentration profiles considering a full set of reversible exchange reactions between different isotope modifications of water and hydrogen (12 molecular species). So the code can be used for simulation of LPCE column operation for detritiation of hydrogen and water feed, which contains H and D not only at low concentrations but above 10 at.% also. EVIO-5 code is used to model a Tritium Removal Facility with a throughput capacity of about 400 m{sup 3}/day. Simulation results show that a huge amount of wet-proofed catalyst is required (about 6000 m{sup 3}), mainly (90%) in the first stage. One reason for these large expenses (apart from a big scale of the problem itself) is the relatively high tritium separation factor in the hot tower

  20. Nuclear Fuel Reprocessing

    SciTech Connect

    Harold F. McFarlane; Terry Todd

    2013-11-01

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

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

    SciTech Connect

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

    2003-07-01

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

  2. Evaluating Environmental, Health and Safety Impacts from Two Nuclear Fuel Cycles: A Comparative Analysis of Once-Through Uranium Use and Plutonium Recycle in Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Smith, Bethan L.

    The work presented in this dissertation represents a systems-level approach to investigate potential net impacts with respect to human health and the environment associated with transitioning to the MOC for the U.S. In Chapter 2, an updated systems-level conceptual model of the OTC is presented to more accurately portray the OTC as currently implemented in the U.S. The conceptual model is the basis for estimating the worker collective doses at each operational stage, and the first demonstration of a quantitative comparative radiological impact assessment from expected normal operations is presented. In the course of evaluating worker collective dose associated with modern OTC practices, it was found that the relative contributions from the two grouped operations (front-end operations for preparing reactor fuel and reactor operations) were substantially different from historical data and conventional wisdom. As a bookend to Chapter 2, a summary is provided that describes the nature of the differences and factors that led to these differences. Detailed information of the work as part of the published journal article based off of this corollary work is included as an Appendix (C). In Chapter 3, the study of worker collective doses from the phased introduction of reprocessing in the MOC scenario, and is presented similarly to the results in Chapter 2. MOC performance was also estimated by evaluating the radioactive waste generated that can be disposed and managed through known disposal practices in shallow-land burial. Relative to the OTC, MOC performance with respect to worker collective dose was not discernibly different; while the volume of radioactive waste generated decreased. It was found that although the sheer volume of radioactive waste avoided is large, the waste disposition pathway is known for the majority of this waste. The radioactive waste that requires disposal at a licensed off-site facility is examined in closer detail. The verification process for

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

    NASA Astrophysics Data System (ADS)

    Straka, Martin

    2016-12-01

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

  4. Light water reactor health physics.

    PubMed

    Prince, Robert J; Bradley, Scott E

    2005-06-01

    In this article an overview of the historical development of light water reactor health physics programs is presented. Operational health physics programs have developed and matured as experience in operating and maintaining light water reactors has been gained. Initial programs grew quickly in both size and complexity with the number and size of nuclear units under construction and in operation. Operational health physics programs evolved to face various challenges confronted by the nuclear industry, increasing the effectiveness of radiological safety measures. Industry improvements in radiological safety performance have resulted in significant decreases in annual collective exposures from a high value of 790 person-rem in 1980 to 117 person-rem per reactor in 2002. Though significant gains have been made, the continued viability of the nuclear power industry is confronted with an aging workforce, as well as the challenges posed by deregulation and the need to maintain operational excellence.

  5. Light water reactor health physics.

    PubMed

    Prince, Robert J; Bradley, Scott E

    2004-11-01

    In this article an overview of the historical development of light water reactor health physics programs is presented. Operational health physics programs have developed and matured as experience in operating and maintaining light water reactors has been gained. Initial programs grew quickly in both size and complexity with the number and size of nuclear units under construction and in operation. Operational health physics programs evolved to face various challenges confronted by the nuclear industry, increasing the effectiveness of radiological safety measures. Industry improvements in radiological safety performance have resulted in significant decreases in annual collective exposures from a high value of 790 person-rem in 1980 to 117 person-rem per reactor in 2002. Though significant gains have been made, the continued viability of the nuclear power industry is confronted with an aging workforce, as well as the challenges posed by deregulation and the need to maintain operational excellence.

  6. Nuclear Fuel Reprocessing

    SciTech Connect

    Michael F. Simpson; Jack D. Law

    2010-02-01

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

  7. LIGHT WATER MODERATED NEUTRONIC REACTOR

    DOEpatents

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  8. Reprocessing Microflare Data

    NASA Technical Reports Server (NTRS)

    Ryan, James M.

    1999-01-01

    that covers solar maximum and constructed the microflare distribution shown in the figure. The results of the automated search start to deviate from the manual search results below about 1000/s. Not only do we now have this distribution but we have a database of solar microflares that was used to construct the distribution. This database contains the signal at higher energy channels as well as that in channel zero (and below). From this one can, using software at GSFC, construct a photon spectrum for some of the larger microflares. It can also be used in other solar studies, especially those that correlate the X-ray flux with emission at other wavelengths. With some additional effort we hope to integrate this database into the corresponding one residing at the Solar Data Analysis Center at GSFC. The entire CGRO mission's data can now be reprocessed to obtain the microflare distribution at all phases of the solar cycle. This work is in progress. The results of this work will be presented in forthcoming scientific workshops and conferences.

  9. Light-Water Breeder Reactor

    DOEpatents

    Beaudoin, B. R.; Cohen, J. D.; Jones, D. H.; Marier, Jr, L. J.; Raab, H. F.

    1972-06-20

    Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

  10. Spent nuclear fuel reprocessing modeling

    SciTech Connect

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

    2013-07-01

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

  11. Comparison of the radiological impacts of thorium and uranium nuclear fuel cycles

    SciTech Connect

    Meyer, H.R.; Witherspoon, J.P.; McBride, J.P.; Frederick, E.J.

    1982-03-01

    This report compares the radiological impacts of a fuel cycle in which only uranium is recycled, as presented in the Final Generic Environmental Statement on the Use of Recycle Plutonium in Mixed Oxide Fuel in Light Water Cooled Reactors (GESMO), with those of the light-water breeder reactor (LWBR) thorium/uranium fuel cycle in the Final Environmental Statement, Light Water Breeder Reactor Program. The significant offsite radiological impacts from routine operation of the fuel cycles result from the mining and milling of thorium and uranium ores, reprocessing spent fuel, and reactor operations. The major difference between the impacts from the two fuel cycles is the larger dose commitments associated with current uranium mining and milling operations as compared to thorium mining and milling. Estimated dose commitments from the reprocessing of either fuel type are small and show only moderate variations for specific doses. No significant differences in environmental radiological impact are anticipated for reactors using either of the fuel cycles. Radiological impacts associated with routine releases from the operation of either the thorium or uranium fuel cycles can be held to acceptably low levels by existing regulations.

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

    SciTech Connect

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

    2009-02-01

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

  13. Reprocessing in Luminous Disks

    NASA Technical Reports Server (NTRS)

    Bell, K. Robbins; DeVincenzi, Donald L. (Technical Monitor)

    1999-01-01

    We develop and investigate a procedure that accounts for disk reprocessing of photons that originate in the disk itself. Surface temperatures and simple, black body spectral energy distributions (SEDs) of protostellar disks are calculated. In disks that flare with radius, reprocessing of stellar photons results in temperature profiles considerably shallower than r(sup -3/4). Including the disk as a radiation source (as in the case of actively secreting disks) along with the stellar source further flattens the temperature profile. Disks that flare strongly near the star and then smoothly curve over and become shadowed at some distance ("decreasing curvature" disks) exhibit nearly power-law temperature profiles which result in power-law infrared SEDs with slopes in agreement with typical observations of young stellar objects. Disk models in which the photospheric thickness is controlled by the local opacity and in which the temperature decreases with radius naturally show this shape. Uniformly flaring models do not match observations as well; progressively stronger reprocessing at larger radii leads to SEDs that flatten toward the infrared or even have a second peak at the wavelength corresponding (through the Wien law) to the temperature of the outer edge of the disk. In FU Orionis outbursting systems, the dominant source of energy is the disk itself. The details of the reprocessing depend sensitively on the assumed disk shape and emitted temperature profile. The thermal instability outburst models of Bell Lin reproduce trends in the observed SEDs of Fuors with T varies as r(sup -3/4) in the inner disk (r approx. less than 0.25au corresponding to lambda approx. less than 10 microns) and T varies as r(sup -1/2) in the outer disk. Surface irradiation during outburst and quiescence is compared in the region of planet formation (1 - 10 au). The contrast between the two phases is diminished by the importance of the reprocessing of photons from the relatively high mass

  14. Effects of reprocessing on nanoalumina polymer composites

    NASA Astrophysics Data System (ADS)

    Huang, Chunchia

    The life cycle of reprocessed polymer nanocomposites is a critical factor associated with their growing use, but the limited work on reprocessing of nanocomposites has focused solely on the effects of organoclays. This research investigated of the structure and property changes during reprocessing of polypropylene (PP) and polycarbonate (PC) nanocomposites with 3 wt. % nanoalumina. Neat PP and PC were used as controls. Reprocessing of the neat polymers and nanocomposites produced no indication of oxidation (in FTIR), no changes in the glass transition temperature of PC and the melting temperatures of PP, and no changes in thermal stability (as measured using thermogravimetric analysis). Significant decreases, however, occurred in the melt viscosity of the materials. The introduction of nanoalumina during twin screw extrusion also produced a significant decrease in the viscosity and a 10°C decrease in the glass transition temperature of the PC nanocomposite. Color changes did not correspond to the chain scission in PP and PC; neat PP and PP composite yellowed, neat PC turned brown, and the PC nanocomposite did not change color. Dispersion of the nanoalumina in both PP and PC improved with repeated reprocessing, the crystallinity in the PP/nanoalumina composites remained constant. The Young's moduli of both the PP and PP/A12O3 nanocomposite were similar, whereas the Young's modulus values of the PC/A1 2O3 nanocomposite was slighted lower than that of the neat PC. In contrast, the elongations at break of the PP/A12O3 and PC/A12O3 nanocomposites were, respectively, 50% and 16% of the values measured for the neat resins. All modulus and elongation at break values, however, remained constant over five reprocessing cycles. This behavior suggests that the major degradation mechanism during reprocessing of neat PP and PP nanocomposites was thermal-mechanical polymer chain scission and that the nanoalumina enhanced this degradation in the PC/nanoalumina. This additional

  15. Environmentally assisted cracking in light water reactors

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  16. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    NASA Astrophysics Data System (ADS)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

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

    SciTech Connect

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

    1980-04-01

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

  18. Reprocessing of nonoptimally exposed holograms

    SciTech Connect

    Phipps, G.S.; Robertson, C.E.; Tamashiro, F.M.

    1980-03-01

    Two reprocessing techniques have been investigated that are capable of correcting the effects of nonoptimum optical density of photographic amplitude holograms recorded on Agfa-Gevaert type 10E75 plates. In some cases a reprocessed hologram will exhibit a diffraction efficiency even higher than that obtainable from a hologram exposed and processed to the optimum density. The SNR of the reprocessed holograms is much higher than that of the same holograms belached with cupric bromide. In some cases the SNR approaches the optimum value for a properly exposed amplitude hologram. Subjective image quality and resolution of reprocessed hologram reconstructins appear to be no different than for normal single-development holograms. Repeated reprocessing is feasible and in some cases desirable as a means of increasing diffraction efficiency.

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

    SciTech Connect

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

    1990-01-01

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

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

    SciTech Connect

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

    1990-01-01

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

  1. Plasma coal reprocessing

    NASA Astrophysics Data System (ADS)

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

    2013-12-01

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

  2. Risks of nuclear fuel reprocessing

    SciTech Connect

    Durant, W.S.

    1990-01-01

    The Savannah River Site's primary function is the production of weapons materials. It consists of four reactors, two fuel reprocessing facilities, a fuel fabrication facility, a nuclear fuel facility for the Navy and a heavy water recycle facility. Under construction is a facility to convert the site's liquid wastes into borosilicate glass. The topic of this paper is risks of nuclear fuel reprocessing. Also discussed are facility operations. 18 figs.

  3. Steps of Reprocessing and Equipments

    PubMed Central

    Lee, Yong Kook

    2013-01-01

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

  4. Classic Nuclear Fuel Reprocessing Flowsheet

    SciTech Connect

    Fallgren, Andrew James

    2015-02-13

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

  5. Multi-Applications Small Light Water Reactor - NERI Final Report

    SciTech Connect

    S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.

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

    SciTech Connect

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

    2013-07-22

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

  7. Data validation and security for reprocessing.

    SciTech Connect

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

    2008-10-01

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

  8. Consolidated Fuel Reprocessing Program. Progress report, January 1-March 31, 1985

    SciTech Connect

    Not Available

    1985-04-01

    The DOE has concentrated all US research and development on fuel reprocessing into one major program - the Consolidated Fuel Reprocessing Program (CFRP) - under the management of the Oak Ridge National Laboratory and the Oak Ridge Operations Office. Other major program participants are GA Technologies, Inc., where reprocessing research and development on the HTGR fuel cycle are done, and the Hanford Engineering Development Laboratory (HEDL). The coverage is generally overview in nature. Experimental details and data have been limited to (1) make the report more concise and (2) meet the requirements which would qualify the report for unrestricted distribution in the open literature. All research and development on civilian power reactor fuel reprocessing in the United States is managed under the Consolidated Fuel Reprocessing Program (CFRP) centered at Oak Ridge National Laboratory (ORNL). Technical progress is reported in overview fashion in this series of quarterly progress reports.

  9. FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION

    SciTech Connect

    Carter, J.

    2011-01-03

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

  10. FUEL CYCLE POTENTIAL WASTE FOR DISPOSITION

    SciTech Connect

    Jones, R.; Carter, J.

    2010-10-13

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

  11. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    SciTech Connect

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  12. Methodology for estimating reprocessing costs for nuclear fuels

    SciTech Connect

    Carter, W. L.; Rainey, R. H.

    1980-02-01

    A technological and economic evaluation of reprocessing requirements for alternate fuel cycles requires a common assessment method and a common basis to which various cycles can be related. A methodology is described for the assessment of alternate fuel cycles utilizing a side-by-side comparison of functional flow diagrams of major areas of the reprocessing plant with corresponding diagrams of the well-developed Purex process as installed in the Barnwell Nuclear Fuel Plant (BNFP). The BNFP treats 1500 metric tons of uranium per year (MTU/yr). Complexity and capacity factors are determined for adjusting the estimated facility and equipment costs of BNFP to determine the corresponding costs for the alternate fuel cycle. Costs of capacities other than the reference 1500 MT of heavy metal per year are estimated by the use of scaling factors. Unit costs of reprocessed fuel are calculated using a discounted cash flow analysis for three economic bases to show the effect of low-risk, typical, and high-risk financing methods.

  13. Commercial Light Water Reactor Tritium Extraction Facility

    SciTech Connect

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  14. Light water reactor lower head failure analysis

    SciTech Connect

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

  15. Fabrication of light water reactor tritium targets

    SciTech Connect

    Pilger, J.P.

    1991-11-01

    The mission of the Fabrication Development Task of the Tritium Target Development Project is: to produce a documented technology basis, including specifications and procedures for target rod fabrication; to demonstrate that light water tritium targets can be manufactured at a rate consistent with tritium production requirements; and to develop quality control methods to evaluate target rod components and assemblies, and establish correlations between evaluated characteristics and target rod performance. Many of the target rod components: cladding tubes, end caps, plenum springs, etc., have similar counterparts in LWR fuel rods. High production rate manufacture and inspection of these components has been adequately demonstrated by nuclear fuel rod manufacturers. This summary describes the more non-conventional manufacturing processes and inspection techniques developed to fabricate target rod components whose manufacturability at required production rates had not been previously demonstrated.

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

    SciTech Connect

    Kleykamp, H.

    1988-03-01

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

  17. Advanced Safeguards Approaches for New Reprocessing Facilities

    SciTech Connect

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

    2007-06-24

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

  18. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    SciTech Connect

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  19. Commercial Nuclear Reprocessing in the United States

    SciTech Connect

    Sherrill, Charles Leland; Balatsky, Galya Ivanovna

    2015-09-09

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

  20. Clinical Practice Guidelines for Endoscope Reprocessing

    PubMed Central

    Oh, Hyun Jin

    2015-01-01

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

  1. Recent Update of Gastrointestinal Endoscope Reprocessing

    PubMed Central

    Hong, Kyong Hee

    2013-01-01

    As infection-related issues have become one of the most important concerns in endoscopy centers, proper reprocessing of endoscopes has attracted great interest. Compliance with established guidelines for reprocessing is critical to prevent pathogen transmission. However, hospital compliance with guidelines has not been satisfactory. To increase compliance, efforts have focused on developing new and more innovative disinfectants and an automated endoscope reprocessor. Reprocessing must be performed by appropriately trained personnel and regular monitoring of reprocessing is essential for quality assurance to improve compliance. PMID:23767038

  2. Endoscope Reprocessing: Update on Controversial Issues

    PubMed Central

    Choi, Hyun Ho

    2015-01-01

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

  3. Evaluation and quantification of reprocessing modification in single-use devices in interventional cardiology

    NASA Astrophysics Data System (ADS)

    Tessarolo, Francesco; Ferrari, Paolo; Silvia, Bortoluzzi; Motta, Antonella; Migliaresi, Claudio; Zennaro, Lucio; Rigo, Adelio; Guarrera, Giovanni Maria; Nollo, Giandomenico

    2004-11-01

    The increasing demand in interventional cardiology urges for reprocessing of single-use-labelled medical devices. To fulfil this aim, accurate and validated regeneration protocols are mandatory to guarantee sterility, functionality and safeness. The reprocessing protocol was realized by decontamination with chloro-donors, cleaning with enzymatic solutions and hydrogen peroxide gas plasma sterilization. Reprocessing effects on ablation and electrophysiology catheters were evaluated by assessing physical-chemical changes on surfaces and bulks, as a function of the reprocessing cycles number. Conventional optical microscopy and environmental scanning electron microscopy (ESEM) underlined the presence of micro-scratches on the polyurethane shaft surface. A clear correlation was found between surface damages and number of reprocessing cycles. Atomic force microscopy (AFM) confirmed the occurrence of physical-chemical etching of the polyurethane shaft caused by the hydrogen peroxide plasma sterilization, with increasing of nano-roughness at increasing number of the reprocessing cycles. UV-Vis spectra performed on the incubation solution of polymeric shaft sample, showed an absorbance increase at about 208 nm. This fact could be attributed to the water elution from the polymer of low molecular weight oligomers. The presence of hydrolysis products of the polymeric shaft after incubation demands both the characterization of the products released in the solution and the chemical characterization of the water exposed surface.

  4. Transformative monitoring approaches for reprocessing.

    SciTech Connect

    Cipiti, Benjamin B.

    2011-09-01

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

  5. Consolidated Fuel Reprocessing Program. National Program Plan FY 1984

    SciTech Connect

    Not Available

    1984-01-01

    This FY 1984 National Program Plan for the Consolidated Fuel Reprocessing Program (CFRP) provides specific guidance from the Department of Energy (DOE) for FY 1984 CFRP activities and a baseline for activities in future years. This is the second issue of the Plan, which is updated anually and summarizes program objectives, summary plans and schedules, budget allocations, contractor involvement, and interfaces with other research programs. The National Program Plan is a controlling document for the Technical Program Plan, which is prepared annually by the CFRP at Oak Ridge National Laboratory (ORNL) and is one of the hierarchical group of planning documents. The CFRP is a part of the DOE's program of research and development (R and D) on nuclear fission systems and is directed by the LMFBR Fuel Cycle Projects Division, Office of Spent Fuel Management and Reprocessing Systems at DOE-Headquarters through the Oak Ridge Operations Office (ORO). The strategy of the program is to maintain the capability to commit to a breeder option through a strong R and D program on breeder reprocessing and alternate fuels and fuel cycles in order to achieve operating and economic advantages.

  6. Consolidated Fuel Reprocessing Program: National Program Plan, FY 1983

    SciTech Connect

    Not Available

    1983-01-01

    This FY 1983 National Program Plan for the Consolidated Fuel Reprocessing Program (CFRP) provides specific guidance from the Department of Energy (DOE) for FY 1983 CFRP activities and a baseline for future year activities. This initial issue of the Plan, which will be updated annually, summarizes program objectives, summary plans and schedules, budget allocations, contractor involvement, and interfaces with other research programs. The National Program Plan is a controlling document for the Technical Program Plan, which is prepared annually by the CFRP at ORNL and is one of a hierarchical group of planning documents. The CFRP is a part of the DOE's program of research and development (R and D) on nuclear fission systems and is directed by the LMFBR Fuel Cycle Projects Division, Office of Spent Fuel Management and Reprocessing Systems at DOE-Headquarters through the Oak Ridge Operations Office (ORO). The strategy of the program is to maintain the capability to commit to a breeder option through a strong R and D program on breeder reprocessing and alternate fuels and fuel cycles to achieve operating and economic advantages.

  7. Thorium-Based Transmuter Fuels for Light Water Reactors

    SciTech Connect

    J. Stephen Herring; P. E. MacDonald; K. Weaver

    2004-04-01

    A light water reactor (LWR) fuel cycle is proposed where the reactor core mainly consists of standard uranium-dioxide (UO2) fuel rods with typical 235U enrichment, along with thoria-urania (ThO2-UO2) or yttria-stablized zirconia fertile-free fuel rods containing the plutonium and minor actinides typical of 30-yr old UO2 fuel in 1/9 to 1/3 of the positions. The goals of this mono-recycling strategy or "twice through fuel cycle" are to transmute the great majority of the long lived actinides in existing LWRs and to discharge a fuel form that is a very robust waste form and whose isotopic content is very proliferation resistant. The incorporation of plutonium into a ThO2 or yttria-stablized zirconia fertile-free matrix results in the consumption of already-separated plutonium without breeding significant additional 239Pu. The minor actinides (i.e., neptunium, americium, curium, berkelium, californium, etc.) are also included in the ThO2 or fertile-free transmuter fuel rods to further reduce the overall long-term radiotoxicity of the fuel cycle. Our analyses have shown that thorium-based or fertile-free fuels can reduce the amount of 239Pu needing further transmutation or going to a repository by ~90%. Also, thorium-based fuels produce a mixture of plutonium isotopes high in 238Pu. Because of the high decay heat and spontaneous neutron generation of 238Pu, this isotope provides intrinsic proliferation resistance.

  8. Thorium-Based Transmuter Fuels for Light Water Reactors

    SciTech Connect

    Herring, J. Stephen; MacDonald, Philip E.; Weaver, Kevan D.

    2004-07-15

    A light water reactor (LWR) fuel cycle is proposed where the reactor core mainly consists of standard uranium-dioxide (UO{sub 2}) fuel rods with typical {sup 235}U enrichment, along with thoria-urania (ThO{sub 2}-UO{sub 2}) or yttria-stablized zirconia fertile-free fuel rods containing the plutonium and minor actinides typical of 30-yr old UO{sub 2} fuel in 1/9 to 1/3 of the positions. The goals of this mono-recycling strategy or 'twice through fuel cycle' are to transmute the great majority of the long lived actinides in existing LWRs and to discharge a fuel form that is a very robust waste form and whose isotopic content is very proliferation resistant. The incorporation of plutonium into a ThO{sub 2} or yttria-stablized zirconia fertile-free matrix results in the consumption of already-separated plutonium without breeding significant additional {sup 239}Pu. The minor actinides (i.e., neptunium, americium, curium, berkelium, californium, etc.) are also included in the ThO{sub 2} or fertile-free transmuter fuel rods to further reduce the overall long-term radiotoxicity of the fuel cycle. Our analyses have shown that thorium-based or fertile-free fuels can reduce the amount of {sup 239}Pu needing further transmutation or going to a repository by {approx}90%. Also, thorium-based fuels produce a mixture of plutonium isotopes high in {sup 238}Pu. Because of the high decay heat and spontaneous neutron generation of {sup 238}Pu, this isotope provides intrinsic proliferation resistance.

  9. Environmentally assisted cracking in light water reactors.

    SciTech Connect

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.

    2007-11-06

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2002. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low-alloy steels and austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs in BWRs, (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in PWRs, and (d) cracking in Ni-alloys and welds. A critical review of the ASME Code fatigue design margins and an assessment of the conservation in the current choice of design margins are presented. The existing fatigue {var_epsilon}-N data have been evaluated to define the effects of key material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Experimental data are presented on the effects of surface roughness on fatigue crack initiation in these materials in air and LWR environments. Crack growth tests were performed in BWR environments on SSs irradiated to 0.9 and 2.0 x 10{sup 21} n x cm{sup -2}. The crack growth rates (CGRs) of the irradiated steels are a factor of {approx}5 higher than the disposition curve proposed in NUREG-0313 for thermally sensitized materials. The CGRs decreased by an order of magnitude in low-dissolved oxygen (DO) environments. Slow-strain-rate tensile (SSRT) tests were conducted in high-purity 289 C water on steels irradiated to {approx}3 dpa. The bulk S content correlated well with the susceptibility to intergranular SCC in 289 C water. The IASCC susceptibility of SSs that contain >0.003 wt. % S increased drastically. bend tests in inert environments at 23 C were conducted on broken pieces of SSRT specimens and on unirradiated specimens of the same materials after hydrogen charging. The results of the tests and a review of other data in the literature

  10. Light Water Reactor Sustainability Accomplishments Report

    SciTech Connect

    McCarthy, Kathryn A.

    2015-02-01

    Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you

  11. Isotopic signature of atmospheric xenon released from light water reactors.

    PubMed

    Kalinowski, Martin B; Pistner, Christoph

    2006-01-01

    A global monitoring system for atmospheric xenon radioactivity is being established as part of the International Monitoring System to verify compliance with the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The isotopic activity ratios of (135)Xe, (133m)Xe, (133)Xe and (131m)Xe are of interest for distinguishing nuclear explosion sources from civilian releases. Simulations of light water reactor (LWR) fuel burn-up through three operational reactor power cycles are conducted to explore the possible xenon isotopic signature of nuclear reactor releases under different operational conditions. It is studied how ratio changes are related to various parameters including the neutron flux, uranium enrichment and fuel burn-up. Further, the impact of diffusion and mixing on the isotopic activity ratio variability are explored. The simulations are validated with reported reactor emissions. In addition, activity ratios are calculated for xenon isotopes released from nuclear explosions and these are compared to the reactor ratios in order to determine whether the discrimination of explosion releases from reactor effluents is possible based on isotopic activity ratios.

  12. The Fourth (A)ATSR Data Reprocessing

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

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

  13. Closed Fuel Cycle Waste Treatment Strategy

    SciTech Connect

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

    2015-02-01

    This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significant additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form

  14. Light-water breeder reactors: preliminary safety and environmental information document. Volume III

    SciTech Connect

    Not Available

    1980-01-01

    Information is presented concerning prebreeder and breeder reactors based on light-water-breeder (LWBR) Type 1 modules; light-water backfit prebreeder supplying advanced breeder; light-water backfit prebreeder/seed-blanket breeder system; and light-water backfit low-gain converter using medium-enrichment uranium, supplying a light-water backfit high-gain converter.

  15. Actuation of liquid crystal elastomers reprocessed with carbon nanoparticles

    NASA Astrophysics Data System (ADS)

    Chambers, Martin; Zalar, Boštjan; Remškar, Maja; Žumer, Slobodan; Finkelmann, Heino

    2006-12-01

    Liquid crystal elastomers are currently of great interest due to their large thermally stimulated changes in shape. Here the authors show that by using an existing network and conducting carbon nanoparticles dispersed in a solvent with high swelling capability, a surface integrated layer can be created. This layer allows the effective resistivity to be reduced from highly insulating to usable values for electrical actuation and withstands large changes in geometrical shape both in contraction and expansion. Utilizing a resistive "Joule" heating effect, the reprocessed system shows a 150% length change and can be cycled beyond 10kcycles.

  16. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect

    McCarthy, Kathryn A.; Busby, Jeremy; Hallbert, Bruce; Bragg-Sitton, Shannon; Smith, Curtis; Barnard, Cathy

    2014-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  17. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect

    Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

    2013-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  18. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect

    George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

    2012-01-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

  19. Consolidated Fuel Reprocessing Program. National Program Plan, FY 1985

    SciTech Connect

    Not Available

    1985-03-01

    This FY 1985 National Program Plan for the Consolidated Fuel Reprocessing Program (CFRP) provides specific guidance from the Department of Energy (DOE) for FY 1985 CFRP activities and a baseline for activities in future years. This is the third issue of the Plan, which is updated annually and summarizes program objectives, plans, and schedules, budget allocations, contractor involvements, and interfaces with other research programs. The National Program Plan is a controlling document for the Technical Program Plan, which is prepared annually by the CFRP at Oak Ridge National Laboratory (ORNL) and is one of a hierarchical group of planning documents. The CFRP is a part of the DOE's program of research and development (R and D) on nuclear fission systems and is directed by the LMFBR Fuel Cycle Projects Division, Office of Spent Fuel Management and Reprocessing Systems at DOE-Headquarters through the Oak Ridge Operations Office (ORO). The strategy of the program is to maintain the capability to commit to a breeder option through a strong fuel cycle R and D program and international technical exchanges.

  20. Development of Advanced High Uranium Density Fuels for Light Water Reactors

    SciTech Connect

    Blanchard, James; Butt, Darryl; Meyer, Mitchell; Xu, Peng

    2016-02-15

    This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

  1. MicroRaman Measurements for Nuclear Fuel Reprocessing Applications

    SciTech Connect

    Casella, Amanda; Lines, Amanda; Nelson, Gilbert; Bello, Job; Bryan, Samuel

    2016-01-01

    Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper will focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. This paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO3, and UO2(NO3)2.

  2. MicroRaman measurements for nuclear fuel reprocessing applications

    DOE PAGES

    Casella, Amanda; Lines, Amanda; Nelson, Gilbert; ...

    2016-12-01

    Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper willmore » focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. Finally, this paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO3, and UO2(NO3)2.« less

  3. MicroRaman measurements for nuclear fuel reprocessing applications

    SciTech Connect

    Casella, Amanda; Lines, Amanda; Nelson, Gilbert; Bello, Job; Bryan, Samuel

    2016-12-01

    Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper will focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. Finally, this paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO3, and UO2(NO3)2.

  4. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    SciTech Connect

    Steven J. Piet; Samuel E. Bays; Michael A. Pope; Gilles J. Youinou

    2010-11-01

    From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

  5. Multi-Application Small Light Water Reactor Final Report

    SciTech Connect

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept. Applications such as

  6. Summary of nuclear fuel reprocessing activities around the world

    SciTech Connect

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

    1984-11-01

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

  7. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    SciTech Connect

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

    2014-09-01

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

  8. EOS Data Products Latency and Reprocessing Evaluation

    NASA Astrophysics Data System (ADS)

    Ramapriyan, H. K.; Wanchoo, L.

    2012-12-01

    NASA's Earth Observing System (EOS) Data and Information System (EOSDIS) program has been processing, archiving, and distributing EOS data since the launch of Terra platform in 1999. The EOSDIS Distributed Active Archive Centers (DAACs) and Science-Investigator-led Processing Systems (SIPSs) are generating over 5000 unique products with a daily average volume of 1.7 Petabytes. Initially EOSDIS had requirements to make process data products within 24 hours of receiving all inputs needed for generating them. Thus, generally, the latency would be slightly over 24 and 48 hours after satellite data acquisition, respectively, for Level 1 and Level 2 products. Due to budgetary constraints these requirements were relaxed, with the requirement being to avoid a growing backlog of unprocessed data. However, the data providers have been generating these products in as timely a manner as possible. The reduction in costs of computing hardware has helped considerably. It is of interest to analyze the actual latencies achieved over the past several years in processing and inserting the data products into the EOSDIS archives for the users to support various scientific studies such as land processes, oceanography, hydrology, atmospheric science, cryospheric science, etc. The instrument science teams have continuously evaluated the data products since the launches of EOS satellites and improved the science algorithms to provide high quality products. Data providers have periodically reprocessed the previously acquired data with these improved algorithms. The reprocessing campaigns run for an extended time period in parallel with forward processing, since all data starting from the beginning of the mission need to be reprocessed. Each reprocessing activity involves more data than the previous reprocessing. The historical record of the reprocessing times would be of interest to future missions, especially those involving large volumes of data and/or computational loads due to

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

    SciTech Connect

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

    1992-10-01

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

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

    SciTech Connect

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

    1992-10-01

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

  11. Evaluation of retention and disposal options for tritium in fuel reprocessing

    SciTech Connect

    Grimes, W.R.; Hampson, D.C.; Larkin, D.J.; Skolrud, J.O.; Benjamin, R.W.

    1982-08-01

    Five options were evaluated as means of retaining tritium released from light-water reactor or fast breeder reactor fuel during the head-end steps of a typical Purex reprocessing scheme. Cost estimates for these options were compared with a base case in which no retention of tritium within the facility was obtained. Costs were also estimated for a variety of disposal methods of the retained tritium. The disposal costs were combined with the retention costs to yield total costs (capital plus operating) for retention and disposal of tritium under the conditions envisioned. The above costs were converted to an annual basis and to a dollars per curie retained basis. This then was used to estimate the cost in dollars per man-rem saved by retaining the tritium. Only the options that used the least expensive disposal costs could approach the $1000/man-rem cost used as a guide by the Nuclear Regulatory Commission.

  12. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

    SciTech Connect

    Mac Donald, Philip Elsworth

    2002-06-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  13. Assessment of light water reactor accident management programs and experience

    SciTech Connect

    Hammersley, R.J.

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  14. Monitoring of endoscope reprocessing with an adenosine triphosphate (ATP) bioluminescence method

    PubMed Central

    Parohl, Nina; Stiefenhöfer, Doris; Heiligtag, Sabine; Reuter, Henning; Dopadlik, Dana; Mosel, Frank; Gerken, Guido; Dechêne, Alexander; Heintschel von Heinegg, Evelyn; Jochum, Christoph; Buer, Jan; Popp, Walter

    2017-01-01

    Background: The arising challenges over endoscope reprocessing quality proposes to look for possibilities to measure and control the process of endoscope reprocessing. Aim: The goal of this study was to evaluate the feasibility of monitoring endoscope reprocessing with an adenosine triphosphate (ATP) based bioluminescence system. Methods: 60 samples of eight gastroscopes have been assessed from routine clinical use in a major university hospital in Germany. Endoscopes have been assessed with an ATP system and microbial cultures at different timepoints during the reprocessing. Findings: After the bedside flush the mean ATP level in relative light units (RLU) was 19,437 RLU, after the manual cleaning 667 RLU and after the automated endoscope reprocessor (AER) 227 RLU. After the manual cleaning the mean total viable count (TVC) per endoscope was 15.3 CFU/10 ml, and after the AER 5.7 CFU/10 ml. Our results show that there are reprocessing cycles which are not able to clean a patient used endoscope. Conclusion: Our data suggest that monitoring of flexible endoscope with ATP can identify a number of different influence factors, like the endoscope condition and the endoscopic procedure, or especially the quality of the bedside flush and manual cleaning before the AER. More process control is one option to identify and improve influence factors to finally increase the overall reprocessing quality, best of all by different methods. ATP measurement seems to be a valid technique that allows an immediate repeat of the manual cleaning if the ATP results after manual cleaning exceed the established cutoff of 200 RLU.

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

    SciTech Connect

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

    1980-01-01

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

  16. Estimating reprocessing plant in-process inventories by simulation

    SciTech Connect

    Coulter, C.A.; Burr, T.L.; Hakkila, E.A.; Ai, H.; Kadokura, I.; Fujimaki, K.

    1995-09-01

    The Safeguards Systems Group`s generic simulation program FacSim was used to model the operation of the proposed Rokkasho Reprocessing Plant during an operating cycle consisting of a start-up phase, a period of steady-state operation, and a flush-out phase. The simulation results give a detailed account of nuclear material inventories in various process vessels as a function of time. As expected, it is found that the pulsed columns and the concentrator determine the rate at which the system responds to feed variations and transients; but the in-process inventory is dominated by the contents of the concentrator and tanks, and particularly by the contents of the tanks downstream from the concentrator. The results of the simulation were used for statistical studies of diversion detection, as described elsewhere in the Proceedings.

  17. Asbestos hazard in the reprocessed textile industry

    SciTech Connect

    Quinn, M.M.; Kriebel, D.; Buiatti, E.; Paci, E.; Sini, S.; Vannucchi, G.; Zappa, M.

    1987-01-01

    Epidemiologic studies have identified an excess risk of lung cancer and mesothelioma among workers in the reprocessed textile industry in Prato, Italy. These studies suggested that there may have been asbestos hazard in this industry although exposure was not known to exist. An industrial hygiene investigation was conducted to determine whether there was previous or current asbestos exposure in the industry. Walk-through surveys, environmental sampling, process documentation, and management and worker interviews were conducted in 13 textile reprocessing establishments. Polypropylene bags that once contained asbestos were found in 2 of the 13. Asbestos bags were cut open and used to cover bales of rags which were then distributed throughout the world. Workers were exposed to asbestos while handling the bags which were contaminated with chrysotile, amosite, and crocidolite. Additional sources of asbestos exposure that may have existed in the past in the industry are also discussed.

  18. Reprocessing technology development for irradiated beryllium

    SciTech Connect

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

    1995-09-01

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

  19. Noble gas atmospheric monitoring at reprocessing facilities

    SciTech Connect

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

    1997-05-01

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

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

    SciTech Connect

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

    1987-12-31

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

  1. Spent fuel and residue measurement instrumentation at the Sellafield nuclear fuel reprocessing facility

    SciTech Connect

    Chesterman, A.S.; Clark, P.A.

    1995-12-31

    The Sellafield reprocessing plant receives and reprocesses several thousand tonnes of spent light water reactor (LWR), advanced gas cooled reactor (AGR) and natural uranium magnesium alloy clad (Magnox) fuels each year. The safety and cost effectiveness of these operations has been supported by the development and installation, at key points in the process, of a range of special purpose radiometric instrumentation. Systems in routine operational use verify the cooling time, burn-tip and initial and final U-235 equivalent enrichment of fuel assemblies in the storage and handling ponds. Other systems determine the radionuclide inventories of fuel residues in intermediate level waste arising from plant operations. The measurement techniques employed include high resolution gamma spectrometry, passive neutron counting and neutron interrogation by the use of a Cf-252 source and deuterium-tritium (D-T) pulsed neutron generators. Details of the instruments including mechanical installation arrangements and measurement data are presented in the paper along with a discussion of possible future uses of similar instruments for burn-up credit associated with fuel and residue storage, transportation and disposal.

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

    SciTech Connect

    Cipiti, Benjamin B.; Smith, James D.

    2008-09-01

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

  3. Effect of Reprocessing and Accelerated Weathering on Impact-Modified Recycled Blend

    NASA Astrophysics Data System (ADS)

    Ramesh, V.; Mohanty, Smita; Biswal, Manoranjan; Nayak, Sanjay K.

    2015-12-01

    Recovery of recycled polycarbonate, acrylonitrile butadiene styrene, high-impact polystyrene, and its blends from waste electrical and electronic equipment plastics products properties were enhanced by the addition of virgin polycarbonate and impact modifier. The optimized blend formulation was processed through five cycles, at processing temperature, 220-240 °C and accelerated weathering up to 700 h. Moreover, the effect of reprocessing and accelerated weathering in the physical properties of the modified blends was investigated by mechanical, thermal, rheological, and morphological studies. The results show that in each reprocessing cycle, the tensile strength and impact strength decreased significantly and the similar behavior has been observed from accelerated weathering. Subsequently, the viscosity decreases and this decrease becomes the effect of thermal and photo-oxidative degradation. This can be correlated with FTIR analysis.

  4. Advances in Endoscope Reprocessing Technology and Its Impact on Pathogen Transmission.

    PubMed

    Shellnutt, Cathleen

    Pathogen transmissions via flexible endoscopes have been documented in the literature and have been historically related to human error or omission of steps in the reprocessing cycle. The 2008 Centers for Disease Control and Prevention report challenged manufacturers of automated endoscope reprocessors to improve and advance technology to automate more of the reprocessing steps. A review and synthesis of the literature following the 2008 Centers for Disease Control and Prevention report was performed to evaluate whether advances in reprocessing technology have occurred and whether these have had an impact on pathogen transmission via flexible endoscopes. The Iowa Model of Evidence-Based Practice to Promote Quality Care was used to guide the project. The literature search regarding pathogen transmission related to flexible endoscopes yielded 10 documents citing infections from 2008 to 2015. A total of 353 patients were identified as having been infected with a contaminated gastroscope, bronchoscope, or duodenoscope. An evaluation of reprocessing technology identified 3 automated endoscope reprocessors with enhanced capabilities and flushing devices intended to automate portions of the manual cleaning step.

  5. Improvement of technology for reprocessing of low-level wastes with the use of ozone

    SciTech Connect

    Revenko, Y.A.; Manakov, S.A.; Petrov, A.I.

    1995-12-31

    An original flowsheet is proposed for reprocessing of low-level wastes (LLW) containing surfactants. The flowsheet involves the use of ozone for destruction of surfactants and clinoptilolite for purification from ozonolysis products. Testing of the process in a pilot facility has shown the possibility for reduction of radioactive slurries by a factor of 10--15 with increasing performance of one filter-cycle by a factor of 2--3.

  6. Mechanical design of a light water breeder reactor

    DOEpatents

    Fauth, Jr., William L.; Jones, Daniel S.; Kolsun, George J.; Erbes, John G.; Brennan, John J.; Weissburg, James A.; Sharbaugh, John E.

    1976-01-01

    In a light water reactor system using the thorium-232 -- uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements.

  7. RADIOACTIVE WASTE STREAMS FROM VARIOUS POTENTIAL NUCLEAR FUEL CYCLE OPTIONS

    SciTech Connect

    Nick Soelberg; Steve Piet

    2010-11-01

    Five fuel cycle options, about which little is known compared to more commonly known options, have been studied in the past year for the United States Department of Energy. These fuel cycle options, and their features relative to uranium-fueled light water reactor (LWR)-based fuel cycles, include: • Advanced once-through reactor concepts (Advanced Once-Through, or AOT) – intended for high uranium utilization and long reactor operating life, use depleted uranium in some cases, and avoid or minimize used fuel reprocessing • Fission-fusion hybrid (FFH) reactor concepts – potential variations are intended for high uranium or thorium utilization, produce fissile material for use in power generating reactors, or transmute transuranic (TRU) and some radioactive fission product (FP) isotopes • High temperature gas reactor (HTGR) concepts - intended for high uranium utilization, high reactor thermal efficiencies; they have unique fuel designs • Molten salt reactor (MSR) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, use on-line reprocessing of the used fuel, produce lesser amounts of long-lived, highly radiotoxic TRU elements, and avoid fuel assembly fabrication • Thorium/U-233 fueled LWR (Th/U-233) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, and produce lesser amounts of long-lived, highly radiotoxic TRU elements. These fuel cycle options could result in widely different types and amounts of used or spent fuels, spent reactor core materials, and waste streams from used fuel reprocessing, such as: • Highly radioactive, high-burnup used metal, oxide, or inert matrix U and/or Th fuels, clad in Zr, steel, or composite non-metal cladding or coatings • Spent radioactive-contaminated graphite, SiC, carbon-carbon-composite, metal, and Be reactor core materials • Li-Be-F salts containing U, TRU, Th, and fission products • Ranges of separated or un-separated activation

  8. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    NASA Astrophysics Data System (ADS)

    Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-01

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  9. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    SciTech Connect

    Bahri, Che Nor Aniza Che Zainul Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  10. Vitrification of IFR and MSBR halide salt reprocessing wastes

    SciTech Connect

    Siemer, D.D.

    2013-07-01

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

  11. Corrosion and Corrosion Control in Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Gordon, Barry M.

    2013-08-01

    Serious corrosion problems have plagued the light water reactor (LWR) industry for decades. The complex corrosion mechanisms involved and the development of practical engineering solutions for their mitigation will be discussed in this article. After a brief overview of the basic designs of the boiling water reactor (BWR) and pressurized water reactor (PWR), emphasis will be placed on the general corrosion of LWR containments, flow-accelerated corrosion of carbon steel components, intergranular stress corrosion cracking (IGSCC) in BWRs, primary water stress corrosion cracking (PWSCC) in PWRs, and irradiation-assisted stress corrosion cracking (IASCC) in both systems. Finally, the corrosion future of both plants will be discussed as plants extend their period of operation for an additional 20 to 40 years.

  12. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    SciTech Connect

    Lewis, M R

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  13. Non-linear analysis in Light Water Reactor design

    SciTech Connect

    Rashid, Y.R.; Sharabi, M.N.; Nickell, R.E.; Esztergar, E.P.; Jones, J.W.

    1980-03-01

    The results obtained from a scoping study sponsored by the US Department of Energy (DOE) under the Light Water Reactor (LWR) Safety Technology Program at Sandia National Laboratories are presented. Basically, this project calls for the examination of the hypothesis that the use of nonlinear analysis methods in the design of LWR systems and components of interest include such items as: the reactor vessel, vessel internals, nozzles and penetrations, component support structures, and containment structures. Piping systems are excluded because they are being addressed by a separate study. Essentially, the findings were that nonlinear analysis methods are beneficial to LWR design from a technical point of view. However, the costs needed to implement these methods are the roadblock to readily adopting them. In this sense, a cost-benefit type of analysis must be made on the various topics identified by these studies and priorities must be established. This document is the complete report by ANATECH International Corporation.

  14. Light-water breeder reactor (LWBR Development Program)

    DOEpatents

    Beaudoin, B.R.; Cohen, J.D.; Jones, D.H.; Marier, L.J. Jr.; Raab, H.F.

    1972-06-20

    Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

  15. Materials Inventory Database for the Light Water Reactor Sustainability Program

    SciTech Connect

    Kazi Ahmed; Shannon M. Bragg-Sitton

    2013-08-01

    Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

  16. The Consortium for Advanced Simulation of Light Water Reactors

    SciTech Connect

    Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

    2011-10-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  17. 76 FR 24495 - Reprocessing of Reusable Medical Devices; Public Workshop

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-02

    ... Medical Devices Workshop.'' The purpose of the workshop is to discuss factors affecting the reprocessing... devices with the overall goal to reduce the risk of infection. The topics to be discussed are: Factors... reprocessing may have been a contributing factor in microbial transmission and subsequent infection....

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

    SciTech Connect

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

    2013-07-01

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

  19. Methods of reprocessing complex medical equipment.

    PubMed

    Babb, J R

    1988-02-01

    The choice as to which of the two gaseous processes is best suited to individual hospital needs is a difficult one. Very few items are unable to tolerate 73 degrees C (LTSF) and these few can withstand 37 degrees C or 55 degrees C (EO). Unfortunately, LTSF is a 'moist' process and sterilizers have a poor history of providing sterilization without modification, and consequently few are used. Ethylene oxide is more reliable, but environmental hazards are greater and running costs high. Both processes are time-consuming and the use of sporicidal disinfectants such as glutaraldehyde is often the only practical alternative. Before purchasing any gaseous sterilizer it is essential to consider throughput and the availability of alternative processes. It may prove sensible to share facilities or at least offer a regional facility. It is certainly not worthwhile purchasing expensive gas sterilizers for reprocessing inexpensive single-use items or for those that require disinfection only. Low temperature steam is safe, inexpensive and no special environmental provisions are necessary. It is, however, not a sterilization process. Disinfectants, hot water and steam will continue to be the only suitable methods for reprocessing items outside the hospital sterile supply department or disinfection unit. Concern over the decontamination of blood-stained instruments following use on patients with hepatitis B or HIV has led to an upsurge of interest in boilers and inexpensive bench top ovens and autoclaves. Such processes are likely to prove more effective than disinfectants but should heat treatment prove impractical then 2% glutaraldehyde or 70% alcohol may be used.(ABSTRACT TRUNCATED AT 250 WORDS)

  20. Reprocessing of research reactor fuel the Dounreay option

    SciTech Connect

    Cartwright, P.

    1997-08-01

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

  1. Fatigue and environmentally assisted cracking in light water reactors

    SciTech Connect

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

    1991-12-01

    Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas.

  2. Technologies for Upgrading Light Water Reactor Outlet Temperature

    SciTech Connect

    Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

    2013-07-01

    Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessment of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.

  3. Thermal-Hydraulic Analysis of Supercritical Pressure Light Water Reactors

    SciTech Connect

    Cheng, X.; Schulenberg, T.; Koshizuka, S.; Oka, Y.; Souyri, A.

    2002-07-01

    In the frame of the European project HPLWR, joined by European research institutions, industrial partners and the University of Tokyo, thermal-hydraulic analysis of supercritical pressure light water reactors has been carried out. A thorough literature survey on heat transfer of supercritical fluids indicates a large deficiency in the prediction of the heat transfer coefficient and the onset of heat transfer deterioration under the reactor condition. A CFD code for analysing the thermal-hydraulic behaviour of supercritical fluids was developed. Numerical results show that the heat transfer coefficient, including the heat transfer deterioration region, can be well predicted using this CFD code, at least for circular tube geometries. Such a CFD code is well suitable for understanding the heat transfer mechanism. Based on the numerical results, a new heat transfer correlation has been proposed. For the thermal-hydraulic design of an HPLWR fuel assembly, the subchannel analysis code STAR-SC has been developed with a high numerical efficiency and a high applicability to different kinds of fuel assembly configurations. The results show clearly that design of a HPLWR fuel assembly is a highly challenging task. At the same time, sub-channel analysis provides some important guidelines for the design of a HPLWR fuel assembly. (authors)

  4. Impact of inflow transport approximation on light water reactor analysis

    NASA Astrophysics Data System (ADS)

    Choi, Sooyoung; Smith, Kord; Lee, Hyun Chul; Lee, Deokjung

    2015-10-01

    The impact of the inflow transport approximation on light water reactor analysis is investigated, and it is verified that the inflow transport approximation significantly improves the accuracy of the transport and transport/diffusion solutions. A methodology for an inflow transport approximation is implemented in order to generate an accurate transport cross section. The inflow transport approximation is compared to the conventional methods, which are the consistent-PN and the outflow transport approximations. The three transport approximations are implemented in the lattice physics code STREAM, and verification is performed for various verification problems in order to investigate their effects and accuracy. From the verification, it is noted that the consistent-PN and the outflow transport approximations cause significant error in calculating the eigenvalue and the power distribution. The inflow transport approximation shows very accurate and precise results for the verification problems. The inflow transport approximation shows significant improvements not only for the high leakage problem but also for practical large core problem analyses.

  5. Burnup credit validation of SCALE-4 using light water reactor criticals

    SciTech Connect

    Bowman, S.M.; Hermann, O.W. ); Brady, M.C. )

    1993-01-01

    The ANSI/ANS 8.1 criticality safety standard recommends validation and benchmarking of the calculational methods used in evaluating criticality safety limits for away-from-reactor applications. The lack of critical experiments with burned light-water-reactor (LWR) fuel in racks or in casks necessitates the validation of burnup credit methods by comparison to LWR core criticals. These are relevant benchmarks because they test a methodology's ability to predict spent fuel isotopics and to evaluate the reactivity effects of heterogeneities and strong absorbers. Data are available to perform analyses at precise state points. The US Department of Energy Burnup Credit Program has sponsored analysis of selected reactor core critical configurations from commercial pressurized-water-reactors (PWRs) in order to validate an appropriate analysis methodology. The initial methodology used the SCALE-4 code system to analyze a set of reactor critical configurations from Virginia Power's Surry and North Anna reactors. The methodology has since been revised to simplify both the data requirements and the calculational procedure for the criticality analyst. This revised methodology is validated here by comparison to three reactor critical configurations from Tennessee Valley Authority's Sequoyah Unit 2 Cycle 3 and two from Virginia Power's Surry Unit 1 Cycle 2.

  6. Burnup credit validation of SCALE-4 using light-water-reactor criticals

    SciTech Connect

    Bowman, S.M.; Hermann, O.W.; Brady, M.C.

    1993-03-01

    The ANSI/ANS 8.1 criticality safety standard recommends validation and benchmarking of the calculational methods used in evaluating criticality safety limits for away-from-reactor applications. The lack of critical experiments with burned light-water reactor (LWR) fuel in racks or in casks necessitates the validation of burnup credit methods by comparison with LWR core criticals. These benchmarks are relevant because they test a methodology`s ability to predict spent fuel isotopic and to evaluate the reactivity effects of heterogeneities and strong absorbers. Data are available to perform analyses at precise state points. As part of the Burnup Credit Analysis Verification (BCAV) Task, the U.S. Department of Energy has sponsored analysis of selected reactor core critical configurations from commercial pressurized-water-reactors (PWRs). The initial analysis methodology used the SCALE-4 code system to analyze a set of reactor critical configurations from Virginia Power`s Slurry and North Anna reactors. However, the analysis procedure was complex and included the calculation of lumped fission products. The methodology has since been revised to simplify both the data requirements and the calculational procedure for the criticality analyst. This revised methodology is validated here by a comparison with three reactor critical configurations from Tennessee Valley Authority`s Sequoyah Unit 2 Cycle 3 and two from Virginia Power`s Slurry Unit 1 Cycle 2.

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

    SciTech Connect

    Ehinger, M.H.

    1989-01-01

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

  8. Lessons Learned in International Safeguards - Implementation of Safeguards at the Rokkasho Reprocessing Plant

    SciTech Connect

    Ehinger, Michael H; Johnson, Shirley

    2010-02-01

    The focus of this report is lessons learned at the Rokkasho Reprocessing Plant (RRP). However, the subject of lessons learned for application of international safeguards at reprocessing plants includes a cumulative history of inspections starting at the West Valley (New York, U.S.A.) reprocessing plant in 1969 and proceeding through all of the efforts over the years. The RRP is the latest and most challenging application the International Atomic Energy Agency has faced. In many ways the challenges have remained the same, timely inspection and evaluation with limited inspector resources, with the continuing realization that planning and preparations can never start early enough in the life cycle of a facility. Lessons learned over the years have involved the challenges of using ongoing advances in technology and dealing with facilities with increased throughput and continuous operation. This report will begin with a review of historical developments and lessons learned. This will provide a basis for a discussion of the experiences and lessons learned from the implementation of international safeguards at RRP.

  9. Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

    SciTech Connect

    Snead, Lance Lewis; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Robb, Kevin R.; Snead, Mary A.

    2015-04-01

    This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

  10. Environmental Impacts on Nuclear Reprocessing Solvents

    NASA Astrophysics Data System (ADS)

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

    2009-12-01

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

  11. High quality reprocessed GPS Zenith Total Delay dataset over Europe

    NASA Astrophysics Data System (ADS)

    Pacione, Rosa; Pace, Brigida; Bianco, Giuseppe

    2015-04-01

    The present availability of 18 years of GPS data belonging to the European Permanent Network (EPN, http://www.epncb.oma.be/) is a valuable database for the development of a climate data record of GPS tropospheric products. We homogeneously reprocessed the whole EPN network for the period 1996-2013 in a consistent way using GIPSY-OASIS II software and applying the state-of-the-art models. This ongoing reprocessing effort, part of the EPN Repro2 initiative, will provide a GPS climate data record over Europe with high potential for monitoring trend and variability in atmospheric water vapour thus improving the knowledge of climatic trends of atmospheric water vapour, being useful for global and regional NWP reanalyses and climate model simulations. These reprocessed ZTD time series will be evaluated against radiosonde data as well as independently reprocessed GPS ZTD time-series.

  12. Revised accident source terms for light-water reactors

    SciTech Connect

    Soffer, L.

    1995-02-01

    This paper presents revised accident source terms for light-water reactors incorporating the severe accident research insights gained in this area over the last 15 years. Current LWR reactor accident source terms used for licensing date from 1962 and are contained in Regulatory Guides 1.3 and 1.4. These specify that 100% of the core inventory of noble gases and 25% of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental iodine. These assumptions have strongly affected present nuclear air cleaning requirements by emphasizing rapid actuation of spray systems and filtration systems optimized to retain elemental iodine. A proposed revision of reactor accident source terms and some im implications for nuclear air cleaning requirements was presented at the 22nd DOE/NRC Nuclear Air Cleaning Conference. A draft report was issued by the NRC for comment in July 1992. Extensive comments were received, with the most significant comments involving (a) release fractions for both volatile and non-volatile species in the early in-vessel release phase, (b) gap release fractions of the noble gases, iodine and cesium, and (c) the timing and duration for the release phases. The final source term report is expected to be issued in late 1994. Although the revised source terms are intended primarily for future plants, current nuclear power plants may request use of revised accident source term insights as well in licensing. This paper emphasizes additional information obtained since the 22nd Conference, including studies on fission product removal mechanisms, results obtained from improved severe accident code calculations and resolution of major comments, and their impact upon the revised accident source terms. Revised accident source terms for both BWRS and PWRS are presented.

  13. Cost and availability of gadolinium for nuclear fuel reprocessing plants

    SciTech Connect

    Klepper, O.H.

    1985-06-01

    Gadolinium is currently planned for use as a soluble neutron poison in nuclear fuel reprocessing plants to prevent criticality of solutions of spent fuel. Gadolinium is relatively rare and expensive. The present study was undertaken therefore to estimate whether this material is likely to be available in quantities sufficient for fuel reprocessing and at reasonable prices. It was found that gadolinium, one of 16 rare earth elements, appears in the marketplace as a by-product and that its present supply is a function of the production rate of other more prevalent rare earths. The potential demand for gadolinium in a fuel reprocessing facility serving a future fast reactor industry amounts to only a small fraction of the supply. At the present rate of consumption, domestic supplies of rare earths containing gadolinium are adequate to meet national needs (including fuel reprocessing) for over 100 years. With access to foreign sources, US demands can be met well beyond the 21st century. It is concluded therefore that the supply of gadolinium will quite likely be more than adequate for reprocessing spent fuel for the early generation of fast reactors. The current price of 99.99% pure gadolinium oxide lies in the range $50/lb to $65/lb (1984 dollars). By the year 2020, in time for reprocessing spent fuel from an early generation of large fast reactors, the corresponding values are expected to lie in the $60/lb to $75/lb (1984 dollars) price range. This increase is modest and its economic impact on nuclear fuel reprocessing would be minor. The economic potential for recovering gadolinium from the wastes of nuclear fuel reprocessing plants (which use gadolinium neutron poison) was also investigated. The cost of recycled gadolinium was estimated at over twelve times the cost of fresh gadolinium, and thus recycle using current recovery technology is not economical. 15 refs., 4 figs., 11 tabs.

  14. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    SciTech Connect

    Thomas, Kenneth; Oxstrand, Johanna

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

  15. Visible light water splitting using dye-sensitized oxide semiconductors.

    PubMed

    Youngblood, W Justin; Lee, Seung-Hyun Anna; Maeda, Kazuhiko; Mallouk, Thomas E

    2009-12-21

    Researchers are intensively investigating photochemical water splitting as a means of converting solar to chemical energy in the form of fuels. Hydrogen is a key solar fuel because it can be used directly in combustion engines or fuel cells, or combined catalytically with CO(2) to make carbon containing fuels. Different approaches to solar water splitting include semiconductor particles as photocatalysts and photoelectrodes, molecular donor-acceptor systems linked to catalysts for hydrogen and oxygen evolution, and photovoltaic cells coupled directly or indirectly to electrocatalysts. Despite several decades of research, solar hydrogen generation is efficient only in systems that use expensive photovoltaic cells to power water electrolysis. Direct photocatalytic water splitting is a challenging problem because the reaction is thermodynamically uphill. Light absorption results in the formation of energetic charge-separated states in both molecular donor-acceptor systems and semiconductor particles. Unfortunately, energetically favorable charge recombination reactions tend to be much faster than the slow multielectron processes of water oxidation and reduction. Consequently, visible light water splitting has only recently been achieved in semiconductor-based photocatalytic systems and remains an inefficient process. This Account describes our approach to two problems in solar water splitting: the organization of molecules into assemblies that promote long-lived charge separation, and catalysis of the electrolysis reactions, in particular the four-electron oxidation of water. The building blocks of our artificial photosynthetic systems are wide band gap semiconductor particles, photosensitizer and electron relay molecules, and nanoparticle catalysts. We intercalate layered metal oxide semiconductors with metal nanoparticles. These intercalation compounds, when sensitized with [Ru(bpy)(3)](2+) derivatives, catalyze the photoproduction of hydrogen from sacrificial

  16. Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J.

    1995-09-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289{degrees}C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  17. Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

    SciTech Connect

    Chung, H.M.; Chopra, O.K.; Gavenda, D.J.; Hins, A.G.; Kassner, T.F.; Ruther, W.E.; Shack, W.J.; Soppet, W.K.

    1996-01-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289{degrees}C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  18. Environmentally assisted cracking in light water reactors. Semiannual progress report, January 1996--June 1996

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1997-05-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1996 to June 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288{degrees}C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in air and high-purity, low-DO water. 83 refs., 60 figs., 14 tabs.

  19. Environmentally assisted cracking in light water reactors. Semiannual report July 1996--December 1996

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J.

    1997-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289 C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.

  20. Waste Minimization Study on Pyrochemical Reprocessing Processes

    SciTech Connect

    Boussier, H.; Conocar, O.; Lacquement, J.

    2006-07-01

    Ideally a new pyro-process should not generate more waste, and should be at least as safe and cost effective as the hydrometallurgical processes currently implemented at industrial scale. This paper describes the thought process, the methodology and some results obtained by process integration studies to devise potential pyro-processes and to assess their capability of achieving this challenging objective. As example the assessment of a process based on salt/metal reductive extraction, designed for the reprocessing of Generation IV carbide spent fuels, is developed. Salt/metal reductive extraction uses the capability of some metals, aluminum in this case, to selectively reduce actinide fluorides previously dissolved in a fluoride salt bath. The reduced actinides enter the metal phase from which they are subsequently recovered; the fission products remain in the salt phase. In fact, the process is not so simple, as it requires upstream and downstream subsidiary steps. All these process steps generate secondary waste flows representing sources of actinide leakage and/or FP discharge. In aqueous processes the main solvent (nitric acid solution) has a low boiling point and evaporate easily or can be removed by distillation, thereby leaving limited flow containing the dissolved substance behind to be incorporated in a confinement matrix. From the point of view of waste generation, one main handicap of molten salt processes, is that the saline phase (fluoride in our case) used as solvent is of same nature than the solutes (radionuclides fluorides) and has a quite high boiling point. So it is not so easy, than it is with aqueous solutions, to separate solvent and solutes in order to confine only radioactive material and limit the final waste flows. Starting from the initial block diagram devised two years ago, the paper shows how process integration studies were able to propose process fittings which lead to a reduction of the waste variety and flows leading at an 'ideal

  1. Disposable Bronchoscope Model for Simulating Endoscopic Reprocessing and Surveillance Cultures.

    PubMed

    Yassin, Mohamed H; Hariri, Rahman; Hamad, Yasir; Ferrelli, Juliet; McKibben, Leeanna; Doi, Yohei

    2017-02-01

    BACKGROUND Endoscope-associated infections are reported despite following proper reprocessing methods. Microbiological testing can confirm the adequacy of endoscope reprocessing. Multiple controversies related to the method and interpretation of microbiological testing cultures have arisen that make their routine performance a complex target. OBJECTIVE We conducted a pilot study using disposable bronchoscopes (DBs) to simulate different reprocessing times and soaking times and to compare high-level disinfection versus ethylene oxide sterilization. We also reviewed the time to reprocessing and duration of the procedures. METHODS Bronchoscopes were chosen because an alternative disposable scope is commercially available and because bronchoscopes are more prone to delays in processing. Disposable bronchoscopes were contaminated using a liquid bacterial suspension and were then incubated for 1-4 hours. Standard processing and high-level disinfection were performed on 36 endoscopes. Ethylene oxide sterilization was performed on 21 endoscopes. Endoscope cultures were performed using the standard "brush, flush, brush" technique. RESULTS After brushing was performed, a final water-flush culture procedure was the most effective method of detecting bacterial persistence on the disposable scopes. Klebsiella pneumoniae was the most commonly recovered organism after reprocessing. Ethylene oxide sterilization did not result in total elimination of viable bacteria. CONCLUSION Routine endoscopy cultures may be required to assess the adequacy of endoscopic processing. Infect Control Hosp Epidemiol 2017;38:136-142.

  2. 77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-12

    ... Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY: Nuclear Regulatory Commission... Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES: Submit comments by November...

  3. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power...

  4. 78 FR 41436 - Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-10

    ... COMMISSION Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The NRC seeks public...- Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' This area includes a revised...

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

    SciTech Connect

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

    2010-09-01

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

  6. Equipment specifications for an electrochemical fuel reprocessing plant

    SciTech Connect

    Hemphill, Kevin P

    2010-01-01

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

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

    SciTech Connect

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

    1990-01-01

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

  8. A Review of Current Disinfectants for Gastrointestinal Endoscopic Reprocessing

    PubMed Central

    Park, Sanghoon; Koo, Ja Seol; Park, Jeong Bae; Lim, Yun Jeong; Hong, Su Jin; Kim, Sang-Woo; Chun, Hoon Jai

    2013-01-01

    Gastrointestinal endoscopy is gaining popularity for diagnostic and therapeutic purposes. However, concerns over endoscope-related nosocomial infections are increasing, together with interest by the general public in safe and efficient endoscopy. For this reason, reprocessing the gastrointestinal endoscope is an important step for effective performance of endoscopy. Disinfectants are essential to the endoscope reprocessing procedure. Before selecting an appropriate disinfectant, their characteristics, limitations and means of use must be fully understood. Herein, we review the characteristics of several currently available disinfectants, including their uses, potency, advantages, and disadvantages. Most disinfectants can be used to reprocess gastrointestinal endoscopes if the manufacturer's guidelines are followed. The selection and use of a suitable disinfectant depends on the individual circumstances of each endoscopy suite. PMID:23964330

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

    SciTech Connect

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

    1998-10-02

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

  10. Do the Kepler AGN light curves need reprocessing?

    NASA Astrophysics Data System (ADS)

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

    2015-10-01

    We gauge the impact of spacecraft-induced effects on the inferred variability properties of the light curve of the Seyfert 1 AGN Zw 229-15 observed by Kepler. We compare the light curve of Zw 229-15 obtained from the Kepler MAST data base with a reprocessed light curve constructed from raw pixel data. We use the first-order structure function, SF(δt), to fit both light curves to the damped power-law PSD (power spectral density) of Kasliwal et al. On short time-scales, we find a steeper log PSD slope (γ = 2.90 to within 10 per cent) for the reprocessed light curve as compared to the light curve found on MAST (γ = 2.65 to within 10 per cent) - both inconsistent with a damped random walk (DRW) which requires γ = 2. The log PSD slope inferred for the reprocessed light curve is consistent with previous results that study the same reprocessed light curve. The turnover time-scale is almost identical for both light curves (27.1 and 27.5 d for the reprocessed and MAST data base light curves). Based on the obvious visual difference between the two versions of the light curve and on the PSD model fits, we conclude that there remain significant levels of spacecraft-induced effects in the standard pipeline reduction of the Kepler data. Reprocessing the light curves will change the model inferenced from the data but is unlikely to change the overall scientific conclusions reached by Kasliwal et al. - not all AGN light curves are consistent with the DRW.

  11. Cleaning of instruments: an absolute requirement for successful reprocessing.

    PubMed

    Wilder, Jonathan A; Roth, Klaus

    2012-01-01

    Cleaning of medical devices, and the validation and verification of this cleaning, has advanced greatly, even in the last five years. It is recognized as an essential part of the reprocessing of reusable medical devices, and is becoming more and more consistent as device manufacturers and the FDA give cleaning more consideration when a device is brought to market. Scrupulous care must be taken to ensure that the cleaning approach for each device is appropriate (but similar items can be reprocessed in a similar manner). There is no replacement for excellent training and maintainance of equipment, but available tests can help ensure that failures are caught earlier rather than later.

  12. Head-end reprocessing equipment remote maintenance demonstration

    SciTech Connect

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

    1989-01-01

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

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

    SciTech Connect

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

    2007-09-01

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

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

    SciTech Connect

    Not Available

    1983-08-01

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

  15. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14

    SciTech Connect

    Schneider, K.J.

    1982-09-01

    The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation.

  16. Acoustic Emission and Guided Ultrasonic Waves for Detection and Continuous Monitoring of Cracks in Light Water Reactor Components

    SciTech Connect

    Meyer, Ryan M.; Coble, Jamie B.; Ramuhalli, Pradeep; Watson, Bruce E.; Cumblidge, Stephen E.; Doctor, Steven R.; Bond, Leonard J.

    2012-06-28

    Acoustic emission (AE) and guided ultrasonic waves (GUW) are considered for continuous monitoring and detection of cracks in Light Water Reactor (LWR) components. In this effort, both techniques are applied to the detection and monitoring of fatigue crack growth in a full scale pipe component. AE results indicated crack initiation and rapid growth in the pipe, and significant GUW responses were observed in response to the growth of the fatigue crack. After initiation, the crack growth was detectable with AE for approximately 20,000 cycles. Signals associated with initiation and rapid growth where distinguished based on total rate of activity and differences observed in the centroid frequency of hits. An intermediate stage between initiation and rapid growth was associated with significant energy emissions, though few hits. GUW exhibit a nearly monotonic trend with crack length with an exception of measurements obtained at 41 mm and 46 mm.

  17. Comparative cost-efficiency of the EVOTECH endoscope cleaner and reprocessor versus manual cleaning plus automated endoscope reprocessing in a real-world Canadian hospital endoscopy setting

    PubMed Central

    2011-01-01

    Background Reprocessing of endoscopes generally requires labour-intensive manual cleaning followed by high-level disinfection in an automated endoscope reprocessor (AER). EVOTECH Endoscope Cleaner and Reprocessor (ECR) is approved for fully automated cleaning and disinfection whereas AERs require manual cleaning prior to the high-level disinfection procedure. The purpose of this economic evaluation was to determine the cost-efficiency of the ECR versus AER methods of endoscopy reprocessing in an actual practice setting. Methods A time and motion study was conducted at a Canadian hospital to collect data on the personnel resources and consumable supplies costs associated with the use of EVOTECH ECR versus manual cleaning followed by AER with Medivators DSD-201. Reprocessing of all endoscopes was observed and timed for both reprocessor types over three days. Laboratory staff members were interviewed regarding the consumption and cost of all disposable supplies and equipment. Exact Wilcoxon rank sum test was used for assessing differences in total cycle reprocessing time. Results Endoscope reprocessing was significantly shorter with the ECR than with manual cleaning followed by AER. The differences in median time were 12.46 minutes per colonoscope (p < 0.0001), 6.31 minutes per gastroscope (p < 0.0001), and 5.66 minutes per bronchoscope (p = 0.0040). Almost 2 hours of direct labour time was saved daily with the ECR. The total per cycle cost of consumables and labour for maintenance was slightly higher for EVOTECH ECR versus manual cleaning followed by AER ($8.91 versus $8.31, respectively). Including the cost of direct labour time consumed in reprocessing scopes, the per cycle and annual costs of using the EVOTECH ECR was less than the cost of manual cleaning followed by AER disinfection ($11.50 versus $11.88). Conclusions The EVOTECH ECR was more efficient and less costly to use for the reprocessing of endoscopes than manual cleaning followed by AER disinfection

  18. Metrics for the technical performance evaluation of light water reactor accident-tolerant fuel

    SciTech Connect

    Bragg-Sitton, Shannon M.; Todosow, Michael; Montgomery, Robert; Stanek, Christopher R.; Carmack, W. Jon; Montgomery, Rose

    2016-07-11

    The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors (LWRs) became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident-tolerant fuel (ATF) for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+), fuels with enhanced accident tolerance would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance. Research and development of ATF in the United States is conducted under the U.S. Department of Energy (DOE) Fuel Cycle Research and Development Advanced Fuels Campaign. The DOE is sponsoring multiple teams to develop ATF concepts within multiple national laboratories, universities, and the nuclear industry. Concepts under investigation offer both evolutionary and revolutionary changes to the current nuclear fuel system. This study summarizes the technical evaluation methodology proposed in the United States to aid in the optimization and prioritization of candidate ATF designs.

  19. Metrics for the technical performance evaluation of light water reactor accident-tolerant fuel

    DOE PAGES

    Bragg-Sitton, Shannon M.; Todosow, Michael; Montgomery, Robert; ...

    2016-07-11

    The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors (LWRs) became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident-tolerant fuel (ATF) for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, andmore » economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+), fuels with enhanced accident tolerance would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance. Research and development of ATF in the United States is conducted under the U.S. Department of Energy (DOE) Fuel Cycle Research and Development Advanced Fuels Campaign. The DOE is sponsoring multiple teams to develop ATF concepts within multiple national laboratories, universities, and the nuclear industry. Concepts under investigation offer both evolutionary and revolutionary changes to the current nuclear fuel system. This study summarizes the technical evaluation methodology proposed in the United States to aid in the optimization and prioritization of candidate ATF

  20. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    DTIC Science & Technology

    2010-03-05

    with uranium to make mixed-oxide ( MOX ) fuel, in which the 239Pu largely substitutes for 235U. Two French reprocessing plants at La Hague can each...and France also have older plants to reprocess gas-cooled reactor fuel, and India has a 275-ton plant.53 About 200 metric tons of MOX fuel is used...to make MOX fuel for today’s nuclear power plants are modest. Existing commercial light water reactors use ordinary water to slow down, or “moderate

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

    SciTech Connect

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

    1991-01-01

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

  2. Eye Movement Desensitization and Reprocessing: A Critical Analysis.

    ERIC Educational Resources Information Center

    Erwin, Terry McVannel

    Since Shapiro's introduction of Eye Movement Desensitization and Reprocessing (EMDR) in 1989, it has been a highly controversial therapeutic technique. Critical reviews of Shapiro's initial study have highlighted many methodological shortcomings in her work. And early empirical research that followed Shapiro's original study has been criticized…

  3. Consolidated Fuel Reprocessing Program. Progress report for period, April 1-June 30, 1985

    SciTech Connect

    Not Available

    1985-08-01

    All research and development on civilian power reactor fuel reprocessing in the United States is managed under the Consolidated Fuel Reprocessing Program (CFRP) centered at Oak Ridge National Laboratory (ORNL). Technical progress is reported in overview fashion for the following: (1) process and engineering R and D; (2) engineering systems; (3) integrated equipment test facility operations; (4) strategic planning and analysis; (5) breeder reprocessing engineering test project; and HTGR fuel reprocessing.

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

    NASA Astrophysics Data System (ADS)

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

    2010-05-01

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

  5. Reprocessing VIIRS sensor data records from the early SNPP mission

    NASA Astrophysics Data System (ADS)

    Blonski, Slawomir; Cao, Changyong

    2016-10-01

    The Visible-Infrared Imaging Radiometer Suite (VIIRS) instrument onboard the Suomi National Polar-orbiting Partnership (SNPP) satellite began acquiring Earth observations in November 2011. VIIRS data from all spectral bands became available three months after launch when all infrared-band detectors were cooled down to operational temperature. Before that, VIIRS sensor data record (SDR) products were successfully generated for the visible and near infrared (VNIR) bands. Although VIIRS calibration has been significantly improved through the four years of the SNPP mission, SDR reprocessing for this early mission phase has yet to be performed. Despite a rapid decrease in the telescope throughput that occurred during the first few months on orbit, calibration coefficients for the VNIR bands were recently successfully generated using an automated procedure that is currently deployed in the operational SDR production system. The reanalyzed coefficients were derived from measurements collected during solar calibration events that occur on every SNPP orbit since the beginning of the mission. The new coefficients can be further used to reprocess the VIIRS SDR products. In this study, they are applied to reprocess VIIRS data acquired over pseudo-invariant calibration sites Libya 4 and Sudan 1 in Sahara between November 2011 and February 2012. Comparison of the reprocessed SDR products with the original ones demonstrates improvements in the VIIRS calibration provided by the reprocessing. Since SNPP is the first satellite in a series that will form the Joint Polar Satellite System (JPSS), calibration methods developed for the SNPP VIIRS will also apply to the future JPSS measurements.

  6. Fuel cycles for the 80's

    SciTech Connect

    Not Available

    1980-01-01

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

  7. Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

    SciTech Connect

    Loflin, Leonard; McRimmon, Beth

    2014-12-18

    This report summarizes a project by EPRI to include requirements for small modular light water reactors (smLWR) into the EPRI Utility Requirements Document (URD) for Advanced Light Water Reactors. The project was jointly funded by EPRI and the U.S. Department of Energy (DOE). The report covers the scope and content of the URD, the process used to revise the URD to include smLWR requirements, a summary of the major changes to the URD to include smLWR, and how to use the URD as revised to achieve value on new plant projects.

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

    SciTech Connect

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

    2007-07-01

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

  9. Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

    SciTech Connect

    Hursin, M.; Perret, G.

    2012-07-01

    The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

  10. Next-Generation Online MC&A Technologies for Reprocessing Plants

    SciTech Connect

    Smith, Leon E.; Schwantes, Jon M.; Ressler, Jennifer J.; Douglas, Matt; Anderson, Kevin K.; Fraga, Carlos G.; Durst, Casey; Orton, Chris; Christensen, Robert P.

    2007-08-03

    As power-production nuclear fuel cycles propagate across the globe, a new generation of measurement technologies is needed to support safeguards monitoring of fuel reprocessing facilities. This paper describes the simulation and analysis of two potential technologies for meeting the challenges of 1) direct measurement of fissile isotopic content in irradiated fuel to detect partial defects, and 2) near-real-time monitoring of process chemistry to detect protracted diversion scenarios. Lead slowing-down spectroscopy is the core of the spent fuel assay technology and multi-isotope indicators via high-resolution gamma-ray spectroscopy is the foundation of the process chemistry verification approach. The safeguards context and methods for each technology are described and the results of preliminary performance studies are presented. The quantitative results for both studies are promising but more comprehensive analysis and empirical validation is needed to adequately assess their potential value as next-generation online materials control and accountability measures.

  11. ATLAS level-1 calorimeter trigger: Monitoring and data reprocessing

    NASA Astrophysics Data System (ADS)

    Dimond, David; Hong, Tae; Carlson, Benjamin; Atlas Collaboration

    2017-01-01

    We present the monitoring and data reprocessing for the calorimeter-based hardware level-1 trigger system (L1Calo) for the ATLAS experiment. This trigger system was upgraded after the Run-1 data taking period (2009-2012) to prepare for Run-2 (2015-current), which allowed better control the event rates for algorithms based on jets and/or missing energy. Monitoring tools for the upgraded system is described. We also present a new offline tool to reprocess previous data samples with altered L1Calo settings, such as calibration constants and noise cuts. The samples are used to study the dependence of the event rates and signal efficiencies on the settings. The studies can help plan the appropriate L1Calo settings for upcoming data taking periods as well as for future runs.

  12. Principles of infection prevention and reprocessing in ENT endoscopy

    PubMed Central

    Kramer, Axel; Kohnen, Wolfgang; Israel, Susanne; Ryll, Sylvia; Hübner, Nils-Olaf; Luckhaupt, Horst; Hosemann, Werner

    2015-01-01

    This article gives an overview on the principles of reprocessing of rigid and flexible endoscopes used in ENT units including structural and spatial requirements based on general and ENT-specific risks of infection associated with diagnostic and therapeutic endoscopy. The underlying legal principles as well as recommendations from scientific societies will be exemplified in order to give a practical guidance to the otorhinolaryngologist. Preliminary results of a small nation-wide survey on infection control standards based on data of 29 ENT practices in Germany reveal current deficits of varying degree concerning infection control management including reprocessing of endoscopes. The presented review aims to give support to the establishment of a structured infection control management program including the evaluation of results by means of a prospective surveillance. PMID:26770284

  13. The Effect of Reprocessing on the Tensile Properties of Composites

    NASA Astrophysics Data System (ADS)

    Bodur, Mehmet Safa; Bakkal, Mustafa; Berkalp, Omer Berk; Sadikoglu, Telem Gok

    2011-01-01

    In this study, waste cotton fabric reinforced polymer matrix composite material has been manufactured by a custom made recycling extruder. Composites with different reinforcement ratios as 12,5%wt ( 12,5%wtRPE ) and 25%wt ( 25%wtRPE ) were tested for their mechanical properties such as tensile strength and young's modulus. The material was then granulated down to the size enough to be used in the extrusion process in order to observe the effects of reprocessing. Reprocessing leads to improve Tensile Strength of composite materials and slows down the reduction of tensile strength of polyethylene. It was observed that composite materials were highly affected by the fiber orientation and acts as anisotropic material under the load.

  14. On the Reprocessing and Reanalysis of Observations for Climate

    NASA Technical Reports Server (NTRS)

    Bosilovich, Michael G.; Kennedy, John; Dee, Dick; ONeill, Alan

    2012-01-01

    The long observational record is critical to our understanding of the Earth s climate, but most observing systems were not developed with a climate objective in mind. As a result, tremendous efforts have gone into assessing and reprocessing the data records to improve their usefulness in climate studies. Many challenges remain, such as tracking the improvement of processing algorithms and limited spatial coverage. Reanalyses have fostered significant research, yet reliable global trends in many physical fields are not yet attainable, despite significant advances in data assimilation and numerical modeling. Communication of the strengths, limitations and uncertainties of reprocessed observations and reanalysis data, not only among the community of developers, but also with the extended research community, including the new generations of researchers and the decision makers is crucial for further advancement of the observational data records. WCRP provides the means to bridge the different motivating objectives on which national efforts focus.

  15. LIFE vs. LWR: End of the Fuel Cycle

    SciTech Connect

    Farmer, J C; Blink, J A; Shaw, H F

    2008-10-02

    The worldwide energy consumption in 2003 was 421 quadrillion Btu (Quads), and included 162 quads for oil, 99 quads for natural gas, 100 quads for coal, 27 quads for nuclear energy, and 33 quads for renewable sources. The projected worldwide energy consumption for 2030 is 722 quads, corresponding to an increase of 71% over the consumption in 2003. The projected consumption for 2030 includes 239 quads for oil, 190 quads for natural gas, 196 quads for coal, 35 quads for nuclear energy, and 62 quads for renewable sources [International Energy Outlook, DOE/EIA-0484, Table D1 (2006) p. 133]. The current fleet of light water reactors (LRWs) provides about 20% of current U.S. electricity, and about 16% of current world electricity. The demand for electricity is expected to grow steeply in this century, as the developing world increases its standard of living. With the increasing price for oil and gasoline within the United States, as well as fear that our CO2 production may be driving intolerable global warming, there is growing pressure to move away from oil, natural gas, and coal towards nuclear energy. Although there is a clear need for nuclear energy, issues facing waste disposal have not been adequately dealt with, either domestically or internationally. Better technological approaches, with better public acceptance, are needed. Nuclear power has been criticized on both safety and waste disposal bases. The safety issues are based on the potential for plant damage and environmental effects due to either nuclear criticality excursions or loss of cooling. Redundant safety systems are used to reduce the probability and consequences of these risks for LWRs. LIFE engines are inherently subcritical, reducing the need for systems to control the fission reactivity. LIFE engines also have a fuel type that tolerates much higher temperatures than LWR fuel, and has two safety systems to remove decay heat in the event of loss of coolant or loss of coolant flow. These features of

  16. On the Reprocessing and Reanalysis of Observations for Climate

    NASA Technical Reports Server (NTRS)

    Bosilovich, Michael G.; Kennedy, John; Dee, Dick; Allan, R.; O'Neill, Alan

    2013-01-01

    The long observational record is critical to our understanding of the Earths climate, but most observing systems were not developed with a climate objective in mind. As a result, tremendous efforts have gone into assessing and reprocessing the data records to improve their usefulness in climate studies. The purpose of this paper is to both review recent progress in reprocessing and reanalyzing observations, and to summarize the challenges that must be overcome in order to improve our understanding of climate and variability. Reprocessing improves data quality through more scrutiny and improved retrieval techniques for individual observing systems, while reanalysis merges many disparate observations with models through data assimilation, yet both aim to provide an climatology of Earth processes. Many challenges remain, such as tracking the improvement of processing algorithms and limited spatial coverage. Reanalyses have fostered significant research, yet reliable global trends in many physical fields are not yet attainable, despite significant advances in data assimilation and numerical modeling. Oceanic reanalyses have made significant advances in recent years, but will only be discussed here in terms of progress toward integrated Earth system analyses. Climate data sets are generally adequate for process studies and large-scale climate variability. Communication of the strengths, limitations and uncertainties of reprocessed observations and reanalysis data, not only among the community of developers, but also with the extended research community, including the new generations of researchers and the decision makers is crucial for further advancement of the observational data records. It must be emphasized that careful investigation of the data and processing methods are required to use the observations appropriately.

  17. Legal implications of single-use medical device reprocessing.

    PubMed

    Larose, Emily

    2013-01-01

    Over 10 years ago, the Public Health Agency of Canada released the results of a nation-wide survey of hospitals that demonstrated that the reuse of single-use medical devices was widespread in Canadian healthcare institutions. In this article, the author discusses the reuse and reprocessing of these devices, as well as the risks this practice presents. She then goes on to outline the legal implications of reusing single-use devices.

  18. Reprocessing of metallurgical slag into materials for the building industry

    SciTech Connect

    Pioro, L.S.; Pioro, I.L

    2004-07-01

    Several methods of reprocessing metallurgical (blast furnace) slag into materials for the building industry, based on melting aggregates with submerged combustion, were developed and tested. The first method involves melting hot slag with some additives directly in a slag ladle with a submerged gas-air burner, with the objective of producing stabilized slag or glass-ceramic. The second method involves direct draining of melted slag from a ladle into the slag receiver, with subsequent control of the slag draining into the converter where special charging materials are added to the melt, with the objective of producing glass-ceramic. A third method involves melting cold slag with some additives inside a melting converter with submerged gas-air burners, with the objective of producing glass-ceramic fillers for use in road construction. Specific to the melting process is the use of a gas-air mixture with direct combustion inside the melt. This feature provides melt bubbling to help achieve maximum heat transfer from combustion products to the melt, improve mixing (and therefore homogeneity of the melt), and increases the rate of chemical reactions. The experimental data for different aspects of the proposed methods are presented. The reprocessed blast-furnace slag in the form of granules can be used as fillers for concretes, asphalts, and as additives in the production of cement, bricks and other building materials. As well, reprocessed blast-furnace slag can be poured into forms for the production of glass-ceramic tiles.

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

    SciTech Connect

    Soelberg, Nick R.; Garn, Troy G.; Greenhalgh, Mitchell R.; Law, Jack D.; Jubin, Robert; Strachan, Denis M.; Thallapally, Praveen K.

    2013-01-01

    The removal of volatile radionuclides generated during used nuclear fuel reprocessing in the US is almost certain to be necessary for the licensing of a reprocessing facility in the US. Various control technologies have been developed, tested, or used over the past 50 years for control of volatile radionuclide emissions from used fuel reprocessing plants. The US DOE has sponsored, since 2009, an Off-gas Sigma Team to perform research and development focused on the most pressing volatile radionuclide control and immobilization problems. In this paper, we focus on the control requirements and methodologies for85Kr and129I. Numerous candidate technologies have been studied and developed at laboratory and pilot-plant scales in an effort to meet the need for high iodine control efficiency and to advance alternatives to cryogenic separations for krypton control. Several of these show promising results. Iodine decontamination factors as high as 105, iodine loading capacities, and other adsorption parameters including adsorption rates have been demonstrated under some conditions for both silver zeolite (AgZ) and Ag-functionalized aerogel. Sorbents, including an engineered form of AgZ and selected metal organic framework materials (MOFs), have been successfully demonstrated to capture Kr and Xe without the need for separations at cryogenic temperatures.

  20. Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

    SciTech Connect

    R. Johansen

    2011-09-01

    Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

  1. GENERIC, COMPONENT FAILURE DATA BASE FOR LIGHT WATER AND LIQUID SODIUM REACTOR PRAs

    SciTech Connect

    S. A. Eide; S. V. Chmielewski; T. D. Swantz

    1990-02-01

    A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs) . The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates . Using this approach, most of the failure rates are based on actual plant data rather than existing estimates .

  2. Grouping of light water reactors for evaluation of decay heat removal capability

    SciTech Connect

    Karol, R.; Fresco, A.; Perkins, K.R.

    1984-06-01

    This grouping report provides a compilation of decay heat removal systems (DHRS) data for operating commercial light water reactors. The reactors have been divided into 12 groups based on similarity of the DHRS and related systems as part of the NRC Task Action Plan on Shutdown Decay Heat Removal Requirements.

  3. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which...

  4. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which...

  5. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which...

  6. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which...

  7. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which...

  8. Effect of peracetic acid reprocessing on the transport characteristics of polysulfone hemodialyzers.

    PubMed

    Wolff, Susanne H; Zydney, Andrew L

    2005-02-01

    Peracetic acid is used extensively for reprocessing hemodialyzers, despite several indications that reprocessing alters the dialyzer transport characteristics. The objective of this study was to obtain quantitative data for the effects of peracetic acid reprocessing on the clearance and sieving coefficients of urea, vitamin B12, and polydisperse dextrans using Fresenius F80A polysulfone dialyzers. Reprocessing restored the urea and vitamin B12 clearance to close to their original values. However, the reprocessed dialyzers had substantially lower clearance of the larger molecular weight dextrans, which was attributed to reductions in the effective pore size caused by residual plasma proteins within the membrane. Storage in peracetic acid provided some additional removal of residual proteins, although the clearance and sieving coefficients of the larger dextrans remained well below their original values. Peracetic acid caused no degradation of the membrane polymer, in sharp contrast to results obtained with bleach reprocessing.

  9. High-level disinfection, sterilization, and antisepsis: current issues in reprocessing medical and surgical instruments.

    PubMed

    Seavey, Rose

    2013-05-01

    Technology is rapidly changing many aspects of health care. The intricate design of instruments, the configuration of instrument trays, and evidence-based practice have resulted in the need for complicated and specific reprocessing recommendations from instrument manufacturers. Patient safety depends on instruments that are appropriately cared for and adequately reprocessed. This article covers current issues that sterile processing and operating room professionals must deal with regarding reprocessing of medical and surgical instruments.

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

    SciTech Connect

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

    1982-04-01

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

  11. High-level disinfection of gastrointestinal endoscope reprocessing.

    PubMed

    Chiu, King-Wah; Lu, Lung-Sheng; Chiou, Shue-Shian

    2015-02-20

    High level disinfection (HLD) of the gastrointestinal (GI) endoscope is not simply a slogan, but rather is a form of experimental monitoring-based medicine. By definition, GI endoscopy is a semicritical medical device. Hence, such medical devices require major quality assurance for disinfection. And because many of these items are temperature sensitive, low-temperature chemical methods, such as liquid chemical germicide, must be used rather than steam sterilization. In summarizing guidelines for infection prevention and control for GI endoscopy, there are three important steps that must be highlighted: manual washing, HLD with automated endoscope reprocessor, and drying. Strict adherence to current guidelines is required because compared to any other medical device, the GI endoscope is associated with more outbreaks linked to inadequate cleaning or disinfecting during HLD. Both experimental evaluation on the surveillance bacterial cultures and in-use clinical results have shown that, the monitoring of the stringent processes to prevent and control infection is an essential component of the broader strategy to ensure the delivery of safe endoscopy services, because endoscope reprocessing is a multistep procedure involving numerous factors that can interfere with its efficacy. Based on our years of experience in the surveillance of culture monitoring of endoscopic reprocessing, we aim in this study to carefully describe what details require attention in the GI endoscopy disinfection and to share our experience so that patients can be provided with high quality and safe medical practices. Quality management encompasses all aspects of pre- and post-procedural care including the efficiency of the endoscopy unit and reprocessing area, as well as the endoscopic procedure itself.

  12. Reprocessing of ices in turbulent protoplanetary disks: Carbon and nitrogen chemistry

    SciTech Connect

    Furuya, Kenji; Aikawa, Yuri

    2014-08-01

    We study the influence of the turbulent transport on ice chemistry in protoplanetary disks, focusing on carbon- and nitrogen-bearing molecules. Chemical rate equations are solved with the diffusion term, mimicking the turbulent mixing in the vertical direction. Turbulence can bring ice-coated dust grains from the midplane to the warm irradiated disk surface, and the ice mantles are reprocessed by photoreactions, thermal desorption, and surface reactions. The upward transport decreases the abundance of methanol and ammonia ices at r ≲ 30 AU because warm dust temperature prohibits their reformation on grain surfaces. This reprocessing could explain the smaller abundances of carbon and nitrogen bearing molecules in cometary coma than those in low-mass protostellar envelopes. We also show the effect of mixing on the synthesis of complex organic molecules (COMs) in two ways: (1) transport of ices from the midplane to the disk surface and (2) transport of atomic hydrogen from the surface to the midplane. The former enhances the COMs formation in the disk surface, while the latter suppresses it in the midplane. Then, when mixing is strong, COMs are predominantly formed in the disk surface, while their parent molecules are (re)formed in the midplane. This cycle expands the COMs distribution both vertically and radially outward compared with that in the non-turbulent model. We derive the timescale of the sink mechanism by which CO and N{sub 2} are converted to less volatile molecules to be depleted from the gas phase and find that the vertical mixing suppresses this mechanism in the inner disks.

  13. Meeting Summary Advanced Light Water Reactor Fuels Industry Meeting Washington DC October 27 - 28, 2011

    SciTech Connect

    Not Listed

    2011-11-01

    The Advanced LWR Fuel Working Group first met in November of 2010 with the objective of looking 20 years ahead to the role that advanced fuels could play in improving light water reactor technology, such as waste reduction and economics. When the group met again in March 2011, the Fukushima incident was still unfolding. After the March meeting, the focus of the program changed to determining what we could do in the near term to improve fuel accident tolerance. Any discussion of fuels with enhanced accident tolerance will likely need to consider an advanced light water reactor with enhanced accident tolerance, along with the fuel. The Advanced LWR Fuel Working Group met in Washington D.C. on October 72-18, 2011 to continue discussions on this important topic.

  14. International academic program in technologies of light-water nuclear reactors. Phases of development and implementation

    NASA Astrophysics Data System (ADS)

    Geraskin, N. I.; Glebov, V. B.

    2017-01-01

    The results of implementation of European educational projects CORONA and CORONA II dedicated to preserving and further developing nuclear knowledge and competencies in the area of technologies of light-water nuclear reactors are analyzed. Present article addresses issues of design and implementation of the program for specialized training in the branch of technologies of light-water nuclear reactors. The systematic approach has been used to construct the program for students of nuclear specialties, which corresponding to IAEA standards and commonly accepted nuclear principles recognized in the European Union. Possibilities of further development of the international cooperation between countries and educational institutions are analyzed. Special attention is paid to e-learning/distance training, nuclear knowledge preservation and interaction with European Nuclear Education Network.

  15. An overview of a nuclear reprocessing plant Human Factors programme.

    PubMed

    Kirwan, Barry

    2003-09-01

    This paper presents a case study of a large Human Factors programme applied in the nuclear fuel reprocessing industry (1987-1991). The paper outlines the key Human Factors issues addressed, as well as the impacts achieved, and gives an indication of the resources utilised (approximately 15 person-years of effort). It also considers the starting point of the programme, in terms of the factors that led to the need for such an extensive programme. Some general lessons learned are given at the end of the paper.

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

    DOEpatents

    Campbell, David O.; Buxton, Samuel R.

    1981-01-01

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

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

    DOEpatents

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

    1980-06-16

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

  18. Method for reprocessing and separating spent nuclear fuels

    DOEpatents

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

    1983-01-01

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

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

    DOEpatents

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

    1982-01-19

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

  20. Materials management in an internationally safeguarded fuels reprocessing plant

    SciTech Connect

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

    1980-04-01

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

  1. Sexual violence: psychiatric healing with eye movement reprocessing and desensitization.

    PubMed

    Posmontier, Bobbie; Dovydaitis, Tiffany; Lipman, Kenneth

    2010-08-01

    Sexual violence, which affects one in three women worldwide, can result in significant psychiatric morbidity and suicide. Eye movement desensitization and reprocessing (EMDR) offers health care providers the option of a brief psychiatric intervention that can result in psychiatric healing in as few as four sessions. Because health care providers often hear stories of sexual violence from their patients, they are in an ideal position to make recommendations for treatment. The purpose of this article is to introduce health care providers to the technique of EMDR, review safety and appropriateness, and discuss clinical and research implications.

  2. Eye movement desensitization and reprocessing: a conceptual framework.

    PubMed

    Menon, Sukanya B; Jayan, C

    2010-07-01

    Eye movement desensitization and reprocessing (EMDR) is a method which was initially used for the treatment of post-traumatic stress disorder. But it is now being used in different therapeutic situations. EMDR is an eight-phase treatment method. History taking, client preparation, assessment, desensitization, installation, body scan, closure and reevaluation of treatment effect are the eight phases of this treatment which are briefly described. A case report is also depicted which indicates the efficacy of EMDR. The areas where EMDR is used and the possible ways through which it is working are also described.

  3. Residual bioburden in reprocessed side-view endoscopes used for endoscopic retrograde cholangiopancreatography (ERCP)

    PubMed Central

    Ubhayawardana, D. L. N. L.; Kottahachchi, J.; Weerasekera, M. M.; Wanigasooriya, I. W. M. P.; Fernando, S. S. N.; De Silva, M.

    2013-01-01

    Background and study aim: Worldwide some endoscopy units routinely continue to use manual reprocessing techniques for disinfection of side-view endoscopes. The aim of this study was to evaluate the outcome quality of manual reprocessing techniques for removal and inactivation of the bioburden from side-view endoscopes used for endoscopic retrograde cholangiopancreatography (ERCP) in a tertiary referral endotherapy unit in Sri Lanka. Methods: 102 samples obtained from two different flexible side-view endoscopes (Olympus TJF Q 180V and Olympus TJF 160 R) were tested for microbial growth. Three samples were collected each time; one swab from the tip before and another after manual reprocessing. The third sample was collected by flushing the working channel with sterile normal saline after manual reprocessing. Microorganisms were identified by culturing the samples. Result:: After reprocessing, culture-positive rates were 20 % and 9 % for the samples obtained from the tip and the working channel of the side-view endoscopes, respectively. Klebsiella spp. and Candida spp. were found to be the commonest microorganisms in the samples from the tips and from the working channels, respectively, of the reprocessed side-view endoscopes. Conclusion: There is a high culture-positive rate after reprocessing of the side-view endoscopes using the manual reprocessing procedure, despite strict adherence to the protocol for reprocessing. PMID:26135507

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

    SciTech Connect

    Not Available

    1981-06-01

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

  5. Uranium, its impact on the national and global energy mix; and its history, distribution, production, nuclear fuel-cycle, future, and relation to the environment

    USGS Publications Warehouse

    Finch, Warren Irvin

    1997-01-01

    The many aspects of uranium, a heavy radioactive metal used to generate electricity throughout the world, are briefly described in relatively simple terms intended for the lay reader. An adequate glossary of unfamiliar terms is given. Uranium is a new source of electrical energy developed since 1950, and how we harness energy from it is explained. It competes with the organic coal, oil, and gas fuels as shown graphically. Uranium resources and production for the world are tabulated and discussed by country and for various energy regions in the United States. Locations of major uranium deposits and power reactors in the United States are mapped. The nuclear fuel-cycle of uranium for a typical light-water reactor is illustrated at the front end-beginning with its natural geologic occurrence in rocks through discovery, mining, and milling; separation of the scarce isotope U-235, its enrichment, and manufacture into fuel rods for power reactors to generate electricity-and at the back end-the reprocessing and handling of the spent fuel. Environmental concerns with the entire fuel cycle are addressed. The future of the use of uranium in new, simplified, 'passively safe' reactors for the utility industry is examined. The present resource assessment of uranium in the United States is out of date, and a new assessment could aid the domestic uranium industry.

  6. Design basis for protection of light water nuclear power plants against effects of postulated pipe rupture

    SciTech Connect

    Not Available

    1981-01-01

    This standard addresses the design bases for light water reactor, nuclear power plant structures and components essential for the protection of public health and safety from the potential adverse effects of pipe whip, jet impingement, pressurization of compartments outside containment, environmental conditions and flooding associated with a postulated pipe rupture. The design bases for missile protection and the design bases for containment pressurization are not within this standard.

  7. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, William E.; Trapp, Turner J.

    1985-01-01

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  8. Advanced Light Water Reactor Program: Program management and staff review methodology

    SciTech Connect

    Moran, D.H.

    1986-12-01

    This report summarizes the NRC/EPRI coordinated effort to develop design requirements for a standardized advanced light water reactor (ALWR) and the procedures for screening and applying new generic safety issues to this program. The end-product will be an NRC-approved ALWR Requirements Document for use by the nuclear industry in generating designs of LWRs to be constructed for operation in the 1990s and beyond.

  9. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, W.E.; Trapp, T.J.

    1983-06-10

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  10. Establishment of a Hub for the Light Water Reactor Sustainability Online Monitoring Community

    SciTech Connect

    Nancy J. Lybeck; Magdy S. Tawfik; Binh T. Pham

    2011-08-01

    Implementation of online monitoring and prognostics in existing U.S. nuclear power plants will involve coordinating the efforts of national laboratories, utilities, universities, and private companies. Internet-based collaborative work environments provide necessary communication tools to facilitate interaction between geographically diverse participants. Available technologies were considered, and a collaborative workspace was established at INL as a hub for the light water reactor sustainability online monitoring community.

  11. Overview of the US Department of Energy Light Water Reactor Sustainability Program

    SciTech Connect

    K. A. McCarthy; D. L. Williams; R. Reister

    2012-05-01

    The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.

  12. Immobilization of Fast Reactor First Cycle Raffinate

    SciTech Connect

    Langley, K. F.; Partridge, B. A.; Wise, M.

    2003-02-26

    This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.

  13. Replication of the Apparent Excess Heat Effect in a Light Water-Potassium Carbonate-Nickel Electrolytic Cell

    NASA Technical Reports Server (NTRS)

    Niedra, Janis M.; Myers, Ira T.; Fralick, Gustave C.; Baldwin, Richard S.

    1996-01-01

    Replication of experiments claiming to demonstrate excess heat production in light water-Ni-K2CO3 electrolytic cells was found to produce an apparent excess heat of 11 W maximum, for 60 W electrical power into the cell. Power gains range from 1.06 to 1.68. The cell was operated at four different dc current levels plus one pulsed current run at 1 Hz, 10% duty cycle. The 28 liter cell used in these verification tests was on loan from a private corporation whose own tests with similar cells are documented to produce 50 W steady excess heat for a continuous period exceeding hundreds of days. The apparent excess heat can not be readily explained either in terms of nonlinearity of the cell's thermal conductance at a low temperature differential or by thermoelectric heat pumping. However, the present data do admit efficient recombination of dissolved hydrogen-oxygen as an ordinary explanation. Calorimetry methods and heat balance calculations for the verification tests are described. Considering the large magnitude of benefit if this effect is found to be a genuine new energy source, a more thorough investigation of evolved heat in the nickel-hydrogen system in both electrolytic and gaseous loading cells remains warranted.

  14. Acoustic emission and guided ultrasonic waves for detection and continuous monitoring of cracks in light water reactor components

    SciTech Connect

    Meyer, R. M.; Coble, J.; Ramuhalli, P.; Watson, B.; Cumblidge, S. E.; Doctor, S. R.; Bond, L. J.

    2012-07-01

    Acoustic emission (AE) and guided ultrasonic waves (GUW) are considered for continuous monitoring and detection of cracks in Light Water Reactor (LWR) components. In this effort, both techniques are applied to the detection and monitoring of fatigue crack growth in a full scale pipe component. AE results indicated crack initiation and rapid growth in the pipe, and significant GUW responses were observed in response to the growth of the fatigue crack. After initiation, the crack growth was detectable with AE for approximately 20,000 cycles. Signals associated with initiation and rapid growth were distinguished based on total rate of activity and differences observed in the centroid frequency of hits. An intermediate stage between initiation and rapid growth was associated with significant energy emissions, though few hits. GUW exhibit a nearly monotonic trend with crack length with an exception of measurements obtained at crack lengths of 41 mm and 46 mm. Coupling variability and shadowing by the electro-discharge machining (EDM) starter notch set the lower limit of detectability. (authors)

  15. Environmentally assisted cracking in light-water reactors: Semi-annual report, January--June 1997. Volume 24

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1998-04-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1997 to June 1997. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Types 304 and 304L SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle is equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in low-DO, simulated pressurized water reactor environments.

  16. Neptunium flow-sheet verification at reprocessing plants

    SciTech Connect

    Rance, P.; Chesnay, B.; Killeen, T.; Murray, M.; Nikkinen, M.; Petoe, A.; Plumb, J.; Saukkonen, H.

    2007-07-01

    Due to their fissile nature, neptunium and americium have at least a theoretical potential application as nuclear explosives and their proliferation potential was considered by the IAEA in studies in the late 1990's. This work was motivated by an increased awareness of the proliferation potential of americium and neptunium and a number of emerging projects in peaceful nuclear programmes which could result in an increase in the available quantities of these minor actinides. The studies culminated in proposals for various voluntary measures including the reporting of international transfers of separated americium and neptunium, declarations concerning the amount of separated neptunium and americium held by states and the application of flow-sheet verification to ensure that facilities capable of separating americium or neptunium are operated in a manner consistent with that declared. This paper discusses the issue of neptunium flowsheet verification in reprocessing plants. The proliferation potential of neptunium is first briefly discussed and then the chemistry of neptunium relevant to reprocessing plants described with a view to indicating a number of issues relevant to the verification of neptunium flow-sheets. Finally, the scope of verification activities is discussed including analysis of process and engineering design information, plant monitoring and sampling and the potential application of containment and surveillance measures. (authors)

  17. Development of engineering technology basis for industrialization of pyrometallurgical reprocessing

    SciTech Connect

    Koyama, Tadafumi; Hijikata, Takatoshi; Yokoo, Takeshi; Inoue, Tadashi

    2007-07-01

    Development of the engineering technology basis of pyrometallurgical reprocessing is a key issue for industrialization. For development of the transport technologies of molten salt and liquid cadmium at around 500 deg. C, a salt transport test rig and a metal transport test rig were installed in Ar glove box. Function of centrifugal pump and 1/2' declined tubing were confirmed with LiCl- KCl molten salt. The transport behavior of molten salt was found to follow that of water. Function of centrifugal pump, vacuum sucking and 1/2' declined tubing were confirmed with liquid Cd. With employing the transport technologies, industrialization applicable electro-refiner was newly designed and engineering-scale model was fabricated in Ar glove box. The electro-refiner has semi-continuous liquid Cd cathode instead of conventional one used in small-scale tests. With using actinide-simulating elements, demonstration of industrial-scale throughput will be carried out in this electro-refiner for more precise evaluation of industrialization potential of pyrometallurgical reprocessing. (authors)

  18. Development of high-fidelity multiphysics system for light water reactor analysis

    NASA Astrophysics Data System (ADS)

    Magedanz, Jeffrey W.

    There has been a tendency in recent years toward greater heterogeneity in reactor cores, due to the use of mixed-oxide (MOX) fuel, burnable absorbers, and longer cycles with consequently higher fuel burnup. The resulting asymmetry of the neutron flux and energy spectrum between regions with different compositions causes a need to account for the directional dependence of the neutron flux, instead of the traditional diffusion approximation. Furthermore, the presence of both MOX and high-burnup fuel in the core increases the complexity of the heat conduction. The heat transfer properties of the fuel pellet change with irradiation, and the thermal and mechanical expansion of the pellet and cladding strongly affect the size of the gap between them, and its consequent thermal resistance. These operational tendencies require higher fidelity multi-physics modeling capabilities, and this need is addressed by the developments performed within this PhD research. The dissertation describes the development of a High-Fidelity Multi-Physics System for Light Water Reactor Analysis. It consists of three coupled codes -- CTF for Thermal Hydraulics, TORT-TD for Neutron Kinetics, and FRAPTRAN for Fuel Performance. It is meant to address these modeling challenges in three ways: (1) by resolving the state of the system at the level of each fuel pin, rather than homogenizing entire fuel assemblies, (2) by using the multi-group Discrete Ordinates method to account for the directional dependence of the neutron flux, and (3) by using a fuel-performance code, rather than a Thermal Hydraulics code's simplified fuel model, to account for the material behavior of the fuel and its feedback to the hydraulic and neutronic behavior of the system. While the first two are improvements, the third, the use of a fuel-performance code for feedback, constitutes an innovation in this PhD project. Also important to this work is the manner in which such coupling is written. While coupling involves combining

  19. Analysis of pellet cladding interaction and creep of U 3SIi2 fuel for use in light water reactors

    NASA Astrophysics Data System (ADS)

    Metzger, Kathryn E.

    Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels that "in comparison with the standard UO2- Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events." To be economically viable, proposed accident tolerant fuels and claddings should be backward compatible with LWR designs, provide significant operating cost improvements such as power uprates, increased fuel burnup, or increased cycle length. In terms of safety, an alternative fuel pellet must have resistance to water corrosion comparable to UO2, thermal conductivity equal to or larger than that of UO2, and a melting temperature that allows the material to remain solid under power reactor conditions. Among the candidates, U3Si2 has a number of advantageous thermophysical properties, including; high density, high thermal conductivity at room temperature, and a high melting temperature. These properties support its use as an accident tolerant fuel while its high uranium density is capable of supporting uprates to the LWR fleet. This research characterizes U3Si2 pellets and analyzes U3Si2 under light water reactor conditions using the fuel performance code BISON. While some thermophysical properties for U3Si2 have been found in the literature, the irradiation behavior is sparse and limited to experience with dispersion fuels. Accordingly, the creep behavior for U3Si2 has been unknown, making it

  20. [HYGEA (Hygiene in gastroenterology--endoscope reprocessing): Study on quality of reprocessing flexible endoscopes in hospitals and in the practice setting].

    PubMed

    Bader, L; Blumenstock, G; Birkner, B; Leiss, O; Heesemann, J; Riemann, J F; Selbmann, H-K

    2002-03-01

    The quality of reprocessing gastroscopes, colonoscopes and duodenoscopes in daily routine of 25 endoscopy departments in hospitals and 30 doctors with their own practices was evaluated by microbiological testing in the HYGEA interventional study. In 2 test periods, endoscopes ready for use in patients were found contaminated at high rates (period 1: 49 % of 152 endoscopes; period 2: 39 % of 154 endoscopes). Culture of bacterial fecal flora (E. coli, coliform enterobacteriaceae, enterococci) was interpreted indicating failure of cleaning procedure and disinfection of endoscopes. Detection of Pseudomonas spp. (especially P. aeruginosa) and other non-fermenting rods - indicating microbially insufficient final rinsing and incomplete drying of the endoscope or a contaminated flushing equipment for the air/water-channel - pointed out endoscope recontamination during reprocessing or afterwards. Cause for complaint was found in more than 50 % of endoscopy facilities tested (period 2: 5 in hospitals, 25 practices). Reprocessing endoscopes in fully automatic chemo-thermally decontaminating washer-disinfectors with disinfection of final rinsing water led to much better results than manual or semi-automatic procedures (failure rate of endoscopy facilities in period 2 : 3 of 28 with fully automatic, 8 of 12 with manual, 9 of 15 with semi-automatic reprocessing). The study results give evidence for the following recommendations: 1. Manual brushing of all accessible endoscope channels has to be performed even before further automatic reprocessing; 2. For final endoscope rinsing, water or aqua dest. should only be used disinfected or sterile-filtered; 3. Endoscopes have to be dried thoroughly using compressed air prior to storage; 4. Bottle and tube for air/water-channel flushing have to be reprocessed daily by disinfection or sterilization, and in use, the bottle have to be filled exclusively with sterile water. The HYGEA study shows that microbiological testing of endoscopes is

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

    SciTech Connect

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

    1991-01-01

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

  2. Using Eye Movement Desensitization and Reprocessing To Enhance Treatment of Couples.

    ERIC Educational Resources Information Center

    Protinsky, Howard; Sparks, Jennifer; Flemke, Kimberly

    2001-01-01

    Eye Movement Desensitization and Reprocessing (EMDR) as a clinical technique may enhance treatment effectiveness when applied in couple therapy that is emotionally and experientially oriented. Clinical experience indicates EMDR-based interventions are useful for accessing and reprocessing intense emotions in couple interactions. EMDR can amplify…

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

    SciTech Connect

    Not Available

    1980-09-01

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

  4. The Economic, repository and proliferation implications of advanced nuclear fuel cycle

    SciTech Connect

    Deinert, Mark; Cady, K B

    2011-09-04

    The goal of this project was to compare the effects of recycling actinides using fast burner reactors, with recycle that would be done using inert matrix fuel burned in conventional light water reactors. In the fast reactor option, actinides from both spent light water and fast reactor fuel would be recycled. In the inert matrix fuel option, actinides from spent light water fuel would be recycled, but the spent inert matrix fuel would not be reprocessed. The comparison was done over a limited 100-year time horizon. The economic, repository and proliferation implications of these options all hinge on the composition of isotopic byproducts of power production. We took the perspective that back-end economics would be affected by the cost of spent fuel reprocessing (whether conventional uranium dioxide fuel, or fast reactor fuel), fuel manufacture, and ultimate disposal of high level waste in a Yucca Mountain like geological repository. Central to understanding these costs was determining the overall amount of reprocessing needed to implement a fast burner, or inert matrix fuel, recycle program. The total quantity of high level waste requiring geological disposal (along with its thermal output), and the cost of reprocessing were also analyzed. A major advantage of the inert matrix fuel option is that it could in principle be implemented using the existing fleet of commercial power reactors. A central finding of this project was that recycling actinides using an inert matrix fuel could achieve reductions in overall actinide production that are nearly very close to those that could be achieved by recycling the actinides using a fast burner reactor.

  5. Reduction of Sodium Nitrate Liquid Waste in Nuclear Reprocessing Plants

    SciTech Connect

    Numata, M.; Mihara, S.; Kojima, S.; Ito, H.; Kato, T.

    2006-07-01

    Sodium nitrate solution has been generated from nuclear reprocessing plant as a result of neutralization of nitric acid. The sodium nitrate has been immobilized by bitumen, cement or other material in the site and waste packages have been produced. In order to reduce an environmental impact of the waste packages from the reprocessing plant, it is preferable to decompose nitrate ion to harmless gases such as nitrogen. A combination of formic acid and catalyst has been proposed for this purpose. But, the method is inadequate for a full decomposition of the nitrate ion. In addition, a mixture of NO and NO{sub 2} is produced during the reaction. Formaldehyde and hydrazine were selected as reductants and a combined use of Pd-Cu catalyst was tried to decompose the nitrate ion. As a result, the nitrate ion can almost entirely be decomposed without any generation of NO and NO{sub 2}. The test was conducted by 1 L flask. In case of formaldehyde, nitrate ion concentration can be reduced from 0.017 mol/l to 3.9x10{sup -4} mol/l. In case of hydrazine, nitrate concentration can be decreased from 2.8 mol/l to 9.5 x 10{sup -3} mol/l and ammonium ion is detected. The ammonium ion concentration in the final solution is 0.12 mol/l when 2.8 mol/l nitrate is reduced by hydrazine. Chemical reactions for formaldehyde on the Pd-Cu catalyst are estimated as combination of: NO{sub 3-} + HCHO = NO{sub 2-} + HCOOH; 2NO{sub 2-} + 3HCOOH = N{sub 2} + 3CO{sub 2} + 2H{sub 2}O + 2OH-; 4NO{sub 2-} + 3HCHO = 2N{sub 2} + 3CO{sub 2} + H{sub 2}O + 4OH-. the other hand, for hydrazine with the Pd-Cu catalyst: 3N{sub 2}H{sub 4} = 2NH{sub 3} + 2N{sub 2} + 3H{sub 2}; NO{sub 3-} + H{sub 2} = NO{sub 2-} + H{sub 2}O; NO{sub 2-} + NH{sub 3} = N{sub 2} + H{sub 2}O + OH-. The fundamental research shows that the combination usage of the Pd-Cu catalyst and formaldehyde or hydrazine is applicable for the reduction of nitrate liquid waste in the nuclear reprocessing plant. (authors)

  6. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    SciTech Connect

    Holbrook, Mark; Kinsey, Jim

    2015-03-01

    In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

  7. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    SciTech Connect

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

  8. Molten tin reprocessing of spent nuclear fuel elements

    DOEpatents

    Heckman, Richard A.

    1983-01-01

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

  9. Alternate extractants to tributyl phosphate for reactor fuel reprocessing

    SciTech Connect

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

    1983-01-01

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

  10. Survey of Dynamic Simulation Programs for Nuclear Fuel Reprocessing

    SciTech Connect

    Troy J. Tranter; Daryl R. Haefner

    2008-06-01

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

  11. GRACE Mission And Data Re-Processing Status

    NASA Astrophysics Data System (ADS)

    Bettadpur, S. V.; Tapley, B. D.

    2006-05-01

    The Gravity Recovery And Climate Experiment was launched on Mar 17, 2002 in order to measure mass flux within the Earth system through its effects on Earth's gravity field. Since that time, using the inter-satellite tracking data between the twin GRACE satellites, monthly gravity field estimates for nearly 4 years have been delivered to the user community. These fields have shown clear evidence of hydrological, oceanographic & glaciological phenomena. The GRACE Science Data System is in the midst of a re-processing activity, focusing on improvements to the background models and processing methodology. In addition to an overview of the mission status and future plans, this paper describes the status of the new results from the re- processing. These include changes to the background models, improvement in the processing, and the resulting error characteristics.

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

    SciTech Connect

    FRANCIS,A.J.

    2006-10-18

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

  13. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    SciTech Connect

    Pan, Paul Y

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

  14. Utility leadership in reopening the nuclear option with advanced light water reactors

    SciTech Connect

    Marston, T.U.; Layman, W.H. )

    1992-01-01

    Since 1981, the Electric Power Research Institute (EPRI) has been pursing the development of the advanced light water reactor (ALWR). The ALWR Program is comprised of five phases and are described in the paper. In order to meet the anticipated baseline power generation requirements in the US, the Nuclear Power Oversight Committee (NPOC) has developed a strategic plan for ALWR implementation in order to regain the nuclear option in the United States. The paper also covers the policies behind the utility requirements, the status of ALWR developments in the United States, the electricity demands during the period 1990-2010, and some of the innovative features of the passive plants presently under design.

  15. Analysis of nuclear proliferation resistance reprocessing and recycling technologies

    SciTech Connect

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

    2011-05-01

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

  16. ''Super 2D,'' Innovative seismic reprocessing: A case history

    SciTech Connect

    Conne, D.K.M.; Bolander, A.G.; MacDonald, R.J.; Strelioff, D.M.

    1988-01-01

    The ''Super 2D'' processing sequence involves taking a randomly oriented grid of multivintage two-dimensional seismic data and reprocessing to tie the data where required, then interpolating the data set to a regular grid suitable for three-dimensional processing and interpretation. A data set from Alberta, provided by a Canadian oil company, comprises 15 two-dimensional seismic lines collected and processed over a period of 6 years by various contractors. Field conditions, advances in technology, and changing objectives combined to result in a data set that densely sampled a small area, but did not tie in well enough to be interpreted as a whole. The data mistied in time, phase, and frequency, as well as having a problem with multiples in the zone of interest that had been partly attenuated in varying degrees. Therefore, the first objective of reprocessing was to resolve these problems. The authors' current land data processing sequence, which includes frequency balancing followed by source wavelet designature, F/K multiple attenuation, trim statics, and F-X filtering, as well as close attention to statics and velocity control, resolved all the mistie issues and produced a standardized data volume. This data volume was now suitable for the second stage of this sequence (i.e., interpolating to a regular grid and subsequent three-dimensional processing). The volume was three-dimensionally migrated (finite difference), filtered, and scaled. The full range of three-dimensional display and interpretational options, including loading on an interactive system, are now possible. This, along with standardizing the data set and improving the spatial location of events via three-dimensional migration are the key results of the ''Super 2D'' sequence.

  17. Clinical experience with heat sterilization for reprocessing dialyzers.

    PubMed

    Kaufman, A M; Frinak, S; Godmere, R O; Levin, N W

    1992-01-01

    Use of heat sterilization for dialysis reprocessing offers significant advantages over chemical germicides. Polysulfone dialyzers (Fresenius 60M or 80M) can be sterilized by heating to 105 degrees C for 20 hr, thus permitting clinical trials of this method. One hundred eighty patients received 9,000 treatments. Pyrogenic reactions, sepsis, and subjective symptoms have not occurred. In vitro clearances (Qb 500 ml/min, Qd 800 ml/min) at baseline and after 2-8 uses did not differ (340 +/- 29 vs. 352 +/- 4 ml/min, respectively). KoA determined in vivo did not decrease (baseline 709 +/- 131 vs. 7th use 632 +/- 50 ml/min). Kt/V for urea was not different in 18 patients treated with heat sterilized dialyzers over 6 months when compared with a baseline period with formaldehyde sterilized dialyzers (1.37 +/- 0.12 vs. 1.32 +/- 0.11 at similar time and blood flows). Mean use number was 7.4 (dialyzers limited to 11 uses). Of discarded dialyzers, 44% failed a bedside integrity test (blood side pressurized at > 400 mmHg for 1 min), 36% failed automated fiber bundle or pressure holding tests, 8% had a blood leak, and 12% reached 11 uses. Clinical blood leaks occur in < 0.5% of treatments. Heat sterilization is a safe and effective method of dialysis reprocessing, but quality control of the process is essential. Based on initial clinical experience, heat sterilization of dialyzers for reuse is a promising alternative to chemical disinfection.

  18. (Design of nuclear fuel reprocessing plants, Neuherberg, Munich, Hannover and Wackersdorf, FRG, July 5--19, 1989): Foreign trip report

    SciTech Connect

    Fields, D.E.

    1989-07-24

    The proposed fuel reprocessing site was characterized as to meteorological and hydrological characteristics and population geographical distribution. Data were gathered characterizing the fuel reprocessing plant licensing procedure currently used in the FRG. Comparisons were made of fuel reprocessing in the FRG, France, and Great Britain.

  19. Thermal Properties of Structural Materials Found in Light Water Reactor Vessels

    SciTech Connect

    J. E. Daw; J. L. Rempe; D. L. Knudson

    2009-11-01

    High temperature material property data for structural materials used in existing Light Water Reactors (LWRs) are limited. Often, extrapolated values recommended in the literature differ significantly. To reduce such uncertainties, new data for SA533 Grade B, Class 1 (SA533B1) low alloy steel, Stainless Steel 304 (SS304), and Inconel 600, found in Light Water Reactor (LWR) vessels and penetrations, were acquired and tested using material property systems available at the High Temperature Test Laboratory (HTTL) at the Idaho National Laboratory (INL). Properties measured include thermal expansion, specific heat capacity, and thermal diffusivity for temperatures up to 1200 oC. From these results, thermal conductivity and density were calculated. Results show that, in some cases, previously recommended values for these material differ significantly from measured values at high temperatures. This is especially true for SA533B1, as previous data do not account for the phase transformation of this material between 740 oC and 840 oC.

  20. Conceptual design of a pressure tube light water reactor with variable moderator control

    SciTech Connect

    Rachamin, R.; Fridman, E.; Galperin, A.

    2012-07-01

    This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

  1. Nondestructive examination (NDE) reliability for inservice inspection of light waters reactors

    SciTech Connect

    Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T. )

    1989-11-01

    Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from April 1988 through September 1988. 33 refs., 70 figs., 12 tabs.

  2. Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

    SciTech Connect

    Shannon M. Bragg-Sitton

    2013-09-01

    Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

  3. Overview of the Consortium for the Advanced Simulation of Light Water Reactors (CASL)

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Franceschini, Fausto; Evans, Thomas M.; Gehin, Jess C.

    2016-02-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR) operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a first-principles approach. In order to pursue these goals, CASL has participation from laboratory, academic, and industry partners. These partners are pursuing the solution of ten major "Challenge Problems" in order to advance the state-of-the-art in reactor design and analysis to permit power uprates, higher burnup, life extension, and increased safety. At present, the problems being addressed by CASL are primarily reactor physics-oriented; however, this paper is intended to introduce CASL to the reactor dosimetry community because of the importance of reactor physics modelling and nuclear data to define the source term for that community and the applicability and extensibility of the transport methods being developed.

  4. Nondestructive examination (NDE) reliability for inservice inspection of light water reactors

    SciTech Connect

    Doctor, S.R.; Good, M.S.; Heasler, P.G.; Hockey, R.L.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V. )

    1992-05-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section 11 of the ASME Code. This is a progress report covering the programmatic work from October 1989 through September 1990.

  5. Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

    SciTech Connect

    Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N.; Malstrom, R.A.; Nobles, B.R.; Carson, S.D.; Peterson, P.K.

    1997-11-30

    The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term.

  6. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Progress Report for Work Through September 2003, 2nd Annual/8th Quarterly Report

    SciTech Connect

    Philip E. MacDonald

    2003-09-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation-IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus the need for recirculation and jet pumps, a pressurizer, steam generators, steam separators and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which is also in use around the world.

  7. Spectroscopic Online Monitoring for Process Control and Safeguarding of Radiochemical Fuel Reprocessing Streams

    SciTech Connect

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

    2013-02-24

    There is a renewed interest worldwide to promote the use of nuclear power and close the nuclear fuel cycle. The long term successful use of nuclear power is critically dependent upon adequate and safe processing and disposition of the spent nuclear fuel. Liquid-liquid extraction is a separation technique commonly employed for the processing of the dissolved spent nuclear fuel. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. In addition, the ability for continuous online monitoring allows for numerous benefits. This paper reviews application of the absorption and vibrational spectroscopic techniques supplemented by physicochemical measurements for radiochemical process monitoring. In this context, our team experimentally assessed the potential of Raman and spectrophotometric techniques for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(IV)/Np(V)/Nd(III), respectively, in solutions relevant to spent fuel reprocessing. Both techniques demonstrated robust performance in the repetitive batch measurements of each analyte in a wide concentration range using simulant and commercial dissolved spent fuel solutions. Static spectroscopic measurements served as training sets for the multivariate data analysis to obtain partial least squares predictive models, which were validated using on-line centrifugal contactor extraction tests. Satisfactory prediction of the analytes concentrations in these preliminary experiments warrants further development of the spectroscopy-based methods for radiochemical safeguards and process control.

  8. Spectroscopic Online Monitoring for Process Control and Safeguarding of Radiochemical Fuel Reprocessing Streams - 13553

    SciTech Connect

    Bryan, S.A.; Levitskaia, T.G.; Casella, Amanda; Peterson, James

    2013-07-01

    There is a renewed interest worldwide to promote the use of nuclear power and close the nuclear fuel cycle. The long term successful use of nuclear power is critically dependent upon adequate and safe processing and disposition of the used nuclear fuel. Liquid-liquid extraction is a separation technique commonly employed for the processing of the dissolved spent nuclear fuel. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. This paper discusses application of absorption and vibrational spectroscopic techniques supplemented by physicochemical measurements for radiochemical process monitoring. In this context, our team experimentally assessed the potential of Raman and spectrophotometric techniques for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(IV)/Np(V)/Nd(III), respectively, in solutions relevant to spent fuel reprocessing. Both techniques demonstrated robust performance in the repetitive batch measurements of each analyte in a wide concentration range using simulant and commercial dissolved spent fuel solutions. Static spectroscopic measurements served as training sets for the multivariate data analysis to obtain partial least squares predictive models, which were validated using on-line centrifugal contactor extraction tests. Satisfactory prediction of the analytes concentrations in these preliminary experiments warrants further development of the spectroscopy-based methods for radiochemical safeguards and process control. (authors)

  9. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, Progress Report for Work Through September 2002, 4th Quarterly Report

    SciTech Connect

    Mac Donald, Philip Elsworth

    2002-09-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR. The Generation IV Roadmap effort has identified the thermal spectrum SCWR (followed by the fast spectrum SCWR) as one of the advanced concepts that should be developed for future use. Therefore, the work in this NERI project is addressing both types of SCWRs.

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

    SciTech Connect

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

    2013-07-01

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

  11. Global Assessment of Reprocessed ERS-1 and ERS-2 Altimetry (REAPER) Wind and Wave Products

    NASA Astrophysics Data System (ADS)

    Abdalla, Saleh; Bidlot, Jean-Raymond; Janssen, Peter A. E. M.

    2016-08-01

    ESA has been committed to reprocess its Earth Observation Missions of ERS and ENVISAT to produce long and homogeneous time series suitable for climate studies. The full-mission reprocessing of the ERS Altimetry Level 2 REAPER was completed sometime ago. The REAPER wind and wave products are assessed mainly against ERA-Interim which is a rather consistent model run. Several numerical experiments were carried out in order to assess the impact of assimilating the reprocessed significant wave height data in the ECMWF wave model for potential use in the most recent ECMWF atmospheric reanalysis ERA-5. The results show positive impact.

  12. The Path to Sustainable Nuclear Energy. Basic and Applied Research Opportunities for Advanced Fuel Cycles

    SciTech Connect

    Finck, P.; Edelstein, N.; Allen, T.; Burns, C.; Chadwick, M.; Corradini, M.; Dixon, D.; Goff, M.; Laidler, J.; McCarthy, K.; Moyer, B.; Nash, K.; Navrotsky, A.; Oblozinsky, P.; Pasamehmetoglu, K.; Peterson, P.; Sackett, J.; Sickafus, K. E.; Tulenko, J.; Weber, W.; Morss, L.; Henry, G.

    2005-09-01

    The objective of this report is to identify new basic science that will be the foundation for advances in nuclear fuel-cycle technology in the near term, and for changing the nature of fuel cycles and of the nuclear energy industry in the long term. The goals are to enhance the development of nuclear energy, to maximize energy production in nuclear reactor parks, and to minimize radioactive wastes, other environmental impacts, and proliferation risks. The limitations of the once-through fuel cycle can be overcome by adopting a closed fuel cycle, in which the irradiated fuel is reprocessed and its components are separated into streams that are recycled into a reactor or disposed of in appropriate waste forms. The recycled fuel is irradiated in a reactor, where certain constituents are partially transmuted into heavier isotopes via neutron capture or into lighter isotopes via fission. Fast reactors are required to complete the transmutation of long-lived isotopes. Closed fuel cycles are encompassed by the Department of Energy?s Advanced Fuel Cycle Initiative (AFCI), to which basic scientific research can contribute. Two nuclear reactor system architectures can meet the AFCI objectives: a ?single-tier? system or a ?dual-tier? system. Both begin with light water reactors and incorporate fast reactors. The ?dual-tier? systems transmute some plutonium and neptunium in light water reactors and all remaining transuranic elements (TRUs) in a closed-cycle fast reactor. Basic science initiatives are needed in two broad areas: ? Near-term impacts that can enhance the development of either ?single-tier? or ?dual-tier? AFCI systems, primarily within the next 20 years, through basic research. Examples: Dissolution of spent fuel, separations of elements for TRU recycling and transmutation Design, synthesis, and testing of inert matrix nuclear fuels and non-oxide fuels Invention and development of accurate on-line monitoring systems for chemical and nuclear species in the nuclear

  13. Advanced reprocessing developments in Europe contribution of European projects ACSEPT and ACTINET-I3

    SciTech Connect

    Bourg, S.; Poinssot, C.; Geist, A.; Cassayre, L.; Rhodes, C.; Ekberg, C.

    2012-07-01

    Nuclear energy has more than ever to demonstrate that it can contribute safely and on a sustainable way to answer the international increase in energy needs. Actually, in addition to an increased safety of the reactors themselves, its acceptance is still closely associated to our capability to reduce the lifetime of the nuclear waste, to manage them safely and to propose options for a better use of the natural resources. Spent fuel reprocessing can help to reach these objectives. But this cannot be achieved only by optimizing industrial processes through engineering studies. It is of a primary importance to increase our fundamental knowledge in actinide sciences in order to build the future of nuclear energy on reliable and scientifically-founded results, and therefore meet the needs of the future fuel cycles in terms of fabrication and performance of fuels, reprocessing and waste management. At the European level, both the collaborative project ACSEPT and the Integrated Infrastructure Initiative ACTINET-I3 work together to improve our knowledge in actinides chemistry and therefore develop advanced separation processes. These tools are complementary and work in close connection on some specific issues such as the understanding of the selectivity of extracting organic ligands. By offering trans-national access to the main nuclear research facility in Europe, ACTINET-I3 aims at increasing the knowledge in actinide sciences by gathering all the expertise available in European nuclear research institutes or university and giving them the opportunity to come and work in hot-labs (ITU, Atalante...) or beamlines (ESFR, ANKA, PSI) ACSEPT is focused on the development of advanced separation processes, both aqueous and pyrochemical. Head-end steps, fuel re-fabrication, solvent treatment, waste management are also taken into account. In aqueous process development, the SANEX and innovative SANEX flowsheets demonstration were successfully achieved. Chemical systems were

  14. Development of High-Temperature Transport Technologies of Molten Salt Slurry in Pyrometallurgical Reprocessing

    NASA Astrophysics Data System (ADS)

    Hijikata, Takatoshi; Koyama, Tadafumi

    Pyrometallurgical-reprocessing is one of the most promising technologies for advanced fuel cycle with favorable economic potential and intrinsic proliferation resistance. The development of transport technology for molten salt is a key issue in the industrialization of pyro-reprocessing. As for pure molten LiCl-KCl eutectic salt at approximately 773 K, we have already reported the successful results of transport using gravity and a centrifugal pump. However, molten salt in an electrorefiner mixes with insoluble fines when spent fuel is dissolved in porous anode basket. The insoluble consists of noble metal fission products, such as Pd, Ru, Mo, and Zr. There have been very few transport studies of a molten salt slurry (metal fines-molten salt mixture). Hence, transport experiments on a molten salt slurry were carried out to investigate the behavior of the slurry in a tube. The apparatus used in the transport experiments on the molten salt slurry consisted of a supply tank, a 10° inclined transport tube (10 mm inner diameter), a valve, a filter, and a recovery tank. Stainless steel (SS) fines with diameters from 53 to 415 μm were used. To disperse these fines homogenously, the molten salt and fines were stirred in the supply tank by an impeller at speeds from 1200 to 2100 rpm. The molten salt slurry containing 0.04 to 0.4 vol.% SS fines was transported from the supply tank to the recovery tank through the transportation tube. In the recovery tank, the fines were separated from the molten salt by the filter to measure the transport behavior of molten salt and SS fines. When the velocity of the slurry was 0.02 m/s, only 1% of the fines were transported to the recovery tank. On the other hand, most of the fines were transported when the velocity of the slurry was more than 0.8 m/s. Consequently, the molten salt slurry can be transported when the velocity is more than 0.8 m/s.

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

    SciTech Connect

    Bryan, Samuel A.; Levitskaia, Tatiana G.; Casella, Amanda J.; Peterson, James M.; Lines, Amanda M.; Jordan, Elizabeth A.; Verdugo, Dawn E.; Skomurski, Frances N.

    2011-07-19

    There is a renewed interest worldwide to promote the use of nuclear power and close the nuclear fuel cycle. The long term successful use of nuclear power is critically dependent upon adequate and safe processing and disposition of the spent nuclear fuel Liquid-liquid extraction is a separation technique commonly employed for the processing of the dissolved spent nuclear fuel. Our approach is based on prerequisite that real time monitoring of the solvent extraction flowsheets provides unique capability to quickly detect unwanted manipulations with fissile isotopes present in the radiochemical streams during reprocessing activities. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. In addition, the ability for continuous on-line monitoring allows for numerous benefits. Our team experimentally assessed the potential of Raman and vis-NIR spectrophotometric techniques for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(IV)/Np(V)/Nd(III), respectively, in solutions relevant to spent fuel reprocessing. Both techniques demonstrated robust performance in the repetitive batch measurements of each analyte in a wide concentration range using simulant and commercial dissolved spent fuel solutions. Static spectroscopic measurements served as training sets for the multivariate data analysis to obtain partial least squares predictive models, which were validated using on-line centrifugal contactor extraction tests. The corresponding spectrometers used under the laboratory conditions are easily convertible to the process-friendly configurations allowing remote measurements under the flow conditions. A fiber optic Raman probe allows monitoring of the high concentration species encountered in both aqueous and organic phases within the PUREX suite of flowsheets, including metal oxide ions, such as

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

    SciTech Connect

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

    2014-04-01

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

  17. Special issue on the "Consortium for Advanced Simulation of Light Water Reactors Research and Development Progress"

    NASA Astrophysics Data System (ADS)

    Turinsky, Paul J.; Martin, William R.

    2017-04-01

    In this special issue of the Journal of Computational Physics, the research and development completed at the time of manuscript submission by the Consortium for Advanced Simulation of Light Water Reactors (CASL) is presented. CASL is the first of several Energy Innovation Hubs that have been created by the Department of Energy. The Hubs are modeled after the strong scientific management characteristics of the Manhattan Project and AT&T Bell Laboratories, and function as integrated research centers that combine basic and applied research with engineering to accelerate scientific discovery that addresses critical energy issues. Lifetime of a Hub is expected to be five or ten years depending upon performance, with CASL being granted a ten year lifetime.

  18. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    SciTech Connect

    Budd, W.A.

    1986-03-01

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  19. Neutron collar calibration for assay of LWR (light-water reactor) fuel assemblies

    SciTech Connect

    Menlove, H.O.; Pieper, J.E.

    1987-03-01

    The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the /sup 235/U content, and the /sup 238/U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities.

  20. Qualification of a Tritium-Producing Target for the light water reactor application

    SciTech Connect

    Apley, W.J.; Beeman, G.H.; Ethridge, J.L.

    1992-06-01

    The Pacific Northwest Laboratory (PNL) currently manages the Light Water Reactor (LWR) Tritium Target Development Project (TTDP) for the Office of New Production Reactors (NP), US Department of Energy. The project`s objective is to demonstrate and qualify a high-temperature LWR tritium target system with fabrication and extraction processes sufficiently confirmed to ensure a deployable system consistent with variable tritium production demands. The project has also examined and reported on technical and institutional issues associated with acquisition and conversion of the yet uncompleted Washington Public Power Supply System Unit I (WNP-1) for tritium production purposes. WNP-1 is a 63% complete, Babcock and Wilcox, 3800 MW thermal, 205 assembly, pressurized water reactor located at Hanford, Washington. WNP-1 has served as the reference LWR plant for the technical evaluation and target development activities. This report discussed the evaluation and development necessary to provide a complete LWR target qualification package.

  1. Categorization of failed and damaged spent LWR (light-water reactor) fuel currently in storage

    SciTech Connect

    Bailey, W.J.

    1987-11-01

    The results of a study that was jointly sponsored by the US Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to (1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, (2) categorize those fuel assemblies, and (3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies. 78 refs., 23 figs., 19 tabs.

  2. Report from the Light Water Reactor Sustainability Workshop on On-Line Monitoring Technologies

    SciTech Connect

    Thomas Baldwin; Magdy Tawfik; Leonard Bond

    2010-06-01

    In support of expanding the use of nuclear power, interest is growing in methods of determining the feasibility of longer term operation for the U.S. fleet of nuclear power plants, particularly operation beyond 60 years. To help establish the scientific and technical basis for such longer term operation, the DOE-NE has established a research and development (R&D) objective. This objective seeks to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors. The Light Water Reactor Sustainability (LWRS) Program, which addresses the needs of this objective, is being developed in collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of nuclear power plants. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. In moving to identify priorities and plan activities, the Light Water Reactor Sustainability Workshop on On-Line Monitoring (OLM) Technologies was held June 10–12, 2010, in Seattle, Washington. The workshop was run to enable industry stakeholders and researchers to identify the nuclear industry needs in the areas of future OLM technologies and corresponding technology gaps and research capabilities. It also sought to identify approaches for collaboration that would be able to bridge or fill the technology gaps. This report is the meeting proceedings, documenting the presentations and discussions of the workshop and is intended to serve as a basis for a plan which is under development that will enable the I&C research pathway to achieve its goals. Benefits to the nuclear industry accruing from On Line Monitoring Technology cannot be ignored. Information gathered thus far has contributed significantly to the Department of Energy’s Light Water Reactor Sustainability Program. DOE has

  3. A flammability and combustion model for integrated accident analysis. [Advanced light water reactors

    SciTech Connect

    Plys, M.G.; Astleford, R.D.; Epstein, M. )

    1988-01-01

    A model for flammability characteristics and combustion of hydrogen and carbon monoxide mixtures is presented for application to severe accident analysis of Advanced Light Water Reactors (ALWR's). Flammability of general mixtures for thermodynamic conditions anticipated during a severe accident is quantified with a new correlation technique applied to data for several fuel and inertant mixtures and using accepted methods for combining these data. Combustion behavior is quantified by a mechanistic model consisting of a continuity and momentum balance for the burned gases, and considering an uncertainty parameter to match the idealized process to experiment. Benchmarks against experiment demonstrate the validity of this approach for a single recommended value of the flame flux multiplier parameter. The models presented here are equally applicable to analysis of current LWR's. 21 refs., 16 figs., 6 tabs.

  4. Recent performance experience with US light water reactor self-actuating safety and relief valves

    SciTech Connect

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  5. Overview of the U.S. DOE Light Water Reactor Sustainability Program

    SciTech Connect

    Shannong M. Bragg-Sitton; Jeremy T. Busby; Bruce P. Hallbert; Kathryn A. McCarthy; Richard Reister; Curtis L. Smith; Donald L. Williams

    2013-05-01

    The U.S. Department of Energy's Light Water Reactor Sustainability (LWRS) Program focuses on re­search and development to support the long-term operation of the nation's com­ mercial nuclear power plants. Extending the operation of current plants is essential to re­ alizing the administration's goals of reduc­ inggreenhouse gas emissions to 80 percent below 1990 levels by the year 2050. The science-based technical results from the LWRS Program provide data to help own­ ers make informed decisions on long-term operation and subsequent license renewal (the Nuclear Regulatory Commission's term for a second license renewal), thereby re­ ducing the uncertainty, and therefore the risk, associated with those decisions.

  6. Passive and inherent safety technologies for light-water nuclear reactors

    SciTech Connect

    Forsberg, C.W.

    1990-07-01

    Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs.

  7. REACTOR PRESSURE VESSEL ISSUES FOR THE LIGHT-WATER REACTOR SUSTAINABILITY PROGRAM

    SciTech Connect

    Nanstad, Randy K; Odette, George Robert

    2010-01-01

    The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable operation to at least 80 years. The reactor pressure vessel (RPV) is one of the primary components requiring significant research to enable such long-term operation. There are significant issues that need to be addressed to reduce the uncertainties in regulatory application, such as, 1) high neutron fluence/long irradiation times, and flux effects, 2) material variability, 3) high-nickel materials, 4)specimen size effects and the fracture toughness master curve, etc. The first issue is the highest priority to obtain the data and mechanistic understanding to enable accurate, reliable embrittlement predictions at high fluences. This paper discusses the major issues associated with long-time operation of existing RPVs and the LWRSP plans to address those issues.

  8. Design and optimization of a back-flow limiter for the high performance light water reactor

    SciTech Connect

    Fischer, Kai; Laurien, Eckart; Claas, Andreas G.; Schulenberg, Thomas

    2007-07-01

    Design and Analysis of a back-flow limiter are presented, which is implemented as a safety device in the four inlet lines of the Reactor Pressure Vessel (RPV) of the High Performance Light Water Reactor (HPLWR). As a passive component, the back-flow limiter has no moving parts and belongs to the group of fluid diodes. It has low flow resistance for regular operation condition and a high flow resistance when the flow direction is reversed which is the case if a break of the feedwater line occurs. The increased flow resistance is due to a substantially increased swirl for reverse flow condition. The design is optimized employing 1D flow analyses in combination with 3D CFD analyses with respect to geometrical modifications, like the nozzle shape and swirler angles. (authors)

  9. Accuracy Based Generation of Thermodynamic Properties for Light Water in RELAP5-3D

    SciTech Connect

    Cliff B. Davis

    2010-09-01

    RELAP5-3D interpolates to obtain thermodynamic properties for use in its internal calculations. The accuracy of the interpolation was determined for the original steam tables currently used by the code. This accuracy evaluation showed that the original steam tables are generally detailed enough to allow reasonably accurate interpolations in most areas needed for typical analyses of nuclear reactors cooled by light water. However, there were some regions in which the original steam tables were judged to not provide acceptable accurate results. Revised steam tables were created that used a finer thermodynamic mesh between 4 and 21 MPa and 530 and 640 K. The revised steam tables solved most of the problems observed with the original steam tables. The accuracies of the original and revised steam tables were compared throughout the thermodynamic grid.

  10. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    SciTech Connect

    Not Available

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  11. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    SciTech Connect

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  12. Expert assessments of the cost of light water small modular reactors.

    PubMed

    Abdulla, Ahmed; Azevedo, Inês Lima; Morgan, M Granger

    2013-06-11

    Analysts and decision makers frequently want estimates of the cost of technologies that have yet to be developed or deployed. Small modular reactors (SMRs), which could become part of a portfolio of carbon-free energy sources, are one such technology. Existing estimates of likely SMR costs rely on problematic top-down approaches or bottom-up assessments that are proprietary. When done properly, expert elicitations can complement these approaches. We developed detailed technical descriptions of two SMR designs and then conduced elicitation interviews in which we obtained probabilistic judgments from 16 experts who are involved in, or have access to, engineering-economic assessments of SMR projects. Here, we report estimates of the overnight cost and construction duration for five reactor-deployment scenarios that involve a large reactor and two light water SMRs. Consistent with the uncertainty introduced by past cost overruns and construction delays, median estimates of the cost of new large plants vary by more than a factor of 2.5. Expert judgments about likely SMR costs display an even wider range. Median estimates for a 45 megawatts-electric (MWe) SMR range from $4,000 to $16,300/kWe and from $3,200 to $7,100/kWe for a 225-MWe SMR. Sources of disagreement are highlighted, exposing the thought processes of experts involved with SMR design. There was consensus that SMRs could be built and brought online about 2 y faster than large reactors. Experts identify more affordable unit cost, factory fabrication, and shorter construction schedules as factors that may make light water SMRs economically viable.

  13. Expert assessments of the cost of light water small modular reactors

    PubMed Central

    Abdulla, Ahmed; Azevedo, Inês Lima; Morgan, M. Granger

    2013-01-01

    Analysts and decision makers frequently want estimates of the cost of technologies that have yet to be developed or deployed. Small modular reactors (SMRs), which could become part of a portfolio of carbon-free energy sources, are one such technology. Existing estimates of likely SMR costs rely on problematic top-down approaches or bottom-up assessments that are proprietary. When done properly, expert elicitations can complement these approaches. We developed detailed technical descriptions of two SMR designs and then conduced elicitation interviews in which we obtained probabilistic judgments from 16 experts who are involved in, or have access to, engineering-economic assessments of SMR projects. Here, we report estimates of the overnight cost and construction duration for five reactor-deployment scenarios that involve a large reactor and two light water SMRs. Consistent with the uncertainty introduced by past cost overruns and construction delays, median estimates of the cost of new large plants vary by more than a factor of 2.5. Expert judgments about likely SMR costs display an even wider range. Median estimates for a 45 megawatts-electric (MWe) SMR range from $4,000 to $16,300/kWe and from $3,200 to $7,100/kWe for a 225-MWe SMR. Sources of disagreement are highlighted, exposing the thought processes of experts involved with SMR design. There was consensus that SMRs could be built and brought online about 2 y faster than large reactors. Experts identify more affordable unit cost, factory fabrication, and shorter construction schedules as factors that may make light water SMRs economically viable. PMID:23716682

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

    SciTech Connect

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

    2011-10-01

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

  15. Status of radioiodine control for nuclear fuel reprocessing plants

    SciTech Connect

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

    1983-07-01

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

  16. Physiological correlates of eye movement desensitization and reprocessing.

    PubMed

    Elofsson, Ulf O E; von Schèele, Bo; Theorell, Töres; Söndergaard, Hans Peter

    2008-05-01

    Eye movement desensitization and reprocessing (EMDR) is an established treatment for post-traumatic stress disorder (PTSD). However, its working mechanism remains unclear. This study explored physiological correlates of eye movements during EMDR in relation to current hypotheses; distraction, conditioning, orienting response activation, and REM-like mechanisms. During EMDR therapy, fingertip temperature, heart rate, skin conductance, expiratory carbon dioxide level, and blood pulse oximeter oxygen saturation, were measured in male subjects with PTSD. The ratio between the low and high frequency components of the heart rate power spectrum (LF/HF) were computed as measures of autonomic balance. Respiratory rate was calculated from the carbon dioxide trace. Stimulation shifted the autonomic balance as indicated by decreases in heart rate, skin conductance and LF/HF-ratio, and an increased finger temperature. The breathing frequency and end-tidal carbon dioxide increased; oxygen saturation decreased during eye movements. In conclusion, eye movements during EMDR activate cholinergic and inhibit sympathetic systems. The reactivity has similarities with the pattern during REM-sleep.

  17. Higher-order ionosphere modeling for CODE's next reprocessing activities

    NASA Astrophysics Data System (ADS)

    Lutz, S.; Schaer, S.; Meindl, M.; Dach, R.; Steigenberger, P.

    2009-12-01

    CODE (the Center for Orbit Determination in Europe) is a joint venture between the Astronomical Institute of the University of Bern (AIUB, Bern, Switzerland), the Federal Office of Topography (swisstopo, Wabern, Switzerland), the Federal Agency for Cartography and Geodesy (BKG, Frankfurt am Main, Germany), and the Institut für Astronomische und Phsyikalische Geodäsie of the Technische Universität München (IAPG/TUM, Munich, Germany). It acts as one of the global analysis centers of the International GNSS Service (IGS) and participates in the first IGS reprocessing campaign, a full reanalysis of GPS data collected since 1994. For a future reanalyis of the IGS data it is planned to consider not only first-order but also higher-order ionosphere terms in the space geodetic observations. There are several works (e.g. Fritsche et al. 2005), which showed a significant and systematic influence of these effects on the analysis results. The development version of the Bernese Software used at CODE is expanded by the ability to assign additional (scaling) parameters to each considered higher-order ionosphere term. By this, each correction term can be switched on and off on normal-equation level and, moreover, the significance of each correction term may be verified on observation level for different ionosphere conditions.

  18. Evaluation of the Vienna APL corrections using reprocessed GNSS series

    NASA Astrophysics Data System (ADS)

    Steigenberger, P.; Dach, R.

    2011-12-01

    The Institute of Geodesy and Geophysics of the Vienna University of Technology recently started an operational service to provide non-tidal atmospheric pressure loading (APL) corrections. As the series is based on European Centre for Medium-Range Weather Forecasts (ECMWF) pressure data, it is fully consistent with the Vienna Mapping Function 1 (VMF1) atmospheric delay correction model for microwave measurements. Whereas VMF1 is widely used for, e.g., observations of Global Navigation Satellite Systems (GNSS), applying APL corrections is not yet a standard nowadays. The Center for Orbit Determination in Europe (CODE) - a joint venture between the Astronomical Institute of the University of Bern (AIUB, Bern, Switzerland), the Federal Office of Topography (swisstopo, Wabern, Switzerland), the Federal Office for Cartography and Geodesy (BKG, Frankfurt am Main, Germany), and the Insitute for Astronomical and Physical Geodesy, TU Muenchen (IAPG, Munich, Germany) - uses a recently generated series of reprocessed multi-GNSS data (considering GPS and GLONASS) to evaluate the APL corrections provided by the Vienna group. The results are also used to investigate the propagation of the APL effect in GNSS-derived results if no corrections are applied.

  19. Material balance areas and frequencies for large reprocessing plants

    SciTech Connect

    Burr, T.

    1994-08-01

    It has long been recognized that facilities with a large nuclear material throughput will probably not meet the International Atomic Energy Agency (IAEA) goal for detecting trickle diversion of plutonium over periods of about one year. The reason is that measurement errors for plutonium concentration and for liquid volume are often approximately relative over a fairly wide range of true values. Therefore, large throughput facilities will tend to have large uncertainties assigned to their annual throughput. By the same argument, if frequent balances are performed over small material balance areas, then the uncertainty associated with each balance period for each balance area will be small. However, trickle diversion would still be difficult to detect statistically. Because the IAEA will soon be faced with safeguarding a new large-scale reprocessing plant in Japan, it is timely to reconsider the advantages and disadvantages of performing frequent material balances over small balance areas (individual tanks where feasible). Therefore, in this paper the authors present some simulation results to study the effect of balance frequency on loss detection probability, and further simulation results to study possibilities introduced by choosing small balance areas. They conclude by recommending frequent balances over small areas.

  20. Authentication of reprocessing plant safeguards data through correlation analysis

    SciTech Connect

    Burr, T.L.; Wangen, L.E.; Mullen, M.F.

    1995-04-01

    This report investigates the feasibility and benefits of two new approaches to the analysis of safeguards data from reprocessing plants. Both approaches involve some level of plant modeling. All models involve some form of mass balance, either applied in the usual way that leads to material balances for individual process vessels at discrete times or applied by accounting for pipe flow rates that leads to material balances for individual process vessels at continuous times. In the first case, material balances are computed after each tank-to-tank transfer. In the second case, material balances can be computed at any desired time. The two approaches can be described as follows. The first approach considers the application of a new multivariate sequential test. The test statistic is a scalar, but the monitored residual is a vector. The second approach considers the application of recent nonlinear time series methods for the purpose of empirically building a model for the expected magnitude of a material balance or other scalar variable. Although the report restricts attention to monitoring scalar time series, the methodology can be extended to vector time series.

  1. Eye movement desensitization and reprocessing for adolescent depression.

    PubMed

    Bae, Hwallip; Kim, Daeho; Park, Yong Chon

    2008-03-01

    While cognitive behavior therapy is considered to be the first-line therapy for adolescent depression, there are limited data on whether other psychotherapeutic techniques are also effective in treating adolescents with depression. This report suggests the potential application of eye movement desensitization and reprocessing (EMDR) for treatment of depressive disorder related, not to trauma, but to stressful life events. At present, EMDR has only been empirically validated for only trauma-related disorders such as posttraumatic stress disorder. Two teenagers with major depressive disorder (MDD) underwent three and seven sessions of EMDR aimed at memories of stressful life events. After treatment, their depressive symptoms decreased to the level of full remission, and the therapeutic gains were maintained after two and three months of follow up. The effectiveness of EMDR for depression is explained by the model of adaptive information processing. Given the powerful effects observed within a brief period of time, the authors suggest that further investigation of EMDR for depressive disorders is warranted.

  2. Eye movement desensitization and reprocessing (EMDR): a meta-analysis.

    PubMed

    Davidson, P R; Parker, K C

    2001-04-01

    Eye movement desensitization and reprocessing (EMDR), a controversial treatment suggested for posttraumatic stress disorder (PTSD) and other conditions, was evaluated in a meta-analysis of 34 studies that examined EMDR with a variety of populations and measures. Process and outcome measures were examined separately. and EMDR showed an effect on both when compared with no treatment and with therapies not using exposure to anxiety-provoking stimuli and in pre post EMDR comparisons. However, no significant effect was found when EMDR was compared with other exposure techniques. No incremental effect of eye movements was noted when EMDR was compared with the same procedure without them. R. J. DeRubeis and P. Crits-Christoph (1998) noted that EMDR is a potentially effective treatment for noncombat PTSD. but studies that examined such patient groups did not give clear support to this. In sum, EMDR appears to be no more effective than other exposure techniques, and evidence suggests that the eye movements integral to the treatment, and to its name, are unnecessary.

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

    SciTech Connect

    1981-12-01

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

  4. Preliminary GPS orbit combination results of the IGS 2nd reprocessing campaign

    NASA Astrophysics Data System (ADS)

    Choi, Kevin

    2015-04-01

    International GNSS Service (IGS) has contributed to the International Terrestrial Reference Frame by reprocessing historic GPS network data and submitting Terrestrial Reference Frame solutions and Earth Rotation Parameters. For the 2nd reprocessing campaign, Analysis Centers (ACs) used up to 21 years of GPS observation data with daily integrations. IERS2010 conventions are applied to model the physical effects of the Earth. Total eight ACs have participated (7 Global solutions, and 2 Tide Gauge solutions) by reprocessing entire time series in a consistent way using the latest models and methodology. IGS combined daily SINEX TRF and EOP combinations have already been submitted to the IERS for ITRF2013. This presentation mainly focuses on the preliminary quality assessment of the reprocessed AC orbits. Quality of the orbit products are examined by examining the repeatability between daily AC satellite ephemeris. Power spectral analysis shows the background noise characteristics of each AC products, and its periodic behaviors.

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

    SciTech Connect

    Burch, W D

    1982-09-01

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

  6. Gigawatt-year nuclear-geothermal energy storage for light-water and high-temperature reactors

    SciTech Connect

    Forsberg, C. W.; Lee, Y.; Kulhanek, M.; Driscoll, M. J.

    2012-07-01

    Capital-intensive, low-operating cost nuclear plants are most economical when operated under base-load conditions. However, electricity demand varies on a daily, weekly, and seasonal basis. In deregulated utility markets this implies high prices for electricity at times of high electricity demand and low prices for electricity at times of low electricity demand. We examined coupling nuclear heat sources to geothermal heat storage systems to enable these power sources to meet hourly to seasonal variable electricity demand. At times of low electricity demand the reactor heats a fluid that is then injected a kilometer or more underground to heat rock to high temperatures. The fluid travels through the permeable-rock heat-storage zone, transfers heat to the rock, is returned to the surface to be reheated, and re-injected underground. At times of high electricity demand the cycle is reversed, heat is extracted, and the heat is used to power a geothermal power plant to produce intermediate or peak power. When coupling geothermal heat storage with light-water reactors (LWRs), pressurized water (<300 deg. C) is the preferred heat transfer fluid. When coupling geothermal heat storage with high temperature reactors at higher temperatures, supercritical carbon dioxide is the preferred heat transfer fluid. The non-ideal characteristics of supercritical carbon dioxide create the potential for efficient coupling with supercritical carbon dioxide power cycles. Underground rock cannot be insulated, thus small heat storage systems with high surface to volume ratios are not feasible because of excessive heat losses. The minimum heat storage capacity to enable seasonal storage is {approx}0.1 Gigawatt-year. Three technologies can create the required permeable rock: (1) hydro-fracture, (2) cave-block mining, and (3) selective rock dissolution. The economic assessments indicated a potentially competitive system for production of intermediate load electricity. The basis for a nuclear

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

    SciTech Connect

    Fox, K.; Brinkman, K.

    2011-09-22

    The Savannah River National Laboratory (SRNL) is developing crystalline ceramic waste forms to incorporate CS/LN/TM high Mo waste streams consisting of perovskite, hollandite, pyrochlore, zirconolite, and powellite phase assemblages. Simple raw materials, including Al{sub 2}O{sub 3}, CaO, and TiO{sub 2} were combined with simulated waste components to produce multiphase crystalline ceramics. Fiscal Year 2011 (FY11) activities included (i) expanding the compositional range by varying waste loading and fabrication of compositions rich in TiO{sub 2}, (ii) exploring the processing parameters of ceramics produced by the melt and crystallize process, (iii) synthesis and characterization of select individual phases of powellite and hollandite that are the target hosts for radionuclides of Mo, Cs, and Rb, and (iv) evaluating the durability and radiation stability of single and multi-phase ceramic waste forms. Two fabrication methods, including melting and crystallizing, and pressing and sintering, were used with the intent of studying phase evolution under various sintering conditions. An analysis of the XRD and SEM/EDS results indicates that the targeted crystalline phases of the FY11 compositions consisting of pyrochlore, perovskite, hollandite, zirconolite, and powellite were formed by both press and sinter and melt and crystallize processing methods. An evaluation of crystalline phase formation versus melt processing conditions revealed that hollandite, perovskite, zirconolite, and residual TiO{sub 2} phases formed regardless of cooling rate, demonstrating the robust nature of this process for crystalline phase development. The multiphase ceramic composition CSLNTM-06 demonstrated good resistance to proton beam irradiation. Electron irradiation studies on the single phase CaMoO{sub 4} (a component of the multiphase waste form) suggested that this material exhibits stability to 1000 years at anticipated self-irradiation doses (2 x 10{sup 10}-2 x 10{sup 11} Gy), but that its stability may be rate dependent, therefore limiting the activity of the waste for which it can be employed. Overall, these preliminary results indicate good radiation damage tolerance for the crystalline ceramic materials. The PCT results showed that, for all of the waste forms tested, the normalized release values for most of the elements measured, including all of the lanthanides and noble metals, were either very small or below the instrument detection limits. Elevated normalized release values were measured only for Cs, Mo, and Rb. It is difficult to draw further conclusions from these data until a benchmark material is developed for the PCT with this type of waste form. Calcined, simulated CS/LN/TM High Mo waste without additives had relatively low normalized release values for Cs, Mo, and Rb. A review of the chemical composition data for this sample showed that these elements were well retained after the calcination. Therefore, it will be useful to further characterize the calcined material to determine what form these elements are in after calcining. This, along with single phase studies on Cs containing crystal structures such as hollandite, should provide insight into the most ideal phases to incorporate these elements to produce a durable waste form.

  8. Endoscope reprocessing methods: a prospective study on the impact of human factors and automation.

    PubMed

    Ofstead, Cori L; Wetzler, Harry P; Snyder, Alycea K; Horton, Rebecca A

    2010-01-01

    The main cause of endoscopy-associated infections is failure to adhere to reprocessing guidelines. More information about factors impacting compliance is needed to support the development of effective interventions. The purpose of this multisite, observational study was to evaluate reprocessing practices, employee perceptions, and occupational health issues. Data were collected utilizing interviews, surveys, and direct observation. Written reprocessing policies and procedures were in place at all five sites, and employees affirmed the importance of most recommended steps. Nevertheless, observers documented guideline adherence, with only 1.4% of endoscopes reprocessed using manual cleaning methods with automated high-level disinfection versus 75.4% of those reprocessed using an automated endoscope cleaner and reprocessor. The majority reported health problems (i.e., pain, decreased flexibility, numbness, or tingling). Physical discomfort was associated with time spent reprocessing (p = .041). Discomfort diminished after installation of automated endoscope cleaners and reprocessors (p = .001). Enhanced training and accountability, combined with increased automation, may ensure guideline adherence and patient safety while improving employee satisfaction and health.

  9. X-Ray Reprocessing in Active Galactic Nuclei

    NASA Technical Reports Server (NTRS)

    Begelman, Mitchell C.

    2004-01-01

    This is the final report for research entitled "X-ray reprocessing in active galactic nuclei," into X-ray absorption and emission in various classes of active galaxy via X-ray spectral signatures. The fundamental goal of the research was to use these signatures as probes of the central engine structure and circumnuclear environment of active galactic nuclei. The most important accomplishment supported by this grant involved the detailed analysis and interpretation of the XMM data for the bright Seyfert 1 galaxy MCG-6-30-15. This work was performed by Drs. Christopher Reynolds and Mitchell Begelman in collaboration with Dr. Jorn Wilms (University of Tubingen, Germany; PI of the XMM observation) and other European scientists. With XMM we obtained medium resolution X-ray spectra of unprecedented quality for this Seyfert galaxy. Modeling the X-ray spectrum within the framework of accretion disk reflection models produced the first evidence for energy extraction from the spin of a black hole. Specifically, we found that the extreme gravitational redshifts required to explain the X-ray spectrum suggests that the bulk of the energy dissipation is concentrated very close to the black hole, in contrast with the expectations of any pure accretion disk model. In a second paper we addressed the low- energy spectral complexity and used RXTE specta to pin down the high-energy spectral index, thus firming up our initial interpretation. Additionally, we carried out detailed spectral and variability analyses of a number of Seyfert and radio galaxies (e.g., NGC 5548 and 3C 111) and developed general techniques that will be useful in performing X-ray reverberation mapping of accretion disks in AGN, once adequate data becomes available. A list of papers supported by this research is included.

  10. Contribution of TIGA reprocessing to the ITRF densification

    NASA Astrophysics Data System (ADS)

    Rudenko, S.; Dähnn, M.; Gendt, G.; Brandt, A.; Nischan, T.

    2009-04-01

    Analysis of tide gauge measurements with the purpose of sea level change investigations requires a well defined reference frame. Such reference frame can be realized through precise positions of GPS stations located at or near tide gauges (TIGA stations) and analyzed within the IGS GPS Tide Gauge Benchmark Monitoring Pilot Project (TIGA). To tie this reference frame to the International Terrestrial Reference Frame (ITRF), one should process simultaneously GPS data from TIGA and IGS stations included in the ITRF. A time series of GPS station positions has been recently derived by reprocessing GPS data from about 400 GPS stations globally distributed covering totally time span from 1998 till 2008 using EPOS-Potsdam software developed at GFZ and improved in the recent years. The analysis is based on the use of IERS Conventions 2003, ITRF2005 as a priori reference frame, FES2004 ocean tide loading model, absolute phase centre variations for GPS satellite transmit and ground receive antennae and other models. About 220 stations of the solution are IGS ones and about 180 are TIGA GPS stations that are not IGS ones. The solution includes weekly coordinates of GPS stations, daily values of the Earth rotation parameters and their rates, as well as satellite antenna offsets. On the other hand, our new solution can contribute to the ITRF densification by providing positions of about 200 stations being not present in ITRF2005. The solution can be also used for the integration of regional frames. The paper presents the results of the analysis and the comparison of our solution with ITRF2005 and the solutions of other TIGA and IGS Analysis Centres.

  11. Light Water Reactor Sustainability Program: Digital Technology Business Case Methodology Guide

    SciTech Connect

    Thomas, Ken; Lawrie, Sean; Hart, Adam; Vlahoplus, Chris

    2014-09-01

    The Department of Energy’s (DOE’s) Light Water Reactor Sustainability Program aims to develop and deploy technologies that will make the existing U.S. nuclear fleet more efficient and competitive. The program has developed a standard methodology for determining the impact of new technologies in order to assist nuclear power plant (NPP) operators in building sound business cases. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway is part of the DOE’s Light Water Reactor Sustainability (LWRS) Program. It conducts targeted research and development (R&D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation’s energy and environmental security. The II&C Pathway is conducting a series of pilot projects that enable the development and deployment of new II&C technologies in existing nuclear plants. Through the LWRS program, individual utilities and plants are able to participate in these projects or otherwise leverage the results of projects conducted at demonstration plants. Performance advantages of the new pilot project technologies are widely acknowledged, but it has proven difficult for utilities to derive business cases for justifying investment in these new capabilities. Lack of a business case is often cited by utilities as a barrier to pursuing wide-scale application of digital technologies to nuclear plant work activities. The decision to move forward with funding usually hinges on

  12. Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

    SciTech Connect

    Turinsky, Paul J.; Kothe, Douglas B.

    2016-05-15

    The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics “core simulator” based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M

  13. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    SciTech Connect

    Rebak, Raul B.

    2014-09-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  14. The potential impact of enhanced accident tolerant cladding materials on reactivity initiated accidents in light water reactors

    SciTech Connect

    Brown, Nicholas R.; Wysocki, Aaron J.; Terrani, Kurt A.; Xu, Kevin G.; Wachs, Daniel M.

    2017-01-01

    Here, advanced cladding materials with potentially enhanced accident tolerance will yield different light-water-reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to cladding material properties, reactor physics, thermal, and hydraulic characteristics. Differences in reactors physics characteristics are driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and also by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermal hydraulic limits after transition from the current zirconium alloy cladding to the advanced materials will also affect the transient response of the integral fuel. This paper describes three-dimensional nodal kinetics simulations of a reactivity-initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon-carbide (SiC-SiC)-based cladding materials. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus that of reference Zr cladding is predominantly due to differences in (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores resulting from hardened (or softened) spectrum. This study shows similar behavior for SiC-SiC-based cladding configurations on the transient response versus reference Zircaloy cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. This is due to the shorter neutron generation time of the models with FeCrAl cladding. Therefore, the FeCrAl-based cases have a more

  15. The potential impact of enhanced accident tolerant cladding materials on reactivity initiated accidents in light water reactors

    DOE PAGES

    Brown, Nicholas R.; Wysocki, Aaron J.; Terrani, Kurt A.; ...

    2017-01-01

    Here, advanced cladding materials with potentially enhanced accident tolerance will yield different light-water-reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to cladding material properties, reactor physics, thermal, and hydraulic characteristics. Differences in reactors physics characteristics are driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and also by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermal hydraulic limits after transition from the current zirconium alloy cladding to the advanced materials will also affect the transientmore » response of the integral fuel. This paper describes three-dimensional nodal kinetics simulations of a reactivity-initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon-carbide (SiC-SiC)-based cladding materials. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus that of reference Zr cladding is predominantly due to differences in (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores resulting from hardened (or softened) spectrum. This study shows similar behavior for SiC-SiC-based cladding configurations on the transient response versus reference Zircaloy cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. This is due to the shorter neutron generation time of the models with FeCrAl cladding. Therefore, the FeCrAl-based cases have

  16. Sensitivity Analysis and Optimization of the Nuclear Fuel Cycle: A Systematic Approach

    NASA Astrophysics Data System (ADS)

    Passerini, Stefano

    For decades, nuclear energy development was based on the expectation that recycling of the fissionable materials in the used fuel from today's light water reactors into advanced (fast) reactors would be implemented as soon as technically feasible in order to extend the nuclear fuel resources. More recently, arguments have been made for deployment of fast reactors in order to reduce the amount of higher actinides, hence the longevity of radioactivity, in the materials destined to a geologic repository. The cost of the fast reactors, together with concerns about the proliferation of the technology of extraction of plutonium from used LWR fuel as well as the large investments in construction of reprocessing facilities have been the basis for arguments to defer the introduction of recycling technologies in many countries including the US. In this thesis, the impacts of alternative reactor technologies on the fuel cycle are assessed. Additionally, metrics to characterize the fuel cycles and systematic approaches to using them to optimize the fuel cycle are presented. The fuel cycle options of the 2010 MIT fuel cycle study are re-examined in light of the expected slower rate of growth in nuclear energy today, using the CAFCA (Code for Advanced Fuel Cycle Analysis). The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycle options available in the future. The options include limited recycling in L WRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. Additional fuel cycle scenarios presented for the first time in this work assume the deployment of innovative recycling reactor technologies such as the Reduced Moderation Boiling Water Reactors and Uranium-235 initiated Fast Reactors. A sensitivity study focused on system and technology parameters of interest has been conducted to test

  17. Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Karlsen, T.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K.

    1994-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289{degree}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  18. THERMIT-2: a two-fluid model for light water reactor subchannel transient analysis. Final report

    SciTech Connect

    Kelly, J.E.; Kao, S.P.; Kazimi, M.S.

    1981-04-01

    The broad effort of developing and assessing the two fluid model computer code THERMIT for light water reactor (LWR) subchannel analysis is described. The developmental effort required a reformulation of the coolant-to-fuel rod coupling so that THERMIT is now capable of traditional coolant-centered subchannel analysis. A model that accounts for mass, momentum and energy transport between mesh cells due to turbulent mixing for two-phase conditions has also been introduced. This model is the first such attempt in a two-fluid context. The liquid-vapor interfacial exchange terms in the two-fluid model have been modified for improved accuracy. A systematic evaluation of the exchange models has been performed. The mass and momentum exchange rates between the vapor and the liquid for pre-CHF conditions were evaluated by comparison to void fraction data in over 30 one-dimensional steady-state experiments reported in the open literature. The liquid-vapor energy exchange rate for post-CHF conditions was assessed using 15 steady-state, one-dimensional wall temperature measurements. The mixing model was tested against G.E. and Ispra BWR and PWR rod-bundle measurements. Comparisons with these measurements have shown the appropriateness of this model.

  19. Analysis of assembly serial number usage in domestic light-water reactors

    SciTech Connect

    Reich, W.J. ); Moore, R.S. )

    1991-05-01

    Domestic light-water reactor (LWR) fuel assemblies are identified by a serial number that is placed on each assembly. These serial numbers are used as identifiers throughout the life of the fuel. The uniqueness of assembly serial numbers is important in determining their effectiveness as unambiguous identifiers. The purpose of this study is to determine what serial numbering schemes are used, the effectiveness of these schemes, and to quantify how many duplicate serial numbers occur on domestic LWR fuel assemblies. The serial numbering scheme adopted by the American National Standards Institute (ANSI) ensures uniqueness of assembly serial numbers. The latest numbering scheme adopted by General Electric (GE), was also found to be unique. Analysis of 70,971 fuel assembly serial numbers from permanently discharged fuel identified 11,948 serial number duplicates. Three duplicate serial numbers were found when analysis focused on duplication within the individual fuel inventory at each reactor site, but these were traced back to data entry errors and will be corrected by the Energy Information Administration (EIA). There were also three instances where the serial numbers used to identify assemblies used for hot cell studies differed from the serial numbers reported to the EIA. It is recommended that fuel fabricators and utilities adhere to the ANSI serial numbering scheme to ensure serial number uniqueness. In addition, organizations collecting serial number information, should request that all known serial numbers physically attached or associated with each assembly be reported and identified by the corresponding number scheme. 10 refs., 5 tabs.

  20. Ta3N5 nanotube arrays for visible light water photoelectrolysis.

    PubMed

    Feng, Xinjian; Latempa, Thomas J; Basham, James I; Mor, Gopal K; Varghese, Oomman K; Grimes, Craig A

    2010-03-10

    Tantalum nitride (Ta3N5) has a band gap of approximately 2.07 eV, suitable for collecting more than 45% of the incident solar spectrum energy. We describe a simple method for scale fabrication of highly oriented Ta3N5 nanotube array films, by anodization of tantalum foil to achieve vertically oriented tantalum oxide nanotube arrays followed by a 700 degrees C ammonia anneal for sample crystallization and nitridation. The thin walled amorphous nanotube array structure enables transformation from tantalum oxide to Ta3N5 to occur at relatively low temperatures, while high-temperature annealing related structural aggregation that commonly occurs in particle films is avoided. In 1 M KOH solution, under AM 1.5 illumination with 0.5 V dc bias typical sample (nanotube length approximately 240 nm, wall thickness approximately 7 nm) visible light incident photon conversion efficiencies (IPCE) as high as 5.3% were obtained. The enhanced visible light activity in combination with the ordered one-dimensional nanoarchitecture makes Ta3N5 nanotube arrays films a promising candidate for visible light water photoelectrolysis.

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

    SciTech Connect

    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.

  2. Qualification Requirements of Guided Ultrasonic Waves for Inspection of Piping in Light Water Reactors

    SciTech Connect

    Meyer, Ryan M.; Ramuhalli, Pradeep; Doctor, Steven R.; Bond, Leonard J.

    2013-08-01

    Guided ultrasonic waves (GUW) are being increasingly used for both NDT and monitoring of piping. GUW offers advantages over many conventional NDE technologies due to the ability to inspect large volumes of piping components without significant removal of thermal insulation or protective layers. In addition, regions rendered inaccessible to more conventional NDE technologies may be more accessible using GUW techniques. For these reasons, utilities are increasingly considering the use of GUWs for performing the inspection of piping components in nuclear power plants. GUW is a rapidly evolving technology and its usage for inspection of nuclear power plant components requires refinement and qualification to ensure it is able to achieve consistent and acceptable levels of performance. This paper will discuss potential requirements for qualification of GUW techniques for the inspection of piping components in light water reactors (LWRs). The Nuclear Regulatory Commission has adopted ASME Boiler and Pressure Vessel Code requirements in Sections V, III, and XI for nondestructive examination methods, fabrication inspections, and pre-service and in-service inspections. A Section V working group has been formed to place the methodology of GUW into the ASME Boiler and Pressure Vessel Code but no requirements for technique, equipment, or personnel exist in the Code at this time.

  3. Storage of LWR (light-water-reactor) spent fuel in air

    SciTech Connect

    Thomas, L.E.; Charlot, L.A.; Coleman, J.E. ); Knoll, R.W. )

    1989-12-01

    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to determine the oxidation response of light-water-reactor (LWR) spent fuels under conditions appropriate to fuel storage in air. The program is designed to investigate several independent variables that might affect the oxidation behavior of spent fuel. Included are temperature (135 to 230{degree}C), fuel burnup (to about 34 MWd/kgM), reactor type (pressurized and boiling water reactors), moisture level in the air, and the presence of a high gamma field. In continuing tests with declad spent fuel and nonirradiated UO{sub 2} specimens, oxidation rates were monitored by weight-gain measurements and the microstructures of subsamples taken during the weighing intervals were characterized by several analytical methods. The oxidation behavior indicated by weight gain and time to form powder will be reported in Volume III of this series. The characterization results obtained from x-ray diffractometry, transmission electron microscopy, scanning electron microscopy, and Auger electron spectrometry of oxidized fuel samples are presented in this report. 28 refs., 21 figs., 3 tabs.

  4. Passive ALWR requirements to prevent containment failure. [Advanced light water reactors

    SciTech Connect

    Additon, S.L.; Blanchard, D.P.; Leaver, D.E.; Persinko, D. TENERA, L.P., Bethesda, MD )

    1991-12-01

    The purpose of this report is to document a systematic evaluation of the Passive Advanced Light Water Reactor (ALWR) design requirements which address severe accident mitigation. This evaluation was performed concurrent with completion of the ALWR Requirements Document to assure the adequacy of these mitigation requirements. The passive plant approach to containment integrity assurance reflects an expansion of the approach established earlier for evolutionary ALWRs. The report identifies containment challenges that might occur coincident with or result from a core damage event, compiles the set of passive ALWR design requirements which addresses each challenge, and evaluates each set of requirements on an integrated basis to confirm that the requirements provide substantial assurance that coincident core damage and containment failure are precluded. Based on past PRAs, a review of pertinent safety functions, severe accident analyses, current regulatory requirements, and reviews by ALWR design personnel, twenty-three (23) potential containment challenges were identified. The report concludes that the relevant ALWR requirements severe to limit the likelihood and magnitude of the challenges, and to assure the capability of the containment to accommodate all challenges which remain potentially risk-significant.

  5. Microstructural examination of fatigue accumulation in critical LWR (light water reactor) components: Final report

    SciTech Connect

    Allen, A.J.; Buttle, D.J.; Coleman, C.F.; Smith, F.A.; Smith, R.L.

    1988-01-01

    This report describes a morphological study of the feasibility of measuring the fatigue damage accumulation state of critical light water reactor (LWR) components by microstructural examination. The changes in microstructure associated with fatigue processes are first discussed so that relevant NDE measurement parameters can be identified. (The creep regime is not considered in this report). The candidate NDE techniques are then reviewed in detail under the following headings: positron annihilation, x-ray diffraction, magnetic techniques, the magnetic Barkhausen effect, the magneto acoustic technique, acoustic emission, ultrasonic techniques and finally other miscellaneous techniques applicable to fatigue damage assessment. All the feasible techniques are summarised and rated in a set of comparison tables. A possible programme for the immediate development of the positron annihilation lineshape technique is proposed. It is concluded that the most successful method of measuring the fatigue accumulation in LWR critical components in a way which relates to the intent of the ASME pressure vessel codes, is likely to be the use of several techniques together and the cross-relation of the results obtained by each. Five techniques are highlighted for immediate possible development: 'etching and surface replication', 'positron annihilation lineshapes', 'x-ray diffraction residual stress', 'acoustic emission' and 'ultrasonic surface acoustic waves'.

  6. An innovative fuel design concept for improved light water reactor performance and safety. Final technical report

    SciTech Connect

    Tulenko, J.S.; Connell, R.G.

    1995-07-01

    Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. The purpose of this research was to explore a technique for extending fuel performance by thermally bonding LWR fuel with a non-alkaline liquid metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide (UO{sub 2}) fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Due to the thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high conductivity liquid metal thermally bonds the fuel to the cladding, and eliminates the large temperature change across the gap, while preserving the expansion and pellet loading capabilities. The resultant lower fuel temperature directly impacts fuel performance limit margins and also core transient performance. The application of liquid bonding techniques to LWR fuel was explored for the purposes of increasing LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) has been developed under the program to analyze the in-reactor performance of the liquid metal bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of Liquid Metal Bonded LWR fuel.

  7. Computational Neutronics Methods and Transmutation Performance Analyses for Light Water Reactors

    SciTech Connect

    M. Asgari; B. Forget; S. Piet; R. Ferrer; S. Bays

    2007-03-01

    The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One obvious path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multi-pass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed. These analyses provide comparison of direct thermal recycling of PuNpAmCm, PuNpAm, PuNp, and Pu.

  8. Insights for aging management of light water reactor components: Metal containments. Volume 5

    SciTech Connect

    Shah, V.N.; Sinha, U.P.; Smith, S.K.

    1994-03-01

    This report evaluates the available technical information and field experience related to management of aging damage to light water reactor metal containments. A generic aging management approach is suggested for the effective and comprehensive aging management of metal containments to ensure their safe operation. The major concern is corrosion of the embedded portion of the containment vessel and detection of this damage. The electromagnetic acoustic transducer and half-cell potential measurement are potential techniques to detect corrosion damage in the embedded portion of the containment vessel. Other corrosion-related concerns include inspection of corrosion damage on the inaccessible side of BWR Mark I and Mark II containment vessels and corrosion of the BWR Mark I torus and emergency core cooling system piping that penetrates the torus, and transgranular stress corrosion cracking of the penetration bellows. Fatigue-related concerns include reduction in the fatigue life (a) of a vessel caused by roughness of the corroded vessel surface and (b) of bellows because of any physical damage. Maintenance of surface coatings and sealant at the metal-concrete interface is the best protection against corrosion of the vessel.

  9. Primary disassembly of Light Water Breeder Reactor modules for core evaluation (LWBR Development Program)

    SciTech Connect

    Greenberger, R.J.; Miller, E.L.

    1987-10-01

    After successfully operating for 29,047 effective full power hours, the Light Water Breeder Reactor (LWBR) core was defueled prior to total decommissioning of the Shippingport Atomic Power Station. All nuclear fuel and much of the reactor internal hardware was removed from the reactor vessel. Non-fuel components were prepared for shipment to disposal sites, and the fuel assemblies were partially disassembled and shipped to the Expended Core Facility (ECF) in Idaho. At ECF, the fuel modules underwent further disassembly to provide fuel rods for nondestructive testing to establish the core's breeding efficiency and to provide core components for examinations to assess their performance characteristics. This report presents a basic description of the processes and equipment used to disassemble LWBR fuel modules for subsequent proof-of-breeding (POB) and core examination operations. Included are discussions of module handling fixtures and equipment, the underwater milling machine and bandsaw assemblies, and the associated design and operation of this equipment for LWBR fuel module disassembly.

  10. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    SciTech Connect

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  11. Swelling in light water reactor internal components: Insights from computational modeling

    SciTech Connect

    Stoller, Roger E.; Barashev, Alexander V.; Golubov, Stanislav I.

    2015-08-01

    A modern cluster dynamics model has been used to investigate the materials and irradiation parameters that control microstructural evolution under the relatively low-temperature exposure conditions that are representative of the operating environment for in-core light water reactor components. The focus is on components fabricated from austenitic stainless steel. The model accounts for the synergistic interaction between radiation-produced vacancies and the helium that is produced by nuclear transmutation reactions. Cavity nucleation rates are shown to be relatively high in this temperature regime (275 to 325°C), but are sensitive to assumptions about the fine scale microstructure produced under low-temperature irradiation. The cavity nucleation rates observed run counter to the expectation that void swelling would not occur under these conditions. This expectation was based on previous research on void swelling in austenitic steels in fast reactors. This misleading impression arose primarily from an absence of relevant data. The results of the computational modeling are generally consistent with recent data obtained by examining ex-service components. However, it has been shown that the sensitivity of the model s predictions of low-temperature swelling behavior to assumptions about the primary damage source term and specification of the mean-field sink strengths is somewhat greater that that observed at higher temperatures. Further assessment of the mathematical model is underway to meet the long-term objective of this research, which is to provide a predictive model of void swelling at relevant lifetime exposures to support extended reactor operations.

  12. Development plan for the External Hazards Experimental Group. Light Water Reactor Sustainability Program

    SciTech Connect

    Coleman, Justin Leigh; Smith, Curtis Lee; Burns, Douglas Edward; Kammerer, Annie

    2016-03-01

    This report describes the development plan for a new multi-partner External Hazards Experimental Group (EHEG) coordinated by Idaho National Laboratory (INL) within the Risk-Informed Safety Margin Characterization (RISMC) technical pathway of the Light Water Reactor Sustainability Program. Currently, there is limited data available for development and validation of the tools and methods being developed in the RISMC Toolkit. The EHEG is being developed to obtain high-quality, small- and large-scale experimental data validation of RISMC tools and methods in a timely and cost-effective way. The group of universities and national laboratories that will eventually form the EHEG (which is ultimately expected to include both the initial participants and other universities and national laboratories that have been identified) have the expertise and experimental capabilities needed to both obtain and compile existing data archives and perform additional seismic and flooding experiments. The data developed by EHEG will be stored in databases for use within RISMC. These databases will be used to validate the advanced external hazard tools and methods.

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

    SciTech Connect

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

    2013-10-01

    promises as a replacement for AHA. FHA undergoes hydrolysis to formic acid which is volatile, thus allowing the recycling of nitric acid. Unfortunately, FHA powder was not stable in the experiments we ran in our laboratory. In addition, AHA and FHA also decompose to hydroxylamine which may undergo an autocatalytic reaction. Other reductants are available and could be extremely useful for actinides separation. The review presents the current plutonium reductants used in used nuclear fuel reprocessing and will introduce innovative and novel reductants that could become reducers for future research on UNF separation.

  14. An homogeneously reprocessed Zenith Total Delay long-term time series over Europe

    NASA Astrophysics Data System (ADS)

    Pacione, Rosa; Pace, Brigida; Bianco, Giuseppe

    2014-05-01

    Homogeneously reprocessed observations from permanent GNSS stations have high potential for monitoring trends and variability in atmospheric water vapour which will enable evaluation of systematic biases from several instruments, improve the knowledge of climatic trends of atmospheric water vapour and be useful for global and regional NWP reanalyses and climate model simulations. The present availability of more than 15 years of GNSS data belonging to the European Permanent Network (EPN, http://www.epncb.oma.be/) is a valuable database for the development of a climate data record of GNSS tropospheric products. We are homogeneously reprocessing the whole EPN network for the period 1996-2013. GNSS data are analyzed with GIPSY-OASIS II 6.2 in PPP mode applying the state-of-the-art models and the JPL reprocessed IGS08-products. These reprocessed ZTD time series over Europe will be compared with radiosonde data, VLBI and IGS zenith delays for collocated stations. The ongoing reprocessing efforts is part of the EPN Repro2 initiative and will provide a GNSS climate data record for the WG3 'Use of GNSS tropospheric products for climate monitoring' of the COST Action ES1206 'Advanced Global Navigation Satellite Systems tropospheric products for monitoring severe weather events and climate (GNSS4SWEC)'.

  15. Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

    SciTech Connect

    Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K.

    1995-03-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials.

  16. Light Water Reactor Sustainability Program A Reference Plan for Control Room Modernization: Planning and Analysis Phase

    SciTech Connect

    Jacques Hugo; Ronald Boring; Lew Hanes; Kenneth Thomas

    2013-09-01

    The U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) program is collaborating with a U.S. nuclear utility to bring about a systematic fleet-wide control room modernization. To facilitate this upgrade, a new distributed control system (DCS) is being introduced into the control rooms of these plants. The DCS will upgrade the legacy plant process computer and emergency response facility information system. In addition, the DCS will replace an existing analog turbine control system with a display-based system. With technology upgrades comes the opportunity to improve the overall human-system interaction between the operators and the control room. To optimize operator performance, the LWRS Control Room Modernization research team followed a human-centered approach published by the U.S. Nuclear Regulatory Commission. NUREG-0711, Rev. 3, Human Factors Engineering Program Review Model (O’Hara et al., 2012), prescribes four phases for human factors engineering. This report provides examples of the first phase, Planning and Analysis. The three elements of Planning and Analysis in NUREG-0711 that are most crucial to initiating control room upgrades are: • Operating Experience Review: Identifies opportunities for improvement in the existing system and provides lessons learned from implemented systems. • Function Analysis and Allocation: Identifies which functions at the plant may be optimally handled by the DCS vs. the operators. • Task Analysis: Identifies how tasks might be optimized for the operators. Each of these elements is covered in a separate chapter. Examples are drawn from workshops with reactor operators that were conducted at the LWRS Human System Simulation Laboratory HSSL and at the respective plants. The findings in this report represent generalized accounts of more detailed proprietary reports produced for the utility for each plant. The goal of this LWRS report is to disseminate the technique and provide examples sufficient to

  17. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-08-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  18. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-01-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  19. Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

    NASA Astrophysics Data System (ADS)

    Turinsky, Paul J.; Kothe, Douglas B.

    2016-05-01

    The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics ;core simulator; based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M

  20. Weapons-grade plutonium dispositioning. Volume 4. Plutonium dispositioning in light water reactors

    SciTech Connect

    Sterbentz, J.W.; Olsen, C.S.; Sinha, U.P.

    1993-06-01

    This study is in response to a request by the Reactor Panel Subcommittee of the National Academy of Sciences (NAS) Committee on International Security and Arms Control (CISAC) to evaluate the feasibility of using plutonium fuels (without uranium) for disposal in existing conventional or advanced light water reactor (LWR) designs and in low temperature/pressure LWR designs that might be developed for plutonium disposal. Three plutonium-based fuel forms (oxides, aluminum metallics, and carbides) are evaluated for neutronic performance, fabrication technology, and material and compatibility issues. For the carbides, only the fabrication technologies are addressed. Viable plutonium oxide fuels for conventional or advanced LWRs include plutonium-zirconium-calcium oxide (PuO{sub 2}-ZrO{sub 2}-CaO) with the addition of thorium oxide (ThO{sub 2}) or a burnable poison such as erbium oxide (Er{sub 2}O{sub 3}) or europium oxide (Eu{sub 2}O{sub 3}) to achieve acceptable neutronic performance. Thorium will breed fissile uranium that may be unacceptable from a proliferation standpoint. Fabrication of uranium and mixed uranium-plutonium oxide fuels is well established; however, fabrication of plutonium-based oxide fuels will require further development. Viable aluminum-plutonium metallic fuels for a low temperature/pressure LWR include plutonium aluminide in an aluminum matrix (PuAl{sub 4}-Al) with the addition of a burnable poison such as erbium (Er) or europium (Eu). Fabrication of low-enriched plutonium in aluminum-plutonium metallic fuel rods was initially established 30 years ago and will require development to recapture and adapt the technology to meet current environmental and safety regulations. Fabrication of high-enriched uranium plate fuel by the picture-frame process is a well established process, but the use of plutonium would require the process to be upgraded in the United States to conform with current regulations and minimize the waste streams.

  1. Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009–2013

    SciTech Connect

    Idaho National Laboratory

    2009-12-01

    Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement

  2. Radioactive Fission Product Release from Defective Light Water Reactor Fuel Elements

    SciTech Connect

    Konyashov, Vadim V.; Krasnov, Alexander M.

    2002-04-15

    Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding gap of the defective fuel element into coolant was developed. It takes into account the convective transport in the fuel-to-cladding gap and RFP sorption on the internal cladding surface and is in good agreement with the available experimental data. An approximate analytical solution of the transport equation is given. The calculation dependencies of the RFP release coefficients on the main parameters such as defect size, fuel-to-cladding gap, temperature of the internal cladding surface, and radioactive decay constant were analyzed.It is shown that the change of the RFP release from the fuel elements with the initial defects during long-term irradiation is, mainly, caused by fuel swelling followed by reduction of the fuel-to-cladding gap and the fuel temperature. The calculation model for the RFP release from defective fuel elements applicable to light water reactors (LWRs) was developed. It takes into account the change of the defective fuel element parameters during long-term irradiation. The calculation error according to the program does not exceed 30% over all the linear power change range of the LWR fuel elements (from 5 to 26 kW/m)

  3. Aging of the containment pressure boundary in light-water reactor plants

    SciTech Connect

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

    1997-01-01

    Research is being conducted by the Oak Ridge National Laboratory to address aging of the containment pressure boundary in light-water reactor plants. The objectives of this work are to (1) identify the significant factors related to occurrence of corrosion, efficacy of inspection, and structural capacity reduction of steel containments and liners of concrete containments, and to make recommendations on use of risk models in regulatory decisions; (2) provide NRC reviewers a means of establishing current structural capacity margins for steel containments, and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by NRC reviewers in assessing the seriousness of reported incidences of containment degradation. In meeting these objectives research is being conducted in two primary task areas - pressure boundary condition assessment and root-cause resolution practices, and reliability-based condition assessments. Under the first task area a degradation assessment methodology was developed for use in characterizing the in-service condition of metal and concrete containment pressure boundary components and quantifying the amount of damage that is present. An assessment of available destructive and nondestructive techniques for examining steel containments and liners is ongoing. Under the second task area quantitative structural reliability analysis methods are being developed for application to degraded metallic pressure boundaries to provide assurances that they will be able to withstand future extreme loads during the desired service period with a level of reliability that is sufficient for public safety. To date, mathematical models that describe time-dependent changes in steel due to aggressive environmental factors have been identified, and statistical data supporting their use in time-dependent reliability analysis have been summarized.

  4. Survey of light-water-reactor designs to be offered in the United States

    SciTech Connect

    Spiewak, I.

    1986-03-01

    ORNL has conducted a Nuclear Power Options Viability Study for the Department of Energy. That study is primarily concerned with new technology which could be developed for initial operation in the 2000 to 2010 time frame. Such technology would have to compete not only with coal options but with incrementally improved commercial light-water-reactors. This survey reported here was undertaken to gain an understanding of the nuclear commercial technology likely to be offered in the late 1980s and perhaps beyond. The three US vendors actively marketing NSSSs are each developing a product for the future which they expect to be more reliable, more maintainable, more economical, and safer than the present plants. These are all essentially 3800-MW(t) designs, although all are studying smaller plants. They apparently will be off offered as standard prelicensed designs with much larger scope than earlier NSSS offerings, with the possibility of firm prices. Westinghouse with Mitsubishi Heavy Industries is developing a completely new design (APWR) to be built initially in Japan, hopefully for operation by the mid-1990s. Westinghouse is making a strong effort to have the APWR licensed in the US as a standard plant. Combustion Engineering (C-E) is evaluating potential improvements to the System-80 standard design (CESSAR) that has already received final design approval by the NRC. General Electric (GE), with Hitachi and Toshiba, is developing a new design (ABWR) that incorporates advanced features which have been proven by the worldwide BWR suppliers. The ABWR is to be built initially in Japan, but the design could be adapted to the United States. Westinghouse, C-E, and GE have done some conceptual evaluation of reactors in the 600-MW(e) class. The Westinghouse concept is a two-loop plant intended for factory assembly in a shipyard and delivery to a site by barge. The GE concept is a modification of the ABWR with some additional passive safety features. 16 figs.

  5. Conducting thermomechanical fatigue test in air at light water reactor relevant temperature intervals

    NASA Astrophysics Data System (ADS)

    Ramesh, Mageshwaran; Leber, Hans J.; Diener, Markus; Spolenak, Ralph

    2011-08-01

    In Light Water Reactors (LWR), many structural components are made of austenitic stainless steels (SS). These components are subject to extreme conditions, such as large temperature gradients and pressure loads during service. Hence, the fatigue and fracture behavior of austenitic SS under these conditions has evoked consistent interest over the years. Most studies dealing with this problem in the past, investigated the isothermal fatigue (IF) condition, which is not the case in the service, and less attention has been paid to thermomechanical fatigue (TMF). Moreover, the existing codes of practice and standards for TMF testing are mainly derived from the high temperature TMF tests ( T mean > 400 °C). This work presents the development of a facility to perform TMF tests under LWR relevant temperature interval in air. The realized testing parameters and tolerances are compared with the recommendations of existing codes of practice and standards from high temperature tests. The effectiveness of the testing facility was verified with series of TMF and IF tests performed on specimens made out of a commercial austenitic SS TP347 pipe material. The results revealed that the existing tolerances in standards are quite strict for the application of lower temperature ranges TMF tests. It was found that the synchronous, in-phase (IP) TMF tested specimens possess a higher lifetime than those subjected to the asynchronous, out-of-phase (OP) TMF and IF at T max in the investigated strain range for austenitic SS. Nevertheless, the fatigue lifetime of all the test conditions was similar in the engineering scale.

  6. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    SciTech Connect

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  7. Environmentally assisted cracking in Light Water Reactors. Volume 16: Semiannual report, October 1992--March 1993

    SciTech Connect

    Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

    1993-09-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  8. Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

    SciTech Connect

    Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Kassner, T. F.; Ruther, W. E.; Shack, W. J.; Smith, J. L.; Soppet, W. K.; Strain; R. V.

    1999-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1998 to December 1998. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vessel and piping steels. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in helium at 289 C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments.

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

    SciTech Connect

    Daryl Haefner

    2007-02-01

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

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

    SciTech Connect

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

    1982-05-20

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

  11. Consolidated Fuel Reprocessing Program: Progress report for period October 1 to December 31, 1986

    SciTech Connect

    Groenier, W.S.; Meacham, S.A.; Stradley, J.G.

    1987-06-01

    All research and development (R and D) on civilian power reactor oxide fuel reprocessing in the United States is managed under the Consolidated Fuel Reprocessing Program (CFRP) centered at Oak Ridge National Laboratory (ORNL). A prime focus of present work is on technical exchanges and collaboration with other countries. In this context, the US Department of Energy (DOE) is in the process of negotiating a major collaboration with Japan. Both work associated with the foreign exchanges and collaboration and some on-going work are reported in overview fashion in this series of quarterly progress reports.

  12. Light-water reactors: preliminary safety and environmental information document. Volume I

    SciTech Connect

    Not Available

    1980-01-01

    Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle (Pu/ThO/sub 2/ Burner).

  13. Radionuclide characterization at US commercial light-water reactors for decommissioning assessment: Distributions, inventories, and waste disposal considerations

    SciTech Connect

    Abel, K.H.; Robertson, D.E.; Thomas, C.W.

    1992-09-01

    A continuing research program, conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission, characterizing radionuclide concentrations associated with US light-water reactors has been conducted for more than a decade. The research initially focused upon sampling and analytical measurements for the purpose of establishing radionuclide distributions and inventories for decommissioning assessment, since very little empirical data existed. The initial phase of the research program examined radionuclide concentrations and distributions external to the reactor vessel at seven US light water reactors. Later stages of the research program have examined the radionuclide distributions in the highly radioactive reactor internals and fuel assembly. Most recently, the research program is determining radionuclide concentrations in these highly radioactive components and comparing empirical results with those derived from the several nonempirical methodologies employed to estimate radionuclide inventories for disposal classification. The results of the research program to date are summarized, and their implications and significance for the decommissioning process are noted.

  14. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Semiannual report, October 1990--March 1991: Volume 13

    SciTech Connect

    Doctor, S.R.; Good, M.S.; Heasler, P.G.; Hockey, R.L.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

    1992-07-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties.

  15. Modeling of Complex Wear Behavior Associated with Grid-to-Rod Fretting in Light Water Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Blau, P. J.; Qu, J.; Lu, R.

    2016-11-01

    Fretting wear damage to fuel cladding from flow-induced vibrations can be a significant concern in the operation of light water nuclear reactors. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. The multi-stage model accounts for oxide layers and wear rate transitions. This paper describes the basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.

  16. Modeling of complex wear behavior associated with grid-to-rod fretting in light water nuclear reactors

    SciTech Connect

    Blau, P. J.; Qu, J.; Lu, R.

    2016-09-21

    One significant concern in the operation of light water nuclear reactors is the fretting wear damage to fuel cladding from flow-induced vibrations. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. Furthermore, the multi-stage model accounts for oxide layers and wear rate transitions. Our paper describes the basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.

  17. Modeling of complex wear behavior associated with grid-to-rod fretting in light water nuclear reactors

    DOE PAGES

    Blau, P. J.; Qu, J.; Lu, R.

    2016-09-21

    One significant concern in the operation of light water nuclear reactors is the fretting wear damage to fuel cladding from flow-induced vibrations. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. Furthermore, the multi-stage model accounts for oxide layers and wear rate transitions. Our paper describes themore » basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.« less

  18. 10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Quality Assurance Criteria for Nuclear Power Plants and... Criteria for Nuclear Power Plants and Fuel Reprocessing Plants Introduction. Every applicant for a..., and components of the reactor. Nuclear power plants and fuel reprocessing plants include...

  19. 10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Quality Assurance Criteria for Nuclear Power Plants and... Criteria for Nuclear Power Plants and Fuel Reprocessing Plants Introduction. Every applicant for a..., and components of the reactor. Nuclear power plants and fuel reprocessing plants include...

  20. 10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Quality Assurance Criteria for Nuclear Power Plants and... Criteria for Nuclear Power Plants and Fuel Reprocessing Plants Introduction. Every applicant for a..., and components of the reactor. Nuclear power plants and fuel reprocessing plants include...

  1. American National Standard: design basis for protection of light water nuclear power plants against effects of postulated pipe rupture

    SciTech Connect

    Not Available

    1980-12-31

    This standard addresses the design bases for light water reactor, nuclear power plant structures and components essential for the protection of public health and safety from the potential adverse effects of pipe whip, jet impingement, pressurization of compartments outside containment, environmental conditions and flooding associated with a postulated pipe rupture. The design bases for missile protection and the design bases for containment pressurization are not within this standard.

  2. Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for 2013

    SciTech Connect

    Hallbert, Bruce; Thomas, Ken

    2014-09-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  3. Environmentally assisted cracking in light water reactors annual report January - December 2005.

    SciTech Connect

    Alexandreanu, B.; Chen, Y.; Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.

    2007-08-31

    This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data

  4. Environmentally assisted cracking in light water reactors : semiannual report, July 2000 - December 2000.

    SciTech Connect

    Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.; Energy Technology

    2002-04-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from July 2000 to December 2000. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. The fatigue strain-vs.-life data are summarized for the effects of various material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Effects of the reactor coolant environment on the mechanism of fatigue crack initiation are discussed. Two methods for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations are presented. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. A fracture toughness J-R curve test was conducted on a commercial heat of Type 304 SS that was irradiated to {approx}2.0 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. The results were compared with the data obtained earlier on steels irradiated to 0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) (0.45 and 1.35 dpa). Neutron irradiation at 288 C was found to decrease the fracture toughness of austenitic SSs. Tests were conducted on compact-tension specimens of Alloy 600 under cyclic loading to evaluate the enhancement of crack growth rates in LWR environments. Then, the existing fatigue crack growth data on Alloys 600 and 690 were analyzed to establish the effects of temperature, load ratio, frequency, and stress intensity range

  5. Environmentally assisted cracking in light water reactors - annual report, January-December 2001.

    SciTech Connect

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E; Hiller, R. W.; Shack, W. J.; Soppet, W. K.; Strain, R. V.; Energy Technology

    2003-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2001. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (c) EAC of Alloy 600. The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, dissolved oxygen (DO) level in water, and material heat treatment, on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The mechanism of fatigue crack initiation in austenitic SSs in LWR environments has also been examined. The results indicate that the presence of a surface oxide film or difference in the characteristics of the oxide film has no effect on fatigue crack initiation in austenitic SSs in LWR environments. Slow-strain-rate tensile tests and post-test fractographic analyses were conducted on several model SS alloys irradiated to {approx}2 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) ({approx}3 dpa) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. Corrosion fatigue tests were conducted on nonirradiated austenitic SSs in high-purity water at 289 C to establish the test procedure and conditions that will be used for the tests on irradiated materials. A comprehensive irradiation experiment was initiated to obtain many tensile and disk specimens irradiated under simulated pressurized water reactor conditions at {approx}325 C to 5, 10, 20, and 40 dpa. Crack growth tests were completed on 30% cold-worked Alloy 600 in high-purity water under various environmental and loading conditions. The results are compared with data obtained earlier on several heats of Alloy 600

  6. 3D Simulation of Missing Pellet Surface Defects in Light Water Reactor Fuel Rods

    SciTech Connect

    B.W. Spencer; J.D. Hales; S.R. Novascone; R.L. Williamson

    2012-09-01

    The cladding on light water reactor (LWR) fuel rods provides a stable enclosure for fuel pellets and serves as a first barrier against fission product release. Consequently, it is important to design fuel to prevent cladding failure due to mechanical interactions with fuel pellets. Cladding stresses can be effectively limited by controlling power increase rates. However, it has been shown that local geometric irregularities caused by manufacturing defects known as missing pellet surfaces (MPS) in fuel pellets can lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. Nuclear fuel performance codes commonly use a 1.5D (axisymmetric, axially-stacked, one-dimensional radial) or 2D axisymmetric representation of the fuel rod. To study the effects of MPS defects, results from 1.5D or 2D fuel performance analyses are typically mapped to thermo-mechanical models that consist of a 2D plane-strain slice or a full 3D representation of the geometry of the pellet and clad in the region of the defect. The BISON fuel performance code developed at Idaho National Laboratory employs either a 2D axisymmetric or 3D representation of the full fuel rod. This allows for a computational model of the full fuel rod to include local defects. A 3D thermo-mechanical model is used to simulate the global fuel rod behavior, and includes effects on the thermal and mechanical behavior of the fuel due to accumulation of fission products, fission gas production and release, and the effects of fission gas accumulation on thermal conductivity across the fuel-clad gap. Local defects can be modeled simply by including them in the 3D fuel rod model, without the need for mapping between two separate models. This allows for the complete set of physics used in a fuel performance analysis to be included naturally in the computational representation of the local defect, and for the effects of the

  7. 10 CFR 51.51 - Uranium fuel cycle environmental data-Table S-3.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Uranium fuel cycle environmental data-Table S-3. 51.51... cycle environmental data—Table S-3. (a) Under § 51.50, every environmental report prepared for the construction permit stage or early site permit stage or combined license stage of a light-water-cooled...

  8. 77 FR 38789 - Notice of Availability of Draft Waste Incidental to Reprocessing Evaluation for the Concentrator...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-29

    ... incidental to reprocessing and thus are not high-level radioactive waste (HLW) and may be managed and....1-1, Radioactive Waste Management Manual. DOE is consulting with the Nuclear Regulatory Commission... would be disposed of at a suitable off-site LLW waste disposal facility, either the Area 5...

  9. Evaluation of the effectiveness of manual and automated dialyzers reprocessing after multiple reuses.

    PubMed

    Toniolo, Alexandra do Rosário; Ribeiro, Maíra Marques; Ishii, Marina; da Silva, Cely Barreto; Jenné Mimica, Lycia Mara; Graziano, Kazuko Uchikawa

    2016-06-01

    A cross-sectional study was conducted to evaluate the effectiveness of manual and automated dialyzer reprocessing. Dialyzers were filled with fluid thioglycollate medium from blood and dialysate chambers after being reprocessed and chemically sterilized with 0.2% peracetic acid. They were incubated for 14 days at 35°C ± 2°C, and microbiologic analysis was performed. Microorganisms were identified in 3 of the 11 samples (27.3%) from the blood chambers: Sphingomonas paucimobilis (2/3) and Penicillium spp (1/3) and in 11 of the 11 samples (100%) from the dialysate chambers: S paucimobilis (7/11), Stenotrophomonas maltophilia (4/11), Pseudomonas aeruginosa (3/11), Candida spp (1/11), and Acinetobacter baumannii (1/11). Of the 4 manually reprocessed dialyzers, gram-positive bacillus were identified in 1 sample (25%) from the blood chamber, and Bacillus spp and Burkholderia spp were identified in 1 sample (25%) from the dialysate chamber. The dialyzers reprocessing can pose risks safety because of exposure patient to microorganisms.

  10. Development of a safeguards data acquisition system for the process monitoring of a simulated reprocessing facility

    SciTech Connect

    Wachter, J.W.

    1986-01-01

    As part of the Consolidated Fuel Reprocessing Program of the Fuel Recycle Division at the Oak Ridge National Laboratory (ORNL), an Integrated Process Demonstration (IPD) facility has been constructed for development of reprocessing plant technology. Through the use of cold materials, the IPD facility provides for the integrated operation of the major equipment items of the chemical-processing portion of a nuclear fuel reprocessing plant. The equipment, processes, and the extensive use of computers in data acquisition and control are prototypical of future reprocessing facilities and provide a unique test-bed for nuclear safeguards demonstrations. The data acquisition and control system consists of several microprocessors that communicate with one another and with a host minicomputer over a common data highway. At intervals of a few minutes, a ''snapshot'' is taken of the process variables, and the data are transmitted to a safeguards computer and minicomputer work station for analysis. This paper describes this data acquisition system and the data-handling procedures leading to microscopic process monitoring for safeguards purposes.

  11. Offline Memory Reprocessing: Involvement of the Brain's Default Network in Spontaneous Thought Processes

    PubMed Central

    Wang, Kun; Yu, Chunshui; Xu, Lijuan; Qin, Wen; Li, Kuncheng; Xu, Lin; Jiang, Tianzi

    2009-01-01

    Background Spontaneous thought processes (STPs), also called daydreaming or mind-wandering, occur ubiquitously in daily life. However, the functional significance of STPs remains largely unknown. Methodology/Principal Finding Using functional magnetic resonance imaging (fMRI), we first identified an STPs-network whose activity was positively correlated with the subjects' tendency of having STPs during a task-free state. The STPs-network was then found to be strongly associated with the default network, which has previously been established as being active during the task-free state. Interestingly, we found that offline reprocessing of previously memorized information further increased the activity of the STPs-network regions, although during a state with less STPs. In addition, we found that the STPs-network kept a dynamic balance between functional integration and functional separation among its component regions to execute offline memory reprocessing in STPs. Conclusion/Significance These findings strengthen a view that offline memory reprocessing and STPs share the brain's default network, and thus implicate that offline memory reprocessing may be a predetermined function of STPs. This supports the perspective that memory can be consolidated and modified during STPs, and thus gives rise to a dynamic behavior dependent on both previous external and internal experiences. PMID:19290044

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

    SciTech Connect

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

    2013-07-01

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

  13. Effects of reuse and bleach/formaldehyde reprocessing on polysulfone and polyamide hemodialyzers.

    PubMed

    Cornelius, Rena M; McClung, W Glenn; Barre, Paul; Esguerra, Fe; Brash, John L

    2002-01-01

    The surface features, morphology, and blood interactions of fibers from pristine, bleach/formaldehyde reprocessed, and reused Fresenius Polysulfone High Flux (Hemoflow F80B) hemodialyzers and Gambro Polyflux 21S Polyamide hemodialyzers have been studied. SEM images of fibers from both hemodialyzer types revealed a dense skin layer on the inner surface and a relatively thick porous layer on the outer surface. The 21S polyamide support layer consisted of interconnected highly porous structures. Environmental scanning electron microscopy and atomic force microscopy images of both membrane types showed alterations in morphology due to reprocessing and reuse; however the changes were more marked for the 21S polyamide dialyzers. Fluorescence microscopy images showed only minimal fluorescence associated with the fibers after patient use and reprocessing, suggesting that blood derived deposits were removed by processing. The protein layers formed on pristine and reused hemodialyzer membranes during clinical use were studied using SDS-PAGE and immunoblotting. Before bleach/formaldehyde treatment, protein layers of considerable amount and complexity were found on the blood side of singly and multiply used dialyzers. Proteins adsorbed on the dialysate side were predominantly in the molecular mass region below 30 kDa. However, some higher molecular mass proteins were detected on the dialysate side of the 21 S polyamide dialyzers. Very little protein was detected on dialyzers that were treated with bleach/formaldehyde after dialysis, regardless of whether they had been used/reprocessed once or 12 times.

  14. Reprocessing the Southern Hemisphere ADditional OZonesondes (SHADOZ) Database for Long-Term Trend Analyses

    NASA Astrophysics Data System (ADS)

    Witte, J. C.; Thompson, A. M.; Coetzee, G.; Fujiwara, M.; Johnson, B. J.; Sterling, C. W.; Cullis, P.; Ashburn, C. E.; Jordan, A. F.

    2015-12-01

    SHADOZ is a large archive of tropical balloon-bone ozonesonde data at NASA/Goddard Space Flight Center with data from 14 tropical and subtropical stations provided by collaborators in Europe, Asia, Latin America and Africa . The SHADOZ time series began in 1998, using electrochemical concentration cell (ECC) ozonesondes. Like many long-term sounding stations, SHADOZ is characterized by variations in operating procedures, launch protocols, and data processing such that biases within a data record and among sites appear. In addition, over time, the radiosonde and ozonesonde instruments and data processing protocols have changed, adding to the measurement uncertainties at individual stations and limiting the reliability of ozone profile trends and continuous satellite validation. Currently, the ozonesonde community is engaged in reprocessing ECC data, with an emphasis on homogenization of the records to compensate for the variations in instrumentation and technique. The goals are to improve the information and integrity of each measurement record and to support calculation of more reliable trends. We illustrate the reprocessing activity of SHADOZ with selected stations. We will (1) show reprocessing steps based on the recent WMO report that provides post-processing guidelines for ozonesondes; (2) characterize uncertainties in various parts of the ECC conditioning process; and (3) compare original and reprocessed data to co-located ground and satellite measurements of column ozone.

  15. A Review of Eye Movement Desensitization and Reprocessing (EMDR): Research Findings and Implications for Counsellors.

    ERIC Educational Resources Information Center

    MacCluskie, Kathryn C.

    1998-01-01

    States that within the last six years a new therapeutic technique for the treatment of posttraumatic stress disorder, Eye Movement Desensitization and Reprocessing (EMDR), has emerged. Examines the strengths and weaknesses of published studies concerning EMDR, describes the nature of the debate about the efficacy of EMDR, and reviews implications…

  16. Eye Movement Desensitization and Reprocessing (EMDR) Treatment for Psychologically Traumatized Individuals.

    ERIC Educational Resources Information Center

    Wilson, Sandra A.; And Others

    1995-01-01

    Studies the effects of 3 90-minute Eye Movement Desensitization and Reprocessing (EMDR) treatment sessions on traumatic memories of 80 participants. Participants receiving EMDR showed decreases in complaints and anxiety, and increases in positive cognition. Participants in the delayed-treatment condition showed no improvement in any measures in…

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

    SciTech Connect

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

    2007-07-01

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

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

    SciTech Connect

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

    1980-05-01

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

  19. A Comparison of "at-launch" V03 GPM Data Products and V04 Reprocessed Products

    NASA Astrophysics Data System (ADS)

    Stocker, Erich Franz; Kwiatkowksi, John; Ji, Yimin; Kelley, Owen; Stout, John; Woltz, Lawrence

    2016-04-01

    The first major reprocessing of GPM mission data is scheduled for 1 March 2016. This is an important first-step in the creation of a consistent mission data suite.It will be the first set of data in which the GPM constellation radiometers are intercalibrated using GPM GMI. In addition, the GPROF profile database will be substantially observational and based on the GPM combined data product. The GPM V03 "at-launch" 1C algorithms intercalibrated the constellation radiometers using TRMM TMI as the satellite reference standard. The GPROF profile database used for retrievals was based on TRMM, AMSRE, RadarSat and NMQ quality-controlled ground radar data in addition to model input. The V04 GPROF reprocessing will depend upon a profile database that is constructed using V04 GPM combined (DPR/GMI) observational data. This paper will provide comparison analysis of at launch V03 vs reprocessed V04. It will present global means comparing the V03 and V04 retrievals for both GPROF and the DPR radar. It will also look at the difference of the means separately for Land and Ocean. In addition it will look at specific instances of differences using the GPM gridded text products. In conclusion it will present the plans for a V05 reprocessing which is scheduled to take place in March of 2017.

  20. Scientific Resistance to Research, Training and Utilization of Eye Movement Desensitization and Reprocessing (EMDR) Therapy in Treating Post-War Disorders

    DTIC Science & Technology

    2008-01-01

    and utilization of eye movement desensitization and reprocessing ( EMDR ) therapy in...history: Available online 22 October 2008 Keywords: USA Resistance Zeitgeist Military Eye movement desensitization and reprocessing ( EMDR ) Post...behavioral paradigm or zeitgeist and its chief rival – eye movement desensitization and reprocessing ( EMDR ) in treating

  1. CMORPH Global High-Resolution Precipitation Estimates: Reprocessing, Bias Correction, and Error Quantification

    NASA Astrophysics Data System (ADS)

    Xie, P.; Joyce, R.; Wu, S.

    2015-12-01

    CMORPH global high-resolution satellite precipitation estimates have been reprocessed and bias-corrected for an 18-year period from January 1998 to the present to cover the entire TRMM/GPM era. As the first step of this project, the CMORPH estimates have been reprocessed with a frozen algorithm and with input PMW retrievals and IR observations of identical versions throughout the data period. Bias correction is then performed for the reprocessed CMORPH over the entire data period from 1998 to the present. Over land, the bias in the raw CMORPH is removed by matching the probability density function (PDF) of the CMORPH with that of the CPC unified daily gauge analysis. Over ocean, the raw CMORPH satellite estimates are calibrated against a long-term precipitation analysis (pentad GPCP) to ensure temporal homogeneity for climate applications. The reprocessed, bias-corrected CMORPH estimates present an 18-year homogeneous record of high-resolution precipitation on an 8kmx8km and 30-min resolution covering the globe from 60oS-60oN. Accuracy of the reprocessed, bias corrected CMORPH is examined through comparison against gauge based analysis over land and against buoy measurements over ocean. Overall, the bias corrected CMORPH exhibited improved performance upon the raw CMORPH, with bias removed almost completely over the tropics and extra-tropical regions during warm seasons. Caused by shortcomings in the input PMW retrievals, CMORPH tends to miss precipitation over cold surfaces. Quantitative accuracy of bias corrected CMORPH is a function of season, location, precipitation intensity, and the availability of satellite retrievals. Bias corrected CMORPH is applied to examine the large-scale precipitation variations over the globe. In particular, diurnal variations of precipitation is investigated and compared with those determined by the three new generation reanalyses (CFSR, MERRA, and ERA-Interim). A comprehensive description and diagnostic discussions will be given at

  2. Algorithm Updates for the Fourth SeaWiFS Data Reprocessing

    NASA Technical Reports Server (NTRS)

    Hooker, Stanford, B. (Editor); Firestone, Elaine R. (Editor); Patt, Frederick S.; Barnes, Robert A.; Eplee, Robert E., Jr.; Franz, Bryan A.; Robinson, Wayne D.; Feldman, Gene Carl; Bailey, Sean W.

    2003-01-01

    The efforts to improve the data quality for the Sea-viewing Wide Field-of-view Sensor (SeaWiFS) data products have continued, following the third reprocessing of the global data set in May 2000. Analyses have been ongoing to address all aspects of the processing algorithms, particularly the calibration methodologies, atmospheric correction, and data flagging and masking. All proposed changes were subjected to rigorous testing, evaluation and validation. The results of these activities culminated in the fourth reprocessing, which was completed in July 2002. The algorithm changes, which were implemented for this reprocessing, are described in the chapters of this volume. Chapter 1 presents an overview of the activities leading up to the fourth reprocessing, and summarizes the effects of the changes. Chapter 2 describes the modifications to the on-orbit calibration, specifically the focal plane temperature correction and the temporal dependence. Chapter 3 describes the changes to the vicarious calibration, including the stray light correction to the Marine Optical Buoy (MOBY) data and improved data screening procedures. Chapter 4 describes improvements to the near-infrared (NIR) band correction algorithm. Chapter 5 describes changes to the atmospheric correction and the oceanic property retrieval algorithms, including out-of-band corrections, NIR noise reduction, and handling of unusual conditions. Chapter 6 describes various changes to the flags and masks, to increase the number of valid retrievals, improve the detection of the flag conditions, and add new flags. Chapter 7 describes modifications to the level-la and level-3 algorithms, to improve the navigation accuracy, correct certain types of spacecraft time anomalies, and correct a binning logic error. Chapter 8 describes the algorithm used to generate the SeaWiFS photosynthetically available radiation (PAR) product. Chapter 9 describes a coupled ocean-atmosphere model, which is used in one of the changes

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

    SciTech Connect

    Chen, J.; Wen, M.

    2013-07-01

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

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

    SciTech Connect

    Patterson, M.W.

    1994-10-01

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

  5. Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

    SciTech Connect

    Corradini, M. L.

    2015-06-01

    “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

  6. Light Water Reactor Sustainability Program Grizzly Year-End Progress Report

    SciTech Connect

    Benjamin Spencer; Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner; Marie Backman; Brian Wirth; Stephen Novascone; Jason Hales

    2013-09-01

    The Grizzly software application is being developed under the Light Water Reactor Sustainability (LWRS) program to address aging and material degradation issues that could potentially become an obstacle to life extension of nuclear power plants beyond 60 years of operation. Grizzly is based on INL’s MOOSE multiphysics simulation environment, and can simultaneously solve a variety of tightly coupled physics equations, and is thus a very powerful and flexible tool with a wide range of potential applications. Grizzly, the development of which was begun during fiscal year (FY) 2012, is intended to address degradation in a variety of critical structures. The reactor pressure vessel (RPV) was chosen for an initial application of this software. Because it fulfills the critical roles of housing the reactor core and providing a barrier to the release of coolant, the RPV is clearly one of the most safety-critical components of a nuclear power plant. In addition, because of its cost, size and location in the plant, replacement of this component would be prohibitively expensive, so failure of the RPV to meet acceptance criteria would likely result in the shutting down of a nuclear power plant. The current practice used to perform engineering evaluations of the susceptibility of RPVs to fracture is to use the ASME Master Fracture Toughness Curve (ASME Code Case N-631 Section III). This is used in conjunction with empirically based models that describe the evolution of this curve due to embrittlement in terms of a transition temperature shift. These models are based on an extensive database of surveillance coupons that have been irradiated in operating nuclear power plants, but this data is limited to the lifetime of the current reactor fleet. This is an important limitation when considering life extension beyond 60 years. The currently available data cannot be extrapolated with confidence further out in time because there is a potential for additional damage mechanisms (i

  7. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    SciTech Connect

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  8. Greater-than-Class C low-level radioactive waste characterization. Appendix A-3: Basis for greater-than-Class C low-level radioactive waste light water reactor projections

    SciTech Connect

    Mancini, A.; Tuite, P.; Tuite, K.; Woodberry, S.

    1994-09-01

    This study characterizes low-level radioactive waste types that may exceed Class C limits at light water reactors, estimates the amounts of waste generated, and estimates radionuclide content and distribution within the waste. Waste types that may exceed Class C limits include metal components that become activated during operations, process wastes such as cartridge filters and decontamination resins, and activated metals from decommissioning activities. Operating parameters and current management practices at operating plants are reviewed and used to estimate the amounts of low-level waste exceeding Class C limits that is generated per fuel cycle, including amounts of routinely generated activated metal components and process waste. Radionuclide content is calculated for specific activated metals components. Empirical data from actual low-level radioactive waste are used to estimate radionuclide content for process wastes. Volumes and activities are also estimated for decommissioning activated metals that exceed Class C limits. To estimate activation levels of decommissioning waste, six typical light water reactors are modeled and analyzed. This study does not consider concentration averaging.

  9. Fuel cycle analysis of once-through nuclear systems.

    SciTech Connect

    Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-08-10

    Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium

  10. Prototype Demonstration of Gamma- Blind Tensioned Metastable Fluid Neutron/Multiplicity/Alpha Detector – Real Time Methods for Advanced Fuel Cycle Applications

    SciTech Connect

    McDeavitt, Sean M.

    2016-12-20

    The content of this report summarizes a multi-year effort to develop prototype detection equipment using the Tensioned Metastable Fluid Detector (TMFD) technology developed by Taleyarkhan [1]. The context of this development effort was to create new methods for evaluating and developing advanced methods for safeguarding nuclear materials along with instrumentation in various stages of the fuel cycle, especially in material balance areas (MBAs) and during reprocessing of used nuclear fuel. One of the challenges related to the implementation of any type of MBA and/or reprocessing technology (e.g., PUREX or UREX) is the real-time quantification and control of the transuranic (TRU) isotopes as they move through the process. Monitoring of higher actinides from their neutron emission (including multiplicity) and alpha signatures during transit in MBAs and in aqueous separations is a critical research area. By providing on-line real-time materials accountability, diversion of the materials becomes much more difficult. The Tensioned Metastable Fluid Detector (TMFD) is a transformational technology that is uniquely capable of both alpha and neutron spectroscopy while being “blind” to the intense gamma field that typically accompanies used fuel – simultaneously with the ability to provide multiplicity information as well [1-3]. The TMFD technology was proven (lab-scale) as part of a 2008 NERI-C program [1-7]. The bulk of this report describes the advancements and demonstrations made in TMFD technology. One final point to present before turning to the TMFD demonstrations is the context for discussing real-time monitoring of SNM. It is useful to review the spectrum of isotopes generated within nuclear fuel during reactor operations. Used nuclear fuel (UNF) from a light water reactor (LWR) contains fission products as well as TRU elements formed through neutron absorption/decay chains. The majority of the fission products are gamma and beta emitters and they represent the

  11. Improving proliferation resistance of high breeding gain generation 4 reactors using blankets composed of light water reactor waste

    SciTech Connect

    Hellesen, C.; Grape, S.; Haakanson, A.; Jacobson Svaerd, S.; Jansson, P.

    2013-07-01

    Fertile blankets can be used in fast reactors to enhance the breeding gain as well as the passive safety characteristics. However, such blankets typically result in the production of weapons grade plutonium. For this reason they are often excluded from Generation IV reactor designs. In this paper we demonstrate that using blankets manufactured directly from spent light water (LWR) reactor fuel it is possible to produce a plutonium product with non-proliferation characteristics on a par with spent LWR fuel of 30-50 MWd/kg burnup. The beneficial breeding and safety characteristics are retained. (authors)

  12. Reprocessing of nuclear fuels at the Savannah River Plant

    SciTech Connect

    Gray, L.W.

    1986-10-04

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

  13. Financing Strategies for Nuclear Fuel Cycle Facility

    SciTech Connect

    David Shropshire; Sharon Chandler

    2005-12-01

    To help meet our nation’s energy needs, reprocessing of spent nuclear fuel is being considered more and more as a necessary step in a future nuclear fuel cycle, but incorporating this step into the fuel cycle will require considerable investment. This report presents an evaluation of financing scenarios for reprocessing facilities integrated into the nuclear fuel cycle. A range of options, from fully government owned to fully private owned, was evaluated using a DPL (Dynamic Programming Language) 6.0 model, which can systematically optimize outcomes based on user-defined criteria (e.g., lowest life-cycle cost, lowest unit cost). Though all business decisions follow similar logic with regard to financing, reprocessing facilities are an exception due to the range of financing options available. The evaluation concludes that lowest unit costs and lifetime costs follow a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. Other financing arrangements, however, including regulated utility ownership and a hybrid ownership scheme, led to acceptable costs, below the Nuclear Energy Agency published estimates. Overwhelmingly, uncertainty in annual capacity led to the greatest fluctuations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; the annual operating costs dominate the government case. It is concluded that to finance the construction and operation of such a facility without government ownership could be feasible with measures taken to mitigate risk, and that factors besides unit costs should be considered (e.g., legal issues, social effects, proliferation concerns) before making a decision on financing strategy.

  14. Assessment for advanced fuel cycle options in CANDU

    SciTech Connect

    Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

    2013-07-01

    The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a driver fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.

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

    SciTech Connect

    Mcwilliams, A. J.

    2015-09-08

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

  16. Desensitization of triggers and urge reprocessing for pathological gambling: a case series.

    PubMed

    Bae, Hwallip; Han, Changwoo; Kim, Daeho

    2015-03-01

    This case series introduces the desensitization of triggers and urge reprocessing (DeTUR), as a promising adjunctive therapy in addition to comprehensive treatment package for pathological gambling. This addiction protocol of eye movement desensitization and reprocessing was delivered to four male inpatients admitted to a 10-week inpatient program for pathological gambling. The therapist gave three 60-min weekly sessions of the DeTUR using bilateral stimulation (horizontal eye movements or alternative tactile stimuli) focusing on the hierarchy of triggering situations and the urge to initiate gambling behaviors. After treatment, self-reported gambling symptoms, depression, anxiety, and impulsiveness were all improved, and all the participants reported satisfaction with the therapy. They were followed up for 6 months and all maintained their abstinence from gambling and their symptomatic improvements. Given the efficiency (i.e., brevity and efficacy) of the treatment, a controlled study to confirm the effects of the DeTUR on pathological gambling would be justified.

  17. Eye movement desensitization and reprocessing (EMDR) in the treatment of war veterans.

    PubMed

    Silver, Steven M; Rogers, Susan; Russell, Mark

    2008-08-01

    Recent practice guidelines and meta-analyses have designated eye movement desensitization and reprocessing (EMDR) as a first-line treatment for trauma. Eye movement desensitization and reprocessing is an eight-phase therapeutic approach guided by an information-processing model that addresses the combat veteran's critical incidents, current triggers, and behaviors likely to prove useful in his or her future. Two case examples of combat veterans illustrate the ability of EMDR to achieve symptom reduction in a variety of clinical domains (e.g., anxiety, depression, anger, physical pain) simultaneously without requiring the patient to carry out homework assignments or discuss the details of the event. The treatment of phantom limb pain and other somatic presentations is also reviewed. The ability of EMDR to achieve positive effects without homework indicates that it can be effectively employed on consecutive days, making it especially useful during combat situations.

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

    SciTech Connect

    Marra, J.; Billings, A.

    2009-06-24

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

  19. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Semiannual report, April 1992--September 1992: Volume 16

    SciTech Connect

    Doctor, S.R.; Diaz, A.A.; Friley, J.R.; Greenwood, M.S.; Heasler, P.G.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Vo, T.V.

    1993-11-01

    The Evaluation and Improvement of NDE Reliability for Inservice inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs);using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel and other components inspected in accordance with Section XI of the ASME Code. This is a programs report covering the programmatic work from April 1992 through September 1992.

  20. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Volume 14, Semiannual report, April 1991--September 1991

    SciTech Connect

    Doctor, S.R.; Diaz, A.A.; Friley, J.R.; Good, M.S.; Greenwood, M.S.; Heasler, P.G.; Hockey, R.L.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

    1992-07-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWR`s); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from April 1991 through September 1991.

  1. A GASFLOW analysis of a steam explosion accident in a typical light-water reactor confinement building

    SciTech Connect

    Travis, J.R.; Wilson, T.L.; Spore, J.W.; Lam, K.L.; Rao, D.V.

    1994-09-01

    Steam over-pressurization resulting from ex-vessel steam explosion (fuel-coolant interaction) may pose a serious challenge to the integrity of a typical light-water reactor confinement building. If the steam generation rate exceeds the removal capacity of the Airborne Activity Confinement System, confinement over pressurization occurs. Thus, there is a large potential for an uncontrolled and unfiltered release of fission products from the confinement atmosphere to the environment at the time of the steam explosion. The GASFLOW computer code was used to analyze the effects of a hypothetical steam explosion and the transport of steam and hydrogen throughout a typical light-water reactor confinement building. The effects of rapid pressurization and the resulting forces on the internal structures and the heat exchanger service bay hatch covers were calculated. Pressurization of the ventilation system and the potential damage to the ventilation fans and high-efficiency particulate air filters were assessed. Because of buoyancy forces and the calculated confinement velocity field, the hydrogen diffuses and mixes in the confinement atmosphere but tends to be transported to its upper region.

  2. Nondestructive examination (NDE) reliability for inservice inspection of light water reactors. Annual report, October 1989--September 1990: Volume 12

    SciTech Connect

    Doctor, S.R.; Good, M.S.; Heasler, P.G.; Hockey, R.L.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

    1992-05-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section 11 of the ASME Code. This is a progress report covering the programmatic work from October 1989 through September 1990.

  3. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Volume 15, Semiannual report: October 1991--March 1992

    SciTech Connect

    Doctor, S.R.; Diaz, A.A.; Friley, J.R.

    1993-09-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from October 1991 through March 1992.

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

    DOEpatents

    Horwitz, E. Philip; Delphin, Walter H.

    1979-07-24

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

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

    PubMed

    Bevelacqua, J J

    2011-07-01

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

  6. Tropospheric water vapour over Antarctia from 12 years of globally reprocessed GPS data.

    NASA Astrophysics Data System (ADS)

    Thomas, I. D.; King, M. A.; Clarke, P. J.

    2008-12-01

    Atmospheric water vapour - a natural greenhouse gas of importance in the climate system - remains poorly monitored for some regions of the Earth. This paper presents an analysis of Global Positioning System (GPS) data that results in a new precipitable water (PW) dataset for Antarctica. We have undertaken a global reprocessing of a 60 station GPS network for the 1995-2006 period. In order to obtain optimal and homogeneous estimates of PW, emphasis has been placed on using the latest and most up-to-date GPS observation models in a consistent manner. These models include the VMF1 tropospheric mapping function and absolute antenna phase centre variations. Zenith delays, estimated at two-hourly intervals, were converted to PW for twelve, mainly coastal, Antarctic locations. Comparison with radiosonde derived PW time series shows that the reprocessed GPS measurement technique exhibits good temporal stability. GPS / radiosonde biases are small, at the sub-millimetre level; correlations are of the order of 0.95. Comparison with PW datasets obtained from MODIS, AIRS and AMSR-E satellite instruments shows the AIRS instrument to give the best agreement with the GPS derived time series, again with sub-millimetre biases. The PW time series capture the extremely dry climate of Antarctica. Periodic signals in the PW time series were also observed, including a strong annual signal at all locations, and a semiannual signal at coastal East Antarctic locations. A 12-hourly (S1) periodic variation is observed in summer PW measurements at many sites. There is also a possible increase in summertime PW over the 1995- 2006 period in coastal East Antarctica and at the South Pole. We conclude that globally reprocessed GPS solutions can provide accurate measurements of PW that will become increasingly useful for meteorological and climatological applications, provided that the GPS data are (re)processed consistently and homogeneously.

  7. Removal of actinides from nuclear reprocessing wastes: a pilot plant study using non-radioactive simulants

    SciTech Connect

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

    1980-01-01

    Nuclear fuel reprocessing wastes generated at the ICPP contain small amounts of actinides, primarily Pu and Am. Removal of these actinides reduces the long term storage hazards of the waste. The development of a flowsheet to remove trivalent actinides is discussed in this paper. Pilot plant studies used actinide simulants. As a result of these studies, the Height of a Transfer Unit (HTU) was selected as the better measure of pulse column separation efficiency.

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

    SciTech Connect

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

    1984-08-01

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

  9. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    SciTech Connect

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

  10. Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactors (Progress report for work through June 2002, 12th quarterly report)

    SciTech Connect

    Mac Donald, Philip Elsworth

    2002-09-01

    The overall objective of this NERI project is to evaluate the potential advantages and disadvantages of an optimized thorium-uranium dioxide (ThO2/UO2) fuel design for light water reactors (LWRs). The project is led by the Idaho National Engineering and Environmental Laboratory (INEEL), with the collaboration of three universities, the University of Florida, Massachusetts Institute of Technology (MIT), and Purdue University; Argonne National Laboratory; and all of the Pressurized Water Reactor (PWR) fuel vendors in the United States (Framatome, Siemens, and Westinghouse). In addition, a number of researchers at the Korean Atomic Energy Research Institute and Professor Kwangheon Park at Kyunghee University are active collaborators with Korean Ministry of Science and Technology funding. The project has been organized into five tasks: · Task 1 consists of fuel cycle neutronics and economics analysis to determine the economic viability of various ThO2/UO2 fuel designs in PWRs, · Task 2 will determine whether or not ThO2/UO2 fuel can be manufactured economically, · Task 3 will evaluate the behavior of ThO2/UO2 fuel during normal, off-normal, and accident conditions and compare the results with the results of previous UO2 fuel evaluations and U.S. Nuclear Regulatory Commission (NRC) licensing standards, · Task 4 will determine the long-term stability of ThO2/UO2 high-level waste, and · Task 5 consists of the Korean work on core design, fuel performance analysis, and xenon diffusivity measurements.

  11. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    DTIC Science & Technology

    2009-07-01

    uranium to make mixed-oxide ( MOX ) fuel, in which the 239Pu largely substitutes for 235U. Two French reprocessing plants at La Hague can each...Managing the Nuclear Fuel Cycle Congressional Research Service 17 plant.54 About 200 metric tons of MOX fuel is used annually, about 2% of new nuclear...fuel,55 equivalent to about 2,000 metric tons of mined uranium.56 However, the benefits of reprocessing spent fuel to make MOX fuel for today’s

  12. Nuclear fuel cycle facility accident analysis handbook

    SciTech Connect

    Ayer, J E; Clark, A T; Loysen, P; Ballinger, M Y; Mishima, J; Owczarski, P C; Gregory, W S; Nichols, B D

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH.

  13. A new thermal hydraulics code coupled to agent for light water reactor analysis

    NASA Astrophysics Data System (ADS)

    Eklund, Matthew Deric

    A new numerical model for coupling a thermal hydraulics method based on the Drift Flux and Homogeneous Equilibrium Mixture (HEM) models, with a deterministic neutronics code system AGENT (Arbitrary Geometry Neutron Transport), is developed. Named the TH thermal hydraulics code, it is based on the mass continuity, momentum, and energy equations integrated with appropriate relations for liquid and vapor phasic velocities. The modified conservation equations are then evaluated in one-dimensional (1D) steady-state conditions for LWR coolant subchannel in the axial direction. This permits faster computation times without sacrificing significant accuracy, as compared to other three-dimensional (3D) codes such as RELAP5/TRACE. AGENT is a deterministic neutronics code system based on the Method of Characteristics to solve the 2D/3D neutron transport equation in current and future reactor systems. The coupling scheme between the TH and AGENT codes is accomplished by computing the normalized fission rate profile in the LWR fuel elements by AGENT. The normalized fission rate profile is then transferred to the TH thermal hydraulics code for computing the reactor coolant properties. In conjunction with the 1D axial TH code, a separate 1D radial heat transfer model within the TH code is used to determine the average fuel temperature at each node where coolant properties are calculated. These properties then are entered into Scale 6.1, a criticality analysis code, to recalculate fuel pin neutron interaction cross sections based on thermal feedback. With updated fuel neutron interaction cross sections, the fission rate profile is recalculated in AGENT, and the cycle continues until convergence is reached. The TH code and coupled AGENT-TH code are benchmarked against the TRACE reactor analysis software, showing required agreement in evaluating the basic reactor parameters.

  14. Modification of Wide-Band-Gap Oxide Semiconductors with Cobalt Hydroxide Nanoclusters for Visible-Light Water Oxidation.

    PubMed

    Maeda, Kazuhiko; Ishimaki, Koki; Tokunaga, Yuki; Lu, Daling; Eguchi, Miharu

    2016-07-11

    Cobalt-based compounds, such as cobalt(II) hydroxide, are known to be good catalysts for water oxidation. Herein, we report that such cobalt species can also activate wide-band-gap semiconductors towards visible-light water oxidation. Rutile TiO2 powder, a well-known wide-band-gap semiconductor, was capable of harvesting visible light with wavelengths of up to 850 nm, and thus catalyzed water oxidation to produce molecular oxygen, when decorated with cobalt(II) hydroxide nanoclusters. To the best of our knowledge, this system constitutes the first example that a particulate photocatalytic material that is capable of water oxidation upon excitation by visible light can also operate at such long wavelengths, even when it is based on earth-abundant elements only.

  15. Large-Scale Water-Vapor Two-Phase Flow Simulations in Advanced Light Water Reactor Cores

    SciTech Connect

    Hiroyuki, Yoshida; Kazuyuki, Takase; Hidesada, Tamai; Hajime, Akimoto; Yasuo, Ose

    2004-07-01

    Fluid flow characteristics in a fuel bundle of a reduced-moderation light water reactor (RMWR) with a tight-lattice core were analyzed numerically using a newly developed two-phase flow analysis code under the full bundle size condition. Conventional analysis methods such as subchannel codes need composition equations based on the experimental data. In case that there are no experimental data regarding to the thermal-hydraulics in the tight-lattice core, therefore, it is difficult to obtain high prediction accuracy on the thermal design of the RMWR. Then the large-scale direct numerical simulations with a super computer were chosen. The axial velocity distribution in a fuel bundle changed sharply around a spacer. Momentum transfer of vapor in a tight-lattice core is linear along the flow direction. The interface characteristics between water and vapor were clarified quantitatively. (authors)

  16. Protective structures on the surface of zirconium components of light water reactor cores: Formation, testing, and prototype equipment

    NASA Astrophysics Data System (ADS)

    Begrambekov, L. B.; Gordeev, A. A.; Evsin, A. E.; Ivanova, S. V.; Kaplevsky, A. S.; Sadovskiy, Ya. A.

    2015-12-01

    The results of tests of plasma treatment of zirconium and deposition of protective yttrium coatings used as the methods of protection of zirconium components of light water reactor cores against hydrogenation are detailed. The amount of hydrogen in the treated sample exposed to superheated steam for 2500 h at temperature T = 400°C and pressure p = 1 atm was five times lower than the corresponding value for the untreated one. The amount of hydrogen in the sample coated with yttrium remained almost unchanged in 4000 h of exposure. A plasma method for rapid testing for hydrogen resistance is proposed. The hydrogenation rate provided by this method is 700 times higher than that in tests with superheated steam. The results of preliminary experiments confirm the possibility of constructing a unit for batch processing of the surfaces of fuel rod claddings.

  17. An Estimate of the Cost of Electricity from Light Water Reactors and Fossil Plants with Carbon Capture and Sequestration

    SciTech Connect

    Simon, A J

    2009-08-21

    As envisioned in this report, LIFE technology lends itself to large, centralized, baseload (or 'always on') electrical generation. Should LIFE plants be built, they will have to compete in the electricity market with other generation technologies. We consider the economics of technologies with similar operating characteristics: significant economies of scale, limited capacity for turndown, zero dependence on intermittent resources and ability to meet environmental constraints. The five generation technologies examined here are: (1) Light Water Reactors (LWR); (2) Coal; (3) Coal with Carbon Capture and Sequestration (CCS); (4) Natural Gas; and (5) Natural Gas with Carbon Capture and Sequestration. We use MIT's cost estimation methodology (Du and Parsons, 2009) to determine the cost of electricity at which each of these technologies is viable.

  18. Radiological characteristics of light-water reactor spent fuel: A literature survey of experimental data. [82 references

    SciTech Connect

    Roddy, J.W.; Mailen, J.C.

    1987-12-01

    This survey brings together the experimentally determined light-water reactor spent fuel data comprising radionuclide composition, decay heat, and photon and neutron generation rates as identified in a literature survey. Many citations compare these data with values calculated using a radionuclide generation and depletion computer code, ORIGEN, and these comparisons have been included. ORIGEN is a widely recognized method for estimating the actinide, fission product, and activation product contents of irradiated reactor fuel, as well as the resulting heat generation and radiation levels. These estimates are used as source terms in safety evaluations of operating reactors, for evaluation of fuel behavior and regulation of the at-reactor storage, for transportation studies, and for evaluation of the ultimate geologic storage of spent fuel. 82 refs., 4 figs., 17 tabs.

  19. Computerized operating procedures for shearing and dissolution of segments from LWBR (Light Water Breeder Reactor) fuel rods

    SciTech Connect

    Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

    1987-05-01

    This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work.

  20. Protective structures on the surface of zirconium components of light water reactor cores: Formation, testing, and prototype equipment

    SciTech Connect

    Begrambekov, L. B.; Gordeev, A. A.; Evsin, A. E. Ivanova, S. V.; Kaplevsky, A. S.; Sadovskiy, Ya. A.

    2015-12-15

    The results of tests of plasma treatment of zirconium and deposition of protective yttrium coatings used as the methods of protection of zirconium components of light water reactor cores against hydrogenation are detailed. The amount of hydrogen in the treated sample exposed to superheated steam for 2500 h at temperature T = 400°C and pressure p = 1 atm was five times lower than the corresponding value for the untreated one. The amount of hydrogen in the sample coated with yttrium remained almost unchanged in 4000 h of exposure. A plasma method for rapid testing for hydrogen resistance is proposed. The hydrogenation rate provided by this method is 700 times higher than that in tests with superheated steam. The results of preliminary experiments confirm the possibility of constructing a unit for batch processing of the surfaces of fuel rod claddings.

  1. Light Water Reactor Sustainability Program Risk-Informed Safety Margins Characterization (RISMC) PathwayTechnical Program Plan

    SciTech Connect

    Curtis Smith; Cristian Rabiti; Richard Martineau

    2012-11-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”

  2. Pilot program to identify valve failures which impact the safety and operation of light water nuclear power plants

    SciTech Connect

    Tsacoyeanes, J. C.; Raju, P. P.

    1980-04-01

    The pilot program described has been initiated under the Department of Energy Light Water Reactor Safety Research and Development Program and has the following specific objectives: to identify the principal types and causes of failures in valves, valve operators and their controls and associated hardware, which lead to, or could lead to plant trip; and to suggest possible remedies for the prevention of these failures and recommend future research and development programs which could lead to minimizing these valve failures or mitigating their effect on plant operation. The data surveyed cover incidents reported over the six-year period, beginning 1973 through the end of 1978. Three sources of information on valve failures have been consulted: failure data centers, participating organizations in the nuclear power industry, and technical documents.

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

    USGS Publications Warehouse

    Yager, R.M.

    1987-01-01

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

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

    SciTech Connect

    E. R. Johnson; R. E. Best

    2009-12-28

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

  5. The NASA/GEWEX Surface Radiation Budget: Integrated Data Product With Reprocessed Radiance, Cloud, and Meteorology Inputs

    NASA Astrophysics Data System (ADS)

    Stackhouse, P. W.; Gupta, S. K.; Cox, S. J.; Mikovitz, J. C.; Zhang, T.

    2015-12-01

    The NASA/GEWEX Surface Radiation Budget (SRB) project produces shortwave and longwave surface and top of atmosphere radiative fluxes for the 1983-near present time period. Spatial resolution is 1 degree. The current release 3.0 (available at gewex-srb.larc.nasa.gov) uses the International Satellite Cloud Climatology Project (ISCCP) DX product for pixel level radiance and cloud information. This product is subsampled to 30 km. ISCCP is currently recalibrating and recomputing their entire data series, to be released as the H product, at 10km resolution. The ninefold increase in pixel number will allow SRB a higher resolution gridded product (e.g. 0.5 degree), as well as the production of pixel-level fluxes. Other key input improvements include a detailed aerosol history using the Max Planck Institut Aerosol Climatology (MAC), temperature and moisture profiles from HIRS, and new topography, surface type, and snow/ice. At the time of abstract submission, results from the year 2007 have been produced. More years will be added as ISCCP reprocessing occurs. Here we present results for the improved GEWEX Shortwave and Longwave algorithm (GSW and GLW) with new ISCCP data, the various other improved input data sets and the incorporation of many additional internal SRB model improvements. Improvements in GSW include an expansion of the number of wavelength bands from five to eighteen, and the inclusion of ice cloud vs. water cloud radiative transfer. The SRB data produced will be released as part of the Release 4.0 Integrated Product, recognizing the interdependence of the radiative fluxes with other GEWEX products providing estimates of the Earth's global water and energy cycle (I.e., ISCCP, SeaFlux, LandFlux, NVAP, etc.).

  6. SALT observation of X-ray pulse reprocessing in 4U 1626-67

    NASA Astrophysics Data System (ADS)

    Raman, Gayathri; Paul, Biswajit; Bhattacharya, Dipankar; Mohan, Vijay

    2016-07-01

    We investigate optical reprocessing of X-rays in the LMXB pulsar 4U 1626-67 in its current spin-up phase using observations with Southern African Large Telescope (SALT), near-simultaneous observations with Swift-XRT and non-simultaneous RXTE-PCA observations and present the results of timing analysis. Using SALT observations carried out on 5th and 6th March, 2014, we detect some interesting reprocessing signatures. We detect a weak optical Quasi Periodic Oscillation (QPO) in the power density spectrum on March 5th at 48 mHz with a fractional rms of 3.3% in spite of the fact that source shows no corresponding X-ray QPO in the spin-up phase. In the light curve obtained on March 5th, we detect a coherent pulsation at the spin period of ~7.677 s. A previously known, slightly down-shifted side-band is also detected at 129.92 mHz. The frequency spacing between main pulse and this side-band is different from earlier observations, though the statistical significance of the difference is limited. The light curve of 6th March displays short time-scale variability in the form of flares on time-scales of a few minutes. Folded pulse profiles resulting from data of this night show an interesting trend of pulse peak drifting. This drift could be due to i) rapid changes in the reprocessing agent, like orbital motion of an accretion disk warp around the neutron star or ii) intrinsic pulse phase changes in X-rays. We also examine some X-ray light curves obtained with RXTE-PCA during 2008-2010 for pulse shape changes in short time scales during X-ray flares.

  7. SALT observation of X-ray pulse reprocessing in 4U 1626-67

    NASA Astrophysics Data System (ADS)

    Raman, Gayathri; Paul, Biswajit; Bhattacharya, Dipankar; Mohan, Vijay

    2016-05-01

    We investigate optical reprocessing of X-rays in the low-mass X-ray binary (LMXB) pulsar 4U 1626-67 in its current spin-up phase using observations with Southern African Large Telescope (SALT), near-simultaneous observations with Swift-X-ray Telescope and non-simultaneous RXTE-Proportional Counter Array (PCA) observations and present the results of timing analysis. Using SALT observations carried out on 2014 March 5 and 6, we detect some interesting reprocessing signatures. We detect a weak optical quasiperiodic oscillation (QPO) in the power density spectrum on March 5 at 48 mHz with a fractional rms of 3.3 per cent in spite of the fact that source shows no corresponding X-ray QPO in the spin-up phase. In the light curve obtained on March 5, we detect a coherent pulsation at the spin period of ˜7.677 s. A previously known, slightly down-shifted side-band is also detected at 129.92 mHz. The frequency spacing between main pulse and this side-band is different from earlier observations, though the statistical significance of the difference is limited. The light curve of March 6 displays short time-scale variability in the form of flares on time-scales of a few minutes. Folded pulse profiles resulting from data of this night show an interesting trend of pulse peak drifting. This drift could be due to (i) rapid changes in the reprocessing agent, like orbital motion of an accretion disc warp around the neutron star, or (ii) intrinsic pulse phase changes in X-rays. We also examine some X-ray light curves obtained with RXTE-PCA during 2008-2010 for pulse shape changes in short time-scales during X-ray flares.

  8. Impacts of reprocessed altimetry on the surface circulation and variability of the Western Alboran Gyre

    NASA Astrophysics Data System (ADS)

    Juza, Mélanie; Escudier, Romain; Pascual, Ananda; Pujol, Marie-Isabelle; Taburet, Guillaume; Troupin, Charles; Mourre, Baptiste; Tintoré, Joaquín

    2016-08-01

    New altimetry products in semi-enclosed seas are of major interest given the importance of the coastal-open ocean interactions. This study shows how reprocessed altimetry products in the Mediterranean Sea from Archiving, Validation and Interpolation of Satellite Oceanographic data (AVISO) have improved the representation of the surface circulation over the 1993-2012 period. We focus on the Alboran Sea, which is the highest mesoscale activity area of the western Mediterranean. The respective impacts of the new mean dynamic topography (MDT) and mapped sea level anomaly (MSLA) on the description of the Western Alboran Gyre (WAG) are quantitatively evaluated. The temporal mean and variability of the total kinetic energy have been significantly increased in the WAG considering both the new MDT and MSLA (by more than 50%). The new MDT has added 39% to the mean kinetic energy, while the new MSLA has increased the eddy kinetic energy mean (standard deviation) by 53% (30%). The new MSLA has yielded higher variability of total (eddy) kinetic energy, especially in the annual frequency band by a factor of 2 (3). The MDT reprocessing has particularly increased the low-frequency variability of the total kinetic energy by a factor of 2. Geostrophic velocities derived from the altimetry products have also been compared with drifter data. Both reprocessed MDT and MSLA products intensify the velocities of the WAG making them closer to the in situ estimations, reducing the root mean square differences and increasing the correlation for the zonal and meridional components. The results obtained using refined coastal processing of altimetry products and new observational data are very encouraging to better understand the ocean circulation variability and coastal-open ocean interactions, and for potential improvements in other sub-basins, marginal seas and coastal global ocean.

  9. Toward mechanistic understanding of nuclear reprocessing chemistries by quantifying lanthanide solvent extraction kinetics via microfluidics with constant interfacial area and rapid mixing.

    PubMed

    Nichols, Kevin P; Pompano, Rebecca R; Li, Liang; Gelis, Artem V; Ismagilov, Rustem F

    2011-10-05

    The closing of the nuclear fuel cycle is an unsolved problem of great importance. Separating radionuclides produced in a nuclear reactor is useful both for the storage of nuclear waste and for recycling of nuclear fuel. These separations can be performed by designing appropriate chelation chemistries and liquid-liquid extraction schemes, such as in the TALSPEAK process (Trivalent Actinide-Lanthanide Separation by Phosphorus reagent Extraction from Aqueous Komplexes). However, there are no approved methods for the industrial scale reprocessing of civilian nuclear fuel in the United States. One bottleneck in the design of next-generation solvent extraction-based nuclear fuel reprocessing schemes is a lack of interfacial mass transfer rate constants obtained under well-controlled conditions for lanthanide and actinide ligand complexes; such rate constants are a prerequisite for mechanistic understanding of the extraction chemistries involved and are of great assistance in the design of new chemistries. In addition, rate constants obtained under conditions of known interfacial area have immediate, practical utility in models required for the scaling-up of laboratory-scale demonstrations to industrial-scale solutions. Existing experimental techniques for determining these rate constants suffer from two key drawbacks: either slow mixing or unknown interfacial area. The volume of waste produced by traditional methods is an additional, practical concern in experiments involving radioactive elements, both from disposal cost and experimenter safety standpoints. In this paper, we test a plug-based microfluidic system that uses flowing plugs (droplets) in microfluidic channels to determine absolute interfacial mass transfer rate constants under conditions of both rapid mixing and controlled interfacial area. We utilize this system to determine, for the first time, the rate constants for interfacial transfer of all lanthanides, minus promethium, plus yttrium, under TALSPEAK

  10. Laser-fluorescence monitoring of {sup 129}I in industrial reprocessing of nuclear fuel

    SciTech Connect

    Kireev, S.V.; Protsenko, E.D.; Shnyrev, L.S.; Veselov, V.K.; Isupov, V.K.

    1995-01-01

    A laser-fluorescence system is developed that uses a He-Ne (633 nm) laser for continuous monitoring of {sup 129}I during HNO{sub 3} dissolution of irradiated nuclear fuel. Tests at Khlopin Radium Institute indicated that the sensitivity of the {sup 129}I detection (at worst 8{center_dot}10{sup {minus}5} g/m{sup 3}) ensures reliable monitoring of this isotope both directly in the working gas during industrial reprocessing of irradiated nuclear fuel and following gas-purification systems at radiochemical plants.

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

    SciTech Connect

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

    1988-01-01

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

  12. Reprocessing of multi-channel seismic-reflection data collected in the Chukchi Sea

    USGS Publications Warehouse

    Agena, W.F.; Lee, M.W.; Hart, P.E.

    2001-01-01

    Contained on this set of two CD-ROMs are stacked and migrated multi-channel seismic-reflection data for 44 lines recorded in the Chukchi Sea, northern Alaska, by the United States Geological Survey in 1977, 1978, and 1980. All data were reprocessed by the USGS in 2000 using updated methods. The resulting final data have both increased temporal and spatial resolution thus providing improved interpretability. An added benefit of these CD-ROMs is that they are a more stable, long-term archival medium for the data.

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

    DOEpatents

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

    1982-01-19

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

  14. The Development of a Quality Management Framework for Evaluating Medical Device Reprocessing Practice in Healthcare Facilities.

    PubMed

    Lorv, Bailey; Horodyski, Robin; Welton, Cynthia; Vail, John; Simonetto, Luca; Jokanovic, Danilo; Sharma, Richa; Mahoney, Angela Rea; Savoy-Bird, Shay; Bains, Shalu

    2017-01-01

    There is increasing awareness of the importance of medical device reprocessing (MDR) for the provision of safe patient care. Although industry service standards are available to guide MDR practices, there remains a lack of published key performance indicators (KPIs) and targets that are necessary to evaluate MDR quality for feedback and improvement. This article outlines the development of an initial framework that builds on established guidelines and includes service standards, KPIs and targets for evaluating MDR operations. This framework can support healthcare facilities in strengthening existing practices and enables a platform for collaboration towards better MDR performance management.

  15. Use of process monitoring for verifying facility design of large-scale reprocessing plants

    SciTech Connect

    Hakkila, E.A.; Zack, N.R. ); Ehinger, M.H. ); Franssen, F. )

    1991-01-01

    During the decade of the 1990s, the International Atomic Energy Agency (IAEA) faces the challenge of implementing safeguards in large, new reprocessing facilities. The Agency will be involved in the design, construction, checkout and initial operation of these new facilities to ensure effective safeguards are implemented. One aspect of the Agency involvement is in the area of design verification. The United States Support Program has initiated a task to develop methods for applying process data collection and validation during the cold commissioning phase of plant construction. This paper summarizes the results of this task. 14 refs., 1 tab.

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

    DOEpatents

    Forsberg, Charles W.

    1992-01-01

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

  17. Consolidated Fuel Reprocessing Program. Operating experience with pulsed-column holdup estimators

    SciTech Connect

    Ehinger, M.H.

    1986-01-01

    Methods for estimating pulsed-column holdup are being investigated as part of the Safeguards Assessment task of the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory. The CFRP was a major sponsor of test runs at the Barnwell Nuclear Fuel plant (BNFP) in 1980 and 1981. During these tests, considerable measurement data were collected for pulsed columns in the plutonium purification portion of the plant. These data have been used to evaluate and compare three available methods of holdup estimation.

  18. Design features and remote maintenance test results for equipment racks designed for reprocessing cell applications

    SciTech Connect

    Schrock, S.L.; Chesser, J.B.; Peishel, F.L.

    1989-01-01

    This paper describes a concept for equipment rack design and cell placement for highly radioactive process cells developed by the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL). This concept takes advantage of the dexterity and mobility of advanced bridge-mounted, force-reflecting servomanipulators to minimize cell size and increase facility availability. Several prototype racks have been fabricated and maintenance demonstrations have been performed on equipment mounted on these racks. The results of these tests are also described in this paper. 3 refs., 6 figs.

  19. Prolonged Exposure versus Eye Movement Desensitization and Reprocessing (EMDR) for PTSD rape victims.

    PubMed

    Rothbaum, Barbara Olasov; Astin, Millie C; Marsteller, Fred

    2005-12-01

    This controlled study evaluated the relative efficacy of Prolonged Exposure (PE) and Eye Movement Desensitization and Reprocessing (EMDR) compared to a no-treatment wait-list control (WAIT) in the treatment of PTSD in adult female rape victims (n = 74). Improvement in PTSD as assessed by blind independent assessors, depression, dissociation, and state anxiety was significantly greater in both the PE and EMDR group than the WAIT group (n = 20 completers per group). PE and EMDR did not differ significantly for change from baseline to either posttreatment or 6-month follow-up measurement for any quantitative scale.

  20. Present and future potential of krypton-85 for the detection of clandestine reprocessing plants for treaty verification.

    PubMed

    Schoeppner, Michael; Glaser, Alexander

    2016-10-01

    Burnup calculations are applied to determine the amount of krypton-85 that is produced during the irradiation of nuclear fuel. Since krypton-85 is most likely released into the atmosphere during reprocessing to separate plutonium, atmospheric transport modeling is used to calculate the worldwide distribution of krypton-85 concentrations stemming from emissions from declared reprocessing plants. The results are the basis for scenarios in which emissions from clandestine reprocessing facilities have to be detected against various background levels. It is concluded that today's background imposes heavily on the ability to detect small and medium plutonium separation rates; only high separation rates of 1 SQ per week and higher have a chance to be detected with feasible outlay. A fixed network of monitoring stations seems too costly; instead the high number of samples that are required rather calls for mobile sampling procedures, where air samples are collected at random locations over the world and are analyzed in regional laboratories for their krypton-85 concentration. Further, it is argued that krypton-85 emissions from declared reprocessing activities have to be significantly lowered to enable a worldwide verification of the absence of even smaller clandestine reprocessing. For each scenario the number of samples that have to be taken for probable detection is calculated.

  1. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    SciTech Connect

    Not Available

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  2. BE Ursae Majoris: A detached binary with a unique reprocessing spectrum

    NASA Technical Reports Server (NTRS)

    Ferguson, Donald H.; Liebert, James; Cutri, Roc; Green, Richard F.; Willner, S. P.; Steiner, Joao E.; Tokarz, Susan

    1987-01-01

    New infrared photometry, optical and UV spectrophotometry, and a photographic ephemeris are presented for the detached binary BE UMa. Results show the primary to be a DO white dwarf with an effective temperature of 80,000 + or - 15,000 K and a mass of 0.6 + or - 0.1 solar masses. No evidence is found for variability of the primary. The main sequence secondary star is shown to be of early M spectral type, with a formal range of M1 to M5 being possible. A reflection effect in reprocessed line and continuum radiation is produced by EUV radiation from the primary incident on the secondary atmosphere. It is suggested that the temperature of the reprocessed component of the secondary's atmosphere is in the 5000 to 8500 K range, and that emission lines of decreasing ionization form deeper in the irradiated envelope. Relatively narrow He II and high excitation metal lines are formed from recombination and continuum fluorescence processes.

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

    SciTech Connect

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

    2003-02-25

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

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

    SciTech Connect

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

    1996-09-01

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

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

    SciTech Connect

    Schwantes, Jon M.; Sweet, Lucas E.

    2012-10-01

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

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

    SciTech Connect

    Susan Stacy; Julie Braun

    2006-12-01

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

  7. Flowsheet Analysis of U-Pu Co-Crystallization Process as a New Reprocessing System

    SciTech Connect

    Shunji Homma; Jun-ichi Ishii; Jiro Koga; Shiro Matsumoto; Toshiaki Kikuchi; Takahiro Chikazawa; Atsuhiro Shibata

    2006-07-01

    A new fuel reprocessing system by U-Pu co-crystallization process is proposed and examined by flowsheet analysis. This reprocessing system is based on the fact that hexavalent plutonium in nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate does not exist in the solution. The system consists of five steps: dissolution of spent fuel, plutonium oxidation, U-Pu co-crystallization as a co-decontamination, re-dissolution of the crystals, and U re-crystallization as a U-Pu separation. The system requires a recycling of the mother liquor from the U-Pu co-crystallization step and the appropriate recycle ratio is determined by flowsheet analysis such that the satisfactory decontamination is achieved. Further flowsheet study using four different compositions of LWR spent fuels demonstrates that the constant ratio of plutonium to uranium in mother liquor from the re-crystallization step is achieved for every composition by controlling the temperature. It is also demonstrated by comparing to the Purex process that the size of the plant based on the proposed system is significantly reduced. (authors)

  8. Dome Degradation Characterization of Wide-Field-of-View Nonscanner Aboard ERBE and Its Reprocessing

    NASA Technical Reports Server (NTRS)

    Shrestha, Alok K.; Kato, Seiji; Wong, Takmeng; Su, Wenying; Stackhouse, Paul W., Jr.; Rose, Fred; Miller, Walter F.; Bush, Kathryn; Rutan, David A.; Minnis, Patrick; Doelling, David R.; Smith, George L.

    2015-01-01

    Earth Radiation Budget Experiment (ERBE) wide-field-of-view (WFOV) nonscanners aboard ERBS and NOAA- 9/NOAA-10 provided broadband shortwave and longwave irradiances from 1985 to 1999. The previous analysis showed dome degradation in the shortwave nonscanner instruments. The correction was performed with a constant spectral (gray assumption) degradation. We suspect that the gray assumption affected daytime longwave irradiance and led to a day-minus-night longwave flux differences (little change in night time longwave) increase over time. Based on knowledge from the CERES process, we will reprocess entire ERBE nonscanner radiation dataset by characterizing shortwave dome transmissivity with spectral dependent degradation using the solar data observed by these instruments. Once spectral dependent degradation is derived, imager derived cloud fraction and the cloud phase as well as surface type over the FOV of nonscanner instruments will be used to model unfiltering coefficients. This poster primarily explains the reprocessing techniques and includes initial comparison of several months of data processed with existing and our recent methods.

  9. Iodine Pathways and Off-Gas Stream Characteristics for Aqueous Reprocessing Plants – A Literature Survey and Assessment

    SciTech Connect

    R. T. Jubin; D. M. Strachan; N. R. Soelberg

    2013-09-01

    Used nuclear fuel is currently being reprocessed in only a few countries, notably France, England, Japan, and Russia. The need to control emissions of the gaseous radionuclides to the air during nuclear fuel reprocessing has already been reported for the entire plant. But since the gaseous radionuclides can partition to various different reprocessing off-gas streams, for example, from the head end, dissolver, vessel, cell, and melter, an understanding of each of these streams is critical. These off-gas streams have different flow rates and compositions and could have different gaseous radionuclide control requirements, depending on how the gaseous radionuclides partition. This report reviews the available literature to summarize specific engineering data on the flow rates, forms of the volatile radionuclides in off-gas streams, distributions of these radionuclides in these streams, and temperatures of these streams. This document contains an extensive bibliography of the information contained in the open literature.

  10. Realistic Challenges to Optimize the Closed Nuclear Fuel Cycle

    SciTech Connect

    Shigeo Nomura; Hisao Ojima; Hirohisa Ishikawa

    2002-07-01

    The first generation nuclear fuel cycle system will be closed in Japan by 2010 with the successful combination of LWR-MOX, industrial operation of reprocessing and MOX fuel fabrication plants, interim storage of spent fuel, and plans for the location of final disposal of radioactive wastes. To tailor nuclear power as the globally sustainable and environmentally benign energy source for future society, however, an advanced fuel cycle option is needed that is well suited to the fast reactor. It should recycle fissile materials of plutonium and uranium whilst transmuting minor actinides and generating minimum waste. (authors)

  11. INSTRUMENTS AND METHODS OF INVESTIGATION: Electrophotonuclear energy cycle

    NASA Astrophysics Data System (ADS)

    Eremeev, Igor P.

    2004-12-01

    It is shown that the current problems related to the nuclear power cycle can be resolved if the conventional fission process is complemented with the 'photoneutron reprocessing' of fissile nuclei directly in spent uranium by using the energy of relativistic storage-ring electrons and transforming it to synchrotron/undulator gamma radiation, and by utilizing the fission products as neutron sources while at the same time transmuting them into stable isotopes. A suitable method of 'gamma-neutron transmutation' is described, and the values of its parameters that make the corresponding 'electrophotonuclear' power cycle closed and waste-free are determined.

  12. Advanced Fuel Cycle Cost Basis

    SciTech Connect

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  13. Advanced Fuel Cycle Cost Basis

    SciTech Connect

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  14. Advanced Fuel Cycle Cost Basis

    SciTech Connect

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

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

    NASA Astrophysics Data System (ADS)

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

    2013-12-01

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

  16. Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

    SciTech Connect

    Forsberg, C.W.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' (Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety) is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document.

  17. Fatigue crack initiation in carbon and low-alloy steels in light water reactor environments : mechanism and prediction.

    SciTech Connect

    Chopra, O. K.; Shack, W. J.

    1998-01-27

    Section 111 of the ASME Boiler and Pressure Vessel Code specifies fatigue design curves for structural materials. The effects of reactor coolant environments are not explicitly addressed by the Code design curves. Recent test data illustrate potentially significant effects of light water reactor (LWR) coolant environments on the fatigue resistance of carbon and low-alloy steels. Under certain loading and environmental conditions, fatigue lives of test specimens may be shorter than those in air by a factor of {approx}70. The crack initiation and crack growth characteristics of carbon and low-alloy steels in LWR environments are presented. Decreases in fatigue life of these steels in high-dissolved-oxygen water are caused primarily by the effect of environment on growth of short cracks < 100 {micro}m in depth. The material and loading parameters that influence fatigue life in LWR environments are defined. Fatigue life is decreased significantly when five conditions are satisfied simultaneously, viz., applied strain range, service temperature, dissolved oxygen in water, and S content in steel are above a threshold level, and loading strain rate is below a threshold value. Statistical models have been developed for estimating the fatigue life of these steels in LWR environments. The significance of the effect of environment on the current Code design curve is evaluated.

  18. Human-In-The-Loop Simulation in Support of Long-Term Sustainability of Light Water Reactors

    SciTech Connect

    Hallbert, Bruce P

    2015-01-01

    Reliable instrumentation, information, and control systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration. The NPP owners and operators realize that this analog technology represents a significant challenge to sustaining the operation of the current fleet of NPPs. Beyond control systems, new technologies are needed to monitor and characterize the effects of aging and degradation in critical areas of key structures, systems, and components. The objective of the efforts sponsored by the U.S. Department of Energy is to develop, demonstrate, and deploy new digital technologies for II&C architectures and provide monitoring capabilities to ensure the continued safe, reliable, and economic operation of the nation’s NPPs.

  19. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

    NASA Astrophysics Data System (ADS)

    Yamamoto, Y.; Pint, B. A.; Terrani, K. A.; Field, K. G.; Yang, Y.; Snead, L. L.

    2015-12-01

    Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10-20Cr, 3-5Al, and 0-0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitive to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741 °C.

  20. Human-In-The-Loop Simulation in Support of Long-Term Sustainability of Light Water Reactors

    DOE PAGES

    Hallbert, Bruce P

    2015-01-01

    Reliable instrumentation, information, and control systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration. The NPP owners and operators realize that this analog technology represents a significant challenge to sustaining the operation of the current fleet of NPPs. Beyond control systems, new technologies are neededmore » to monitor and characterize the effects of aging and degradation in critical areas of key structures, systems, and components. The objective of the efforts sponsored by the U.S. Department of Energy is to develop, demonstrate, and deploy new digital technologies for II&C architectures and provide monitoring capabilities to ensure the continued safe, reliable, and economic operation of the nation’s NPPs.« less

  1. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

    DOE PAGES

    Yamamoto, Yukinori; Pint, Bruce A.; Terrani, Kurt A.; ...

    2015-10-19

    Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10–20Cr, 3–5Al, and 0–0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitivemore » to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741°C.« less

  2. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

    SciTech Connect

    Yamamoto, Yukinori; Pint, Bruce A.; Terrani, Kurt A.; Field, Kevin G.; Yang, Ying; Snead, Lance Lewis

    2015-10-19

    Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10–20Cr, 3–5Al, and 0–0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitive to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741°C.

  3. Advanced light water reactor plants System 80+{trademark} design certification program. Annual progress report, October 1, 1995--September 30, 1996

    SciTech Connect

    1996-12-31

    The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1996 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2 and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems.

  4. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    SciTech Connect

    Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A; Matlack, Katie; Ramuhalli, Pradeep; Light, Glenn

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  5. Nuclear Islands: International Leasing of Nuclear Fuel Cycle Sites to Provide Enduring Assurance of Peaceful Use

    SciTech Connect

    Paine, Christopher E.; Cochran, Thomas B.

    2010-11-01

    Current International Atomic Energy Agency safeguards do not provide adequate protection against the diversion to military use of materials or technology from certain types of sensitive nuclear fuel cycle facilities. In view of highly enriched uranium’s relatively greater ease of use as a nuclear explosive material than plutonium and the significant diseconomies of commercial spent fuel reprocessing, this article focuses on the need for improved international controls over uranium enrichment facilities as the proximate justification for creation of an International Nuclear Fuel Cycle Association (INFCA). In principle, the proposal is equally applicable to alleviating the proliferation concerns provoked by nuclear fuel reprocessing plants and other sensitive nuclear fuel cycle facilities. The INFCA would provide significantly increased nonproliferation assurance to its member states and the wider international community by holding long-term leasehold contracts to operate secure restricted zones containing such sensitive nuclear facilities.

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

    SciTech Connect

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

    2006-07-01

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

  7. Concept for dismantling the Hllw treatment facility on the Former Wak Reprocessing Site

    SciTech Connect

    Birringer, K.J.; Fleisch, J.; Graffunder, I.; Pfeifer, W.

    2007-07-01

    The German pilot reprocessing plant WAK was operated until 1990 and processed about 200 tons of nuclear fuels from test and power reactors. In late 1991, the Federal Republic of Germany, the State of Baden-Wuerttemberg, and the utilities decided to shut down the WAK and to dismantle it completely to the green field. In the years 2000/2001, remote-controlled dismantling of the process cells in the reprocessing building was completed. Part of the building has already been subjected to release measurement and released from the obligations under the German Atomic Energy Act. However, a major prerequisite for the complete dismantling of the WAK is the management of the 60 m{sup 3} high-level liquid waste (HLLW) with an activity of 8.0 E 17 Bq resulting from reprocessing. For this purpose, the Karlsruhe vitrification plant (VEK) was constructed and is now under commissioning /1/. Hot operation is foreseen for the years 2007/2008. Following vitrification operation, dismantling of the four HLLW tanks in the storage building will be a particularly challenging task in terms of radiology. The HLLW tanks are located in thick-walled concrete cells that require remote- controlled horizontal access. For this purpose, a new access building, the southern extension, was built. It serves to bring in and operate the remote handling tools and allows for the contamination-safe removal and measurement of the MAW drums. In contrast to the crane in the process building, the manipulator carrier system used here is an 8 Mg excavator. All tools, including the wall cutter, chisel, cutting disk, scissors, and the electric master-slave manipulator (EMSM), can be docked to this excavator. The VEK installations shall be dismantled parallel to the HLLW storage tanks. Due to the dose rates expected after operation, two dismantling areas have to be distinguished in the VEK: The core area with the HLLW transfer cell, melter cell, and exhaust gas cell requires remote dismantling. All remaining cells

  8. Improved methods for reprocessing of GNSS data for climate monitoring over Poland

    NASA Astrophysics Data System (ADS)

    Stepniak, Katarzyna; Bock, Olivier; Wielgosz, Pawel

    2016-04-01

    The goal of this work is to determine the most accurate and homogeneous processing strategy to reprocess ground-based GNSS data for climate monitoring applications (analysis of trends and variability of Zenith Total Delay, ZTD, and Integrated Water Vapor, IWV). Namely, we investigate the impact of network design strategy and tropospheric modeling approach on the quality and homogeneity of both relative (double difference) and absolute (PPP) solutions. A network of 138 GNSS stations (including 33 stations from the EUREF Permanent Network, EPN, and 105 stations from ASG-EUPOS in Poland) is reprocessed for year 2014 using Bernese 5.2 GNSS software with the final IGS (International GNSS Service) orbits and clocks. First a standard (the shortest) "star" baseline design strategy is used in which the EPN stations are connected together defining a reference network and every ASG-EUPOS station is connected to the nearest EPN station. The initial network is modified automatically by the Bernese software every day depending on the availability of observations at the EPN stations. We show that in case of sub-daily gaps in the measurements of the reference stations, small clusters of stations can be disconnected from the main reference network. As a result, offsets of a few centimeters in ZTD estimates and spikes in formal errors can appear. These offsets and spikes cannot always be detected. This phenomenon is quite frequent in a large network such as considered in this study. It is also responsible for significant discontinuities in the estimated ZTD series which are detrimental to climate monitoring applications. We developed a new baseline design strategy algorithm to circumvent this event and assure that all the stations remain connected to the main reference network. It is shown that using this strategy, the reprocessed ZTD series are much more continuous and homogeneous in comparison to the standard strategy. The results are further validated against a Precise Point

  9. Menstrual Cycle

    MedlinePlus

    ... receive Pregnancy email updates Enter email Submit The menstrual cycle Day 1 starts with the first day of ... drop around Day 25 . This signals the next menstrual cycle to begin. The egg will break apart and ...

  10. Process agitator operating problems and equipment failures, F-Canyon Reprocessing Facility

    SciTech Connect

    Durant, W.S.; Starks, J.B.; Low, J.M.; Galloway, W.D.

    1988-09-01

    The Savannah River Laboratory (SRL) maintains a compilation of operating problems and equipment failures that have occurred in the fuel reprocessing areas of the Savannah River Plant (SRP). At present, the data bank contains more than 200,000 entries ranging from minor equipment malfunctions to incidents with the potential for injury or contamination of personnel, or for economic loss. The data bank has been used extensively for a wide variety of purposes, such as failure analyses, trend analyses, and preparation of safety analyses. Typical of the data are problems associated with the F-Canyon process agitators. This report contains a compilation of the agitator operating problems and equipment failures primarily as an aid to organizations with related equipment. Publication of these data was prompted by a number of requests for this information by other Department of Energy (DOE) sites. 4 figs., 4 tabs.

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

    DOEpatents

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

    1982-01-19

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

  12. Reprocessing of Lunar occultations observations carried out in Uzbekistan during 1882-1996

    NASA Astrophysics Data System (ADS)

    Muminov, M. M.; Kazantseva, L. V.; Ehgamberdiev, Sh. A.; Kahhorov, B. B.; Andruk, V. M.

    A unified database of ILOC and Kyiv Observatory by selected locations on the territory of Uzbekistan was presented. Reprocessing of observations using the semi-analytical ephemeris of lunar motion VSOP87A with obtained correction for the relief of the marginal zone according to Watts’s map was carried out. Observation of ocultations was carried out by the staff of Tashkent Observatory in Uzbekistan from the end of the XIX century. The observations are planned to be continued with 48 cm telescope of Maidanak Observatory. Selected database (1145 results of observations) was obtained during a long period (1882-1993 AD), but distribution of observation over that period is not uniform. 4 periods of an active observation 1882-1890, 1927-1938, 1948-1972 and 1993 can be noted.

  13. Modeling the dynamics of evaluation: a multilevel neural network implementation of the iterative reprocessing model.

    PubMed

    Ehret, Phillip J; Monroe, Brian M; Read, Stephen J

    2015-05-01

    We present a neural network implementation of central components of the iterative reprocessing (IR) model. The IR model argues that the evaluation of social stimuli (attitudes, stereotypes) is the result of the IR of stimuli in a hierarchy of neural systems: The evaluation of social stimuli develops and changes over processing. The network has a multilevel, bidirectional feedback evaluation system that integrates initial perceptual processing and later developing semantic processing. The network processes stimuli (e.g., an individual's appearance) over repeated iterations, with increasingly higher levels of semantic processing over time. As a result, the network's evaluations of stimuli evolve. We discuss the implications of the network for a number of different issues involved in attitudes and social evaluation. The success of the network supports the IR model framework and provides new insights into attitude theory.

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

    NASA Astrophysics Data System (ADS)

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

    2005-12-01

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

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

    DOEpatents

    Heckman, R.A.

    1980-12-19

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

  16. Study on release and transport of aerial radioactive materials in reprocessing plants

    SciTech Connect

    Amano, Y.; Tashiro, S.; Uchiyama, G.; Abe, H.; Yamane, Y.; Yoshida, K.; Kodama, T.

    2013-07-01

    The release and transport characteristics of radioactive materials at a boiling accident of the high active liquid waste (HALW) in a reprocessing plant have been studied for improving experimental data of source terms of the boiling accident. In the study, a heating test and a thermogravimetry and differential thermal analysis (TG-DTA) test were conducted. In the heating test using a simulated HALW, it was found that ruthenium was mainly released into the air in the form of gas and that non-volatile elements were released into the air in the form of mist. In the TG-DTA test, the rate constants and reaction heat of thermal decomposition of ruthenium nitrosyl nitrate were obtained from TG and DTA curves. (authors)

  17. Reprocessing of Soft X-ray Emission Lines in Black Hole Accretion Disks

    SciTech Connect

    Mauche, C W; Liedahl, D A; Mathiesen, B F; Jimenez-Garate, M A; Raymond, J C

    2003-10-17

    By means of a Monte Carlo code that accounts for Compton scattering and photoabsorption followed by recombination, we have investigated the radiation transfer of Ly{alpha}, He{alpha}, and recombination continua photons of H- and He-like C, N, O, and Ne produced in the photoionized atmosphere of a relativistic black hole accretion disk. We find that photoelectric opacity causes significant attenuation of photons with energies above the O VIII K-edge; that the conversion efficiencies of these photons into lower-energy lines and recombination continua are high; and that accounting for this reprocessing significantly (by factors of 21% to 105%) increases the flux of the Ly{alpha} and He{alpha} emission lines of H- and He-like C and O escaping the disk atmosphere.

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

    DOEpatents

    Forsberg, C.W.

    1992-03-24

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

  19. Ozone budget in the upper stratosphere: Model studies using the reprocessed LIMS and the HALOE datasets

    NASA Astrophysics Data System (ADS)

    Natarajan, Murali; Remsberg, Ellis E.; Gordley, Larry L.

    2002-04-01

    Recently reprocessed LIMS dataset has been used with a contemporary photochemical model to study the balance between photochemical production and destruction of ozone in the upper stratosphere. Model results corresponding to January 1979 indicate that the ozone deficit is less than 15% in the pressure range of 5 to 0.5 mb between 50°S and 50°N latitude. The imbalance at 40 km is much smaller than reported by the earliest studies with the archived LIMS data. The same model, when initialized with HALOE (version 19) data for January, 1996, shows similar results with peak ozone deficits being less than 10%. For both cases, the model shows a near balance in the ozone budget above 1 mb, contrary to recent studies based on balloon-borne measurements. The magnitude of the ozone imbalance seen in this study is within the uncertainties of the data and model.

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

    SciTech Connect

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

    1991-01-01

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

  1. Eye movement desensitization and reprocessing therapy for personality disorders in older adults?

    PubMed

    Gielkens, E M J; Sobczak, S; Van Alphen, S P J

    2016-10-01

    Eye Movement Desensitization and Reprocessing (EMDR) is a kind of psychotherapy, which is growing in popularity, particularly for treatment of post-traumatic stress disorder (PTSD). When Shapiro first introduced EMDR in 1989, it was approached as a controversial treatment because of lack of evidence. However, nowadays there is growing evidence for EMDR efficacy in PTSD (Mc Guire et al., 2014) and EMDR is recommended by international and national treatment guidelines for PTSD. Moreover, EMDR is also used for the treatment of other anxiety disorders, such as panic disorders (De Jongh et al., 2002). Furthermore, research continues on effects of EMDR in addiction, somatoform disorders and psychosis. So far, there is no empirical research on the efficacy of EMDR treatment in older adults.

  2. Eye movement desensitization and reprocessing in an adolescent with epilepsy and mild intellectual disability.

    PubMed

    Rodenburg, Roos; Benjamin, Anja; Meijer, Anne Marie; Jongeneel, Ruud

    2009-09-01

    Intellectual disability is a comorbid condition in epilepsy. People with epilepsy and intellectual disability are at high risk of developing behavioral problems. Among the many contributors to behavioral problems in people with epilepsy and intellectual disability are those of traumatic experiences. As such, behavioral problems can be seen as a reflection of these traumatic experiences. Among established trauma therapies, eye movement desensitization and reprocessing (EMDR) is an emerging treatment that is effective in adults and also seems to be effective in children. This article is a case report of EMDR in an adolescent with epilepsy and mild intellectual disability, in whom the EMDR children's protocol was used. The aim was to assess whether clinical trauma status significantly diminished to nonclinical status posttreatment. Change in trauma symptoms was evaluated with the Reliable Change Index (RCI). Results showed a significant decrease in trauma symptoms toward nonclinical status from pretreatment to posttreatment. EMDR consequences for epilepsy and intellectual disability are discussed.

  3. [Psychotherapeutic treatment of traumatic stress with the EMDR (Eye Movement Desensitization and Reprocessing) method].

    PubMed

    Liebermann, P; Hofmann, A; Flatten, G

    2003-12-04

    EMDR (Eye Movement Desensitization and Reprocessing) is a method, developed at the end of the nineteen-eighties, for the treatment of the post-traumatic stress disorder (PTSD). The patient is asked to concentrate on certain aspect of the traumatic memory while keeping his eyes fixed on the movements of the therapist's finger. Apparently, this triggers information processing that results in appreciable relief for the patient. The method has proven to be equally as effective as behavioral-therapeutic techniques, and, has in the meantime, been included in national and international guidelines for the treatment of PTSD. The indications for EMDR treatment include not only PTSD, but, increasingly, also other, sometimes more severely chronic, it trauma sequelae. Within the framework of basic care, EMDR must be embedded within a treatment plan and should, where applicable, be combined with other methods.

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

    SciTech Connect

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

    1993-12-31

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

  5. The Efficacy of Eye Movement Desensitization and Reprocessing Therapy Technique in the Treatment of Test Anxiety of College Students.

    ERIC Educational Resources Information Center

    Enright, Matthew; Baldo, Tracy D.; Wykes, Scott D.

    2000-01-01

    Explores the efficacy of eye movement desensitization and reprocessing (EMDR) in the treatment of test anxiety. Thirty-five college students with test anxiety were assigned to either a treatment or delayed treatment control group. EMDR was shown to be effective in reducing overall test anxiety as well as "emotionality" and…

  6. 76 FR 24494 - Draft Guidance for Industry and FDA Staff: Processing/Reprocessing Medical Devices in Health Care...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-02

    ... HUMAN SERVICES Food and Drug Administration Draft Guidance for Industry and FDA Staff: Processing...: Food and Drug Administration, HHS. ACTION: Notice. SUMMARY: The Food and Drug Administration (FDA) is... with processing or reprocessing labeling. This draft guidance is not final; nor is it in effect at...

  7. DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term Operations Program. Joint Research and Development Plan

    SciTech Connect

    Williams, Don

    2014-04-01

    Nuclear power has contributed almost 20% of the total amount of electricity generated in the United States over the past two decades. High capacity factors and low operating costs make nuclear power plants (NPPs) some of the most economical power generators available. Further, nuclear power remains the single largest contributor (nearly 70%) of non-greenhouse gas-emitting electric power generation in the United States. Even when major refurbishments are performed to extend operating life, these plants continue to represent cost-effective, low-carbon assets to the nation’s electrical generation capability. By the end of 2014, about one-third of the existing domestic fleet will have passed their 40th anniversary of power operations, and about one-half of the fleet will reach the same 40-year mark within this decade. Recognizing the challenges associated with pursuing extended service life of commercial nuclear power plants, the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) and the Electric Power Research Institute (EPRI) have established separate but complementary research and development programs (DOE-NE’s Light Water Reactor Sustainability [LWRS] Program and EPRI’s Long-Term Operations [LTO] Program) to address these challenges. To ensure that a proper linkage is maintained between the programs, DOE-NE and EPRI executed a memorandum of understanding in late 2010 to “establish guiding principles under which research activities (between LWRS and LTO) could be coordinated to the benefit of both parties.” This document represents the third annual revision to the initial version (March 2011) of the plan as called for in the memorandum of understanding.

  8. A MELCOR Application to Two Light Water Reactor Nuclear Power Plant Core Melt Scenarios with Assumed Cavity Flooding Action

    SciTech Connect

    Martin-Fuertes, Francisco; Martin-Valdepenas, Juan Manuel; Mira, Jose; Sanchez, Maria Jesus

    2003-10-15

    The MELCOR 1.8.4 code Bottom Head package has been applied to simulate two reactor cavity flooding scenarios for when the corium material relocates to the lower-plenum region in postulated severe accidents. The applications were preceded by a review of two main physical models, which highly impacted the results. A model comparison to available bibliography models was done, which allowed some code modifications on selected default assumptions to be undertaken. First, the corium convective heat transfer to the wall when it becomes liquid was modified, and second, the default nucleate boiling regime curve in a submerged hemisphere was replaced by a new curve (and, to a much lesser extent, the critical heat flux curve was slightly varied).The applications were devoted to two prototypical light water reactor nuclear power plants, a 2700-MW(thermal) pressurized water reactor (PWR) and a 1381-MW(thermal) boiling water reactor (BWR). The main conclusions of the cavity flooding simulations were that the PWR lower-head survivability is extended although it is clearly not guaranteed, while in the BWR sequence the corium seems to be successfully arrested in the lower plenum.Three applications of the CFX 4.4 computational fluid dynamics code were carried out in the context of the BWR scenario to support the first modification of the aforementioned two scenarios for MELCOR.Finally, in the same BWR context, a statistic predictor of selected output parameters as a function of input parameters is presented, which provides reasonable results when compared to MELCOR full calculations in much shorter CPU processing times.

  9. Advanced Light Water Reactor Plants System 80+{trademark} Design Certification Program. Annual progress report, October 1, 1992--September 30, 1993

    SciTech Connect

    Not Available

    1993-12-31

    The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1993 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW{sub t} (1350 MWe) Pressurized Water Reactor (PWR). The design consists of an essentially complete plant. It is based on evolutionary improvements to the Standardized System 80 nuclear steam supply system in operation at Palo Verde Units 1, 2, and 3, and the Duke Power Company P-81 balance-of-plant (BOP) that was designed and partially constructed at the Cherokee plant site. The System 80/P-81 original design has been substantially enhanced to increase conformance with the EPRI ALWR Utility Requirements Document (URD). Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units form the basis of the Korean standardization program. The full System 80+ standard design has been offered to the Republic of China, in response to their recent bid specification. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was submitted to the NRC and a Draft Safety Evaluation Report was issued by the NRC in October 1992. CESSAR-DC contains the technical basis for compliance with the EPRI URD for simplified emergency planning. The Nuclear Steam Supply System (NSSS) is the standard ABB-Combustion Engineering two-loop arrangement with two steam generators, two hot legs and four cold legs each with a reactor coolant pump. The System 80+ standard plant includes a sperical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual containment.

  10. Transmutation Performance Analysis for Inert Matrix Fuels in Light Water Reactors and Computational Neutronics Methods Capabilities at INL

    SciTech Connect

    Michael A. Pope; Samuel E. Bays; S. Piet; R. Ferrer; Mehdi Asgari; Benoit Forget

    2009-05-01

    The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF, thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multi-pass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed. These analyses provide comparison of direct thermal recycling of PuNpAmCmCf, PuNpAm, PuNp, and Pu. The results of this infinite lattice assembly-wise study using SCALE 5.1 indicate that it may be feasible to recycle TRU in this manner using an otherwise typical PWR assembly without violating peaking factor limits.

  11. Validation efforts for the neutronics of a plutonium erbium zirconium oxide inert matrix light water reactor fuel

    NASA Astrophysics Data System (ADS)

    Paratte, J. M.; Chawla, R.; Früh, R.; Joneja, O. P.; Pelloni, S.; Pralong, C.

    1999-08-01

    Light water reactor (LWR) neutronics codes and cross-section libraries need further qualification when used for the calculation of inert matrix fuel (IMF) cells. Three types of validation efforts have been undertaken for the PuO 2-Er 2O 3-ZrO 2 IMF concept under development at the Paul Scherrer Institute (PSI). Firstly, the PSI calculational scheme, based on the BOXER code and its data library, has been applied to the analysis of a range of LWR experiments with PuO 2-UO 2 fuel, conducted earlier at PSI's PROTEUS facility. The generally good agreement obtained between calculated and measured parameters gives confidence in the ability of the employed calculational scheme to correctly modelize Pu-containing fuel cells. Secondly, reactivity effects of various burnable poisons in a ZrO 2 matrix were measured in the CROCUS reactor of the Swiss Federal Institute of Technology at Lausanne. Modelling these experiments with BOXER resulted in satisfactory prediction of measured reactivity ratios (relative to a soluble-boron standard) for most of the experimental rods employed. This was particularly the case for experiments with erbium, as well as with mixtures of erbium and europium (the latter being used to simulate the effects of overlapping resonances, as would be expected in the case of a Pu-Er IMF). Finally, as there are no experimental results available from power reactors employing IMFs, the validation of burnup calculations (at the cell level) has been based on results obtained in the framework of an international benchmark exercise on the physics of LWRs employing IMFs. Certain discrepancies in calculated parameters have been observed in this context, several of which can be attributed to specific differences in cross-section libraries.

  12. Light Water Reactor Sustainability Program: Computer-Based Procedures for Field Activities: Results from Three Evaluations at Nuclear Power Plants

    SciTech Connect

    Oxstrand, Johanna; Le Blanc, Katya; Bly, Aaron

    2014-09-01

    The Computer-Based Procedure (CBP) research effort is a part of the Light-Water Reactor Sustainability (LWRS) Program, which is a research and development (R&D) program sponsored by Department of Energy (DOE) and performed in close collaboration with industry R&D programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants. One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. One area that could yield tremendous savings in increased efficiency and safety is in improving procedure use. Nearly all activities in the nuclear power industry are guided by procedures, which today are printed and executed on paper. This paper-based procedure process has proven to ensure safety; however, there are improvements to be gained. Due to its inherent dynamic nature, a CBP provides the opportunity to incorporate context driven job aids, such as drawings, photos, and just-in-time training. Compared to the static state of paper-based procedures (PBPs), the presentation of information in CBPs can be much more flexible and tailored to the task, actual plant condition, and operation mode. The dynamic presentation of the procedure will guide the user down the path of relevant steps, thus minimizing time spent by the field worker to evaluate plant conditions and decisions related to the applicability of each step. This dynamic presentation of the procedure also minimizes the risk of conducting steps out of order and/or incorrectly assessed applicability of steps.

  13. The Role and Reprocessing of Attitudes in Fostering Employee Work Happiness: An Intervention Study

    PubMed Central

    Williams, Paige; Kern, Margaret L.; Waters, Lea

    2017-01-01

    This intervention study examines the iterative reprocessing of explicit and implicit attitudes as the process underlying associations between positive employee attitudes (PsyCap), perception of positive organization culture (organizational virtuousness, OV), and work happiness. Using a quasi-experimental design, a group of school staff (N = 69) completed surveys at three time points. After the first assessment, the treatment group (n = 51) completed a positive psychology training intervention. Results suggest that employee PsyCap, OV, and work happiness are associated with one another through both implicit and explicit attitudes. Further, the Iterative-Reprocessing Model of attitudes (IRM) provides some insights into the processes underlying these associations. By examining the role and processes through which explicit and implicit attitudes relate to wellbeing at work, the study integrates theories on attitudes, positive organizational scholarship, positive organizational behavior and positive education. It is one of the first studies to apply the theory of the IRM to explain associations amongst PsyCap, OV and work happiness, and to test the IRM theory in a field-based setting. In applying attitude theory to wellbeing research, this study provides insights to mechanisms underlying workplace wellbeing that have not been previously examined and in doing so responds to calls for researchers to learn more about the mechanisms underlying wellbeing interventions. Further, it highlights the need to understand subconscious processes in future wellbeing research and to include implicit measures in positive psychology interventions measurement programs. Practically, this research calls attention to the importance of developing both the positive attitudes of employees and the organizational culture in developing employee work happiness. PMID:28154546

  14. Multivariate analysis of GPS position time series of JPL second reprocessing campaign

    NASA Astrophysics Data System (ADS)

    Amiri-Simkooei, A. R.; Mohammadloo, T. H.; Argus, D. F.

    2017-01-01

    The second reprocessing of all GPS data gathered by the Analysis Centers of IGS was conducted in late 2013 using the latest models and methodologies. Improved models of antenna phase center variations and solar radiation pressure in JPL's reanalysis are expected to significantly reduce errors. In an earlier work, JPL estimates of position time series, termed first reprocessing campaign, were examined in terms of their spatial and temporal correlation, power spectra, and draconitic signal. Similar analyses are applied to GPS time series at 89 and 66 sites of the second reanalysis with the time span of 7 and 21 years, respectively, to study possible improvements. Our results indicate that the spatial correlations are reduced on average by a factor of 1.25. While the white and flicker noise amplitudes for all components are reduced by 29-56 %, the random walk amplitude is enlarged. The white, flicker, and random walk noise amount to rate errors of, respectively, 0.01, 0.12, and 0.09 mm/yr in the horizontal and 0.04, 0.41 and 0.3 mm/yr in the vertical. Signals reported previously, such as those with periods of 13.63, 14.76, 5.5, and 351.4 / n for n=1,2,ldots,8 days, are identified in multivariate spectra of both data sets. The oscillation of the draconitic signal is reduced by factors of 1.87, 1.87, and 1.68 in the east, north and up components, respectively. Two other signals with Chandlerian period and a period of 380 days can also be detected.

  15. Multi-Technique Reprocessing and Combination using “Space-Ties”

    NASA Astrophysics Data System (ADS)

    Springer, T.; Dilssner, F.; Escobar, D.; Flohrer, C.; Otten, M.; Romero, I.; Dow, J. M.

    2009-12-01

    During 2008 and 2009 ESOC has reprocessed the historic data from the IDS, IGS, and ILRS! Our 3 solution were computed with one and the same software running on the same machine and using, as far as possible, identical settings, a unique achievement. Any systematic differences between the technique dependent reference frame solutions must therefore be caused by the techniques itself, and not because of model differences or errors! Our three technique dependent solutions should give us a very good understanding of the technique dependent effects. In the ITRF reference frame realisations the reference frame solutions from the individual techniques are combined using the local ties as measured at the stations. Besides possible errors and inaccuracies in these site-ties, this approach also suffers from the fact that the ILRS and IVS both offer very few stations with a far from global homogeneous distribution. At ESOC we have now made a small but significant step forward by including the SLR observations of the GNSS (GPS and GLONASS) satellites in our GNSS solutions. This allows to combine the ILRS and IGS reference frames by, besides using the site-ties at the stations, also using the “Space-Ties”. With space-ties we mean the vector between the position of the optical phase centre of the SLR laser reflector array and the GPS transmit antenna. Of course these “space-ties” are also not perfectly known but nevertheless they allow a much more rigorous combination of the two reference frames. And, very important for the GNSS technique, they allow the direct estimation of the GNSS satellite transmitter phase centre offset. In our presentation will give an overview of our IDS, IGS, and ILRS reprocessing efforts. We will conclude our presentation by demonstrating the benefits of “Space-Ties”.

  16. Regional seismic lines reprocessed using post-stack processing techniques; National Petroleum Reserve, Alaska

    USGS Publications Warehouse

    Miller, John J.; Agena, W.F.; Lee, M.W.; Zihlman, F.N.; Grow, J.A.; Taylor, D.J.; Killgore, Michele; Oliver, H.L.

    2000-01-01

    This CD-ROM contains stacked, migrated, 2-Dimensional seismic reflection data and associated support information for 22 regional seismic lines (3,470 line-miles) recorded in the National Petroleum Reserve ? Alaska (NPRA) from 1974 through 1981. Together, these lines constitute about one-quarter of the seismic data collected as part of the Federal Government?s program to evaluate the petroleum potential of the Reserve. The regional lines, which form a grid covering the entire NPRA, were created by combining various individual lines recorded in different years using different recording parameters. These data were reprocessed by the USGS using modern, post-stack processing techniques, to create a data set suitable for interpretation on interactive seismic interpretation computer workstations. Reprocessing was done in support of ongoing petroleum resource studies by the USGS Energy Program. The CD-ROM contains the following files: 1) 22 files containing the digital seismic data in standard, SEG-Y format; 2) 1 file containing navigation data for the 22 lines in standard SEG-P1 format; 3) 22 small scale graphic images of each seismic line in Adobe Acrobat? PDF format; 4) a graphic image of the location map, generated from the navigation file, with hyperlinks to the graphic images of the seismic lines; 5) an ASCII text file with cross-reference information for relating the sequential trace numbers on each regional line to the line number and shotpoint number of the original component lines; and 6) an explanation of the processing used to create the final seismic sections (this document). The SEG-Y format seismic files and SEG-P1 format navigation file contain all the information necessary for loading the data onto a seismic interpretation workstation.

  17. The Nuclear Energy Advanced Modeling and Simulation Safeguards and Separations Reprocessing Plant Toolkit

    SciTech Connect

    McCaskey, Alex; Billings, Jay Jay; de Almeida, Valmor F

    2011-08-01

    This report details the progress made in the development of the Reprocessing Plant Toolkit (RPTk) for the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. RPTk is an ongoing development effort intended to provide users with an extensible, integrated, and scalable software framework for the modeling and simulation of spent nuclear fuel reprocessing plants by enabling the insertion and coupling of user-developed physicochemical modules of variable fidelity. The NEAMS Safeguards and Separations IPSC (SafeSeps) and the Enabling Computational Technologies (ECT) supporting program element have partnered to release an initial version of the RPTk with a focus on software usability and utility. RPTk implements a data flow architecture that is the source of the system's extensibility and scalability. Data flows through physicochemical modules sequentially, with each module importing data, evolving it, and exporting the updated data to the next downstream module. This is accomplished through various architectural abstractions designed to give RPTk true plug-and-play capabilities. A simple application of this architecture, as well as RPTk data flow and evolution, is demonstrated in Section 6 with an application consisting of two coupled physicochemical modules. The remaining sections describe this ongoing work in full, from system vision and design inception to full implementation. Section 3 describes the relevant software development processes used by the RPTk development team. These processes allow the team to manage system complexity and ensure stakeholder satisfaction. This section also details the work done on the RPTk ``black box'' and ``white box'' models, with a special focus on the separation of concerns between the RPTk user interface and application runtime. Section 4 and 5 discuss that application runtime component in more detail, and describe the dependencies, behavior, and rigorous testing of its constituent components.

  18. A European perspective on the US nuclear fuel cycle

    SciTech Connect

    Wilkinson, W.L.

    1989-01-01

    Many Europeans believe that the main problems which have impeded progress in solving the back end of the nuclear fuel cycle in the United States have been a series of ideological and political hang-ups and these, coupled with excessive bureaucracy, have made logical decision making on the back-end problems impossible. This situation has been caused by a succession of political nondecisions. Public confidence in nuclear generation was thereby undermined and, because of plentiful supplies of other energy sources, there was no urgent need to expand the nuclear program in the United States. Since uranium was cheap and fast reactors not commercially attractive, there was no economic incentive to reprocess fuel from existing reactors in the United States. The problem facing the United States is that of managing the large stocks of spent fuel which have arisen over many years. A logical way forward for the United States would appear to be as follows: build more storage for spent fuel; consider overseas reprocessing to provide plutonium; develop reprocessing technology; and develop direct disposal technology.

  19. Reprocessable squeezing electrode fabrication of olive-like Fe/Co/O nanoparticle@three dimensional nitrogen-doped reduced graphene oxides for high performance lithium batteries.

    PubMed

    Qi, Li-Ya; Zhang, Yi-Wei; Xin, Yue-Long; Zuo, Zi-Cheng; Wu, Bin; Zhang, Xin-Xiang; Zhou, Heng-Hui

    2015-05-07

    A one step in situ synthesis approach is developed to construct 3D nitrogen-doped reduced graphene oxides, in which olive-like multi-component metal oxides are homogeneously dispersed. The novel hybrid nanoarchitecture shows some particular properties derived from synergistic effects. The size of Fe/Co/O oxides is reduced and better controlled compared to that of individual oxides due to mutual dispersant interactions. Furthermore, the positive synergistic interaction between heterogeneous oxides and graphene nanosheets has effective control on the particle size and dispersion of nanoparticles. Taking advantage of the flexibility and the cohesiveness of graphene nanosheets, the obtained composite can be directly processed into a binder-free electrode through a unique time-saving "squeezing" process. The obtained electrode possesses a reprocessable feature, which provides possibilities for convenient storage and quick fabrication at any time and presents attractive electrochemical performance of robust long-term capability retention (562 mA h g(-1) after 300 cycles at 10 A g(-1)) and superior rate performances (1162 mA h g(-1) at 0.5 A g(-1), 737 mA h g(-1) at 5 A g(-1), and 585 mA h g(-1) at 10 A g(-1)).

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

    SciTech Connect

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

    2003-02-25

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

  1. Reprocessing the Historical Satellite Passive Microwave Record at Enhanced Spatial Resolutions using Image Reconstruction

    NASA Astrophysics Data System (ADS)

    Hardman, M.; Brodzik, M. J.; Long, D. G.; Paget, A. C.; Armstrong, R. L.

    2015-12-01

    Beginning in 1978, the satellite passive microwave data record has been a mainstay of remote sensing of the cryosphere, providing twice-daily, near-global spatial coverage for monitoring changes in hydrologic and cryospheric parameters that include precipitation, soil moisture, surface water, vegetation, snow water equivalent, sea ice concentration and sea ice motion. Currently available global gridded passive microwave data sets serve a diverse community of hundreds of data users, but do not meet many requirements of modern Earth System Data Records (ESDRs) or Climate Data Records (CDRs), most notably in the areas of intersensor calibration, quality-control, provenance and consistent processing methods. The original gridding techniques were relatively primitive and were produced on 25 km grids using the original EASE-Grid definition that is not easily accommodated in modern software packages. Further, since the first Level 3 data sets were produced, the Level 2 passive microwave data on which they were based have been reprocessed as Fundamental CDRs (FCDRs) with improved calibration and documentation. We are funded by NASA MEaSUREs to reprocess the historical gridded data sets as EASE-Grid 2.0 ESDRs, using the most mature available Level 2 satellite passive microwave (SMMR, SSM/I-SSMIS, AMSR-E) records from 1978 to the present. We have produced prototype data from SSM/I and AMSR-E for the year 2003, for review and feedback from our Early Adopter user community. The prototype data set includes conventional, low-resolution ("drop-in-the-bucket" 25 km) grids and enhanced-resolution grids derived from the two candidate image reconstruction techniques we are evaluating: 1) Backus-Gilbert (BG) interpolation and 2) a radiometer version of Scatterometer Image Reconstruction (SIR). We summarize our temporal subsetting technique, algorithm tuning parameters and computational costs, and include sample SSM/I images at enhanced resolutions of up to 3 km. We are actively

  2. NASADEM Overview and First Results: Shuttle Radar Topography Mission (SRTM) Reprocessing and Improvements

    NASA Astrophysics Data System (ADS)

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

    2015-12-01

    NASADEM is a significant modernization of SRTM digital elevation model (DEM) data supported by the NASA MEaSUREs program. We are reprocessing the raw radar signal data using improved algorithms and incorporating ICESat and ASTER-derived DEM data unavailable during the original processing. The NASADEM products will be freely-available through the Land Processes Distributed Active Archive Center (LPDAAC) at 1-arcsecond spacing. The most significant processing improvements involve void reduction through improved phase unwrapping and using ICESat data for control. The updated unwrapping strategy now includes the use of SNAPHU for data processing patches where the unwrapped coverage from the original residue-based unwrapper falls below a coverage threshold. In North America continental processing, first experiments show the strip void area is reduced by more than 50% and the number of strip void patches is reduced by 40%. Patch boundary voids are mitigated by reprocessing with a different starting burst and merging the unwrapping results. We also updated a low-resolution elevation database to aid with unwrapping bootstrapping, retaining isolated component of unwrapped phase, and assessing the quality of the strip DEMs. We introduce a height ripple error correction to reduce artifacts in the strip elevation data. These ripples are a few meters in size with along-track spatial scales of tens of kilometers and are due to uncompensated mast motion most pronounced after Shuttle roll angle adjustment maneuvers. We developed an along-track filter utilizing differences between the SRTM heights and ICESat lidar elevation data. For a test using all data over North America, the algorithm reduced the ICESat-SRTM bias from 80 cm to 3 cm and the RMS from 5m to 4m. After merging and regridding the SRTM strip DEMs into 1x1-degree tiles, remaining voids are primarily filled with the ASTER-derived Global DEM. We use a Delta Surface Fill method to rubbersheet fill data across the void for

  3. Ancestral Structure of the Neuquén Basin, Supported by an Innovative Deep Seismic Reprocessing

    NASA Astrophysics Data System (ADS)

    Comínguez, A. H.; Franzese, J. R.

    2007-05-01

    Seismic-tracings comprising both the eastern and western sectors of Sierra de los Chihuidos, showed the deep structure of the Neuquén basin, Argentina. Deep reprocessing of historical industrial seismic-lines supplied interpretive information down to about 30-33 km. Consequently, seismic data reprocessed with "self-truncating extended correlation" confirmed an objective way for acquiring deep-seismic information where standard Vibroseis records are available. In addition, the FMED algorithm was an appreciated nonlinear mathematical tool to improve seismic resolution. Original results accomplished with the above emphasized techniques, revealed a list of concepts summarized along the subsequent comments. An acoustic contrast at about 24 km depth must be the top of the lower Crust. An oblique reflector between 16 and 18 km depth must be assumed as the local image of the master shear that controlled the extension system during the Late Triassic-Early Jurassic period. A sub-master fault dipping about 8° W, surely have been controlling the evolution of `Las Cárceles' area. An important inversion event initiated during the Bathonian-Callovian, sensibly affected the western sector of `Las Cárceles' (that is the site contiguous to the Neuquén river). Significant deposition of synrift sediments (Precuyo Group) originated in contiguous scarp degradation was detected on the western side of `Los Chihuidos' arch, at about 7 km depth. A Pliensbachian-Toarcian bipolar inversion developed during the transition to the Cuyo Group was evidenced in the western area. In the same sector, a middle Jurassic postrift episode is characterized by a deltaic depositional system prograding to the west with accentuate high energy. A deep discontinuity was related with the ancestral origin of the Basin, its seismic tracing permitted to match field results with a scale tank experiment simulating orogenic collapse. Bulk extension of the ancestral thickened crust could be only justified if a

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

    PubMed

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

    2015-05-19

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

  5. Waste-Incidental-to-Reprocessing Evaluation for the West Valley Demonstration Project Vitrification Melter - 12167

    SciTech Connect

    McNeil, Jim; Kurasch, David; Sullivan, Dan; Crandall, Thomas

    2012-07-01

    The Department of Energy (DOE) has determined that the vitrification melter used in the West Valley Demonstration Project can be disposed of as low-level waste (LLW) after completion of a waste-incidental-to-reprocessing evaluation performed in accordance with the evaluation process of DOE Manual 435.1-1, Radioactive Waste Management Manual. The vitrification melter - which consists of a ceramic lined, electrically heated box structure - was operated for more than 5 years melting and fusing high-level waste (HLW) slurry and glass formers and pouring the molten glass into 275 stainless steel canisters. Prior to shutdown, the melter was decontaminated by processing low-activity decontamination flush solutions and by extracting molten glass from the melter cavity. Because it could not be completely emptied, residual radioactivity conservatively estimated at approximately 170 TBq (4,600 Ci) remained in the vitrification melter. To establish whether the melter was incidental to reprocessing, DOE prepared an evaluation to demonstrate that the vitrification melter: (1) had been processed to remove key radionuclides to the maximum extent technically and economically practical; (2) would be managed to meet safety requirements comparable to the performance objectives for LLW established by the Nuclear Regulatory Commission (NRC); and (3) would be managed by DOE in accordance with DOE's requirements for LLW after it had been incorporated in a solid physical form with radionuclide concentrations that do not exceed the NRC concentration limits for Class C LLW. DOE consulted with the NRC on the draft evaluation and gave other stakeholders an opportunity to submit comments before the determination was made. The NRC submitted a request for additional information in connection with staff review of the draft evaluation; DOE provided the additional information and made improvements to the evaluation, which was issued in January 2012. DOE considered the NRC Technical Evaluation Report

  6. First Indications of Intraplate Deformations in Central Germany from Reprocessed GNSS Time Series and Geological Data

    NASA Astrophysics Data System (ADS)

    Becker, Matthias; Leinen, Stefan; Läufer, Gwendolyn; Lehné, Rouwen

    2013-04-01

    Six years of GPS data have been reprocessed in ITRF2008 for a regional SAPOS CORS network in the federal state of Hesse with 25 stations and some anchor sites of IGS and EPN to derive accurate and consistent coordinate time series. Based on daily network solutions coordinate time series parameters like velocities, offsets in case of antenna changes and annual periodic variation have been estimated. The estimation process includes the fitting of a sophisticated stochastic model for the time series which accounts for inherent time correlation. The results are blended with geological data to verify information from geology on potential recent deformations by the geodetic analyses. Besides of some information on the reprocessing of the GNSS the results the stochastics of the derived velocity field will be discussed in detail. Special emphasis will be on the intra-plate deformation: for the horizontal component the residual velocity field after removal of a plate rotation model is presented, while for the vertical velocities the datum-induced systematic effect is removed in order to analyze the remaining vertical motion. The residual velocity field is then matched with the geology for Hesse. Correlation of both vertical and horizontal movements with major geological structures reveals good accordance. SAPOS stations with documented significant subsidence are mainly located in tertiary Graben structures such as the Lower Hessian Basin (station Kassel), the Wetterau (station Kloppenheim) or the Upper Rhine Graben (Station Darmstadt). From the geological point of view these structures are supposed to be subsiding ones. Other major geological features, i.e. the Rhenish Shield as well as the East Hessian Bunter massif are supposed to be affected by recent uplift. SAPOS stations located in these regions match the assumed movement (e.g. Weilburg, Wiesbaden, Bingen, Fulda). Furthermore SAPOS-derived horizontal movements seem to trace tectonic movements in the region, i

  7. Safety demonstration tests of hypothetical explosive burning in the cell and air ventilation system in a nuclear fuel reprocessing plant

    SciTech Connect

    Nisio, G.; Suzuki, M.; Mukaide, S. )

    1991-09-01

    This paper reports on a nuclear fuel reprocessing plant equipped with an air ventilation system consisting of cells, ducts, dampers, high-efficiency particulate air filters, and blowers. This ventilation system is required to have multiple safeguards in order to confine airborne radioactive materials within the plant in the event of fire, explosion, and criticality. To evaluate these safeguards, three kinds of explosive burning tests are performed using a large-scale facility simulating the ventilation system of a reprocessing plant. In the boilover test, an organic solvent is burned on a layer of water in a burning pan to determine the magnitude of the burning caused by the sudden boiling of the water under the solvent. The optimum conditions for boilover burning are determined by the relationship between the pan size and the ventilation rate.

  8. Next-generation purex flowsheets with acetohydroxamic acid as complexant for FBR and thermal-fuel reprocessing

    SciTech Connect

    Kumar, Shekhar; Koganti, S.B.

    2008-07-01

    Acetohydroxamic acid (AHA) is a novel complexant for recycle of nuclear-fuel materials. It can be used in ordinary centrifugal extractors, eliminating the need for electro-redox equipment or complex maintenance requirements in a remotely maintained hot cell. In this work, the effect of AHA on Pu(IV) distribution ratios in 30% TBP system was quantified, modeled, and integrated in SIMPSEX code. Two sets of batch experiments involving macro Pu concentrations (conducted at IGCAR) and one high-Pu flowsheet (literature) were simulated for AHA based U-Pu separation. Based on the simulation and validation results, AHA based next-generation reprocessing flowsheets are proposed for co-processing based FBR and thermal-fuel reprocessing as well as evaporator-less macro-level Pu concentration process required for MOX fuel fabrication. Utilization of AHA results in significant simplification in plant design and simpler technology implementations with significant cost savings. (authors)

  9. 65 Years of Reprocessed GLDAS Version 2.0 Data and Their Exploration Using the NASA GES DISC Giovanni

    NASA Technical Reports Server (NTRS)

    Rui, Hualan; Vollmer, B.; Teng, W.; Beaudoing, H.; Rodell, M.; Silberstein, D.

    2015-01-01

    GLDAS-2.0 data have been reprocessed with updated Princeton meteorological forcing data within the Land Information System (LIS) Version 7, and temporal coverage have been extended to 1948-2012.Global Land Data Assimilation System Version 2 (GLDAS-2) has two components: GLDAS-2.0: entirely forced with the Princeton meteorological forcing data GLDAS-2.1: forced with atmospheric analysis and observation-based data after 2001In order to create more climatologically consistent data sets, NASA GSFC's Hydrological Sciences Laboratory (HSL) has recently reprocessed the GLDAS-2.0, by using updated Princeton meteorological forcing data within the LIS Version 7.GLDAS-2.0 data and data services are provided at NASA GES DISC Hydrology Data and Information Services Center (HDISC), in collaboration with HSL.

  10. HUPO Brain Proteome Project: summary of the pilot phase and introduction of a comprehensive data reprocessing strategy.

    PubMed

    Hamacher, Michael; Apweiler, Rolf; Arnold, Georg; Becker, Albert; Blüggel, Martin; Carrette, Odile; Colvis, Christine; Dunn, Michael J; Fröhlich, Thomas; Fountoulakis, Michael; van Hall, André; Herberg, Friedrich; Ji, Jianguo; Ji, Juango; Kretzschmar, Hans; Lewczuk, Piotr; Lubec, Gert; Marcus, Katrin; Martens, Lennart; Palacios Bustamante, Nadine; Park, Young Mok; Pennington, Stephen R; Robben, Johan; Stühler, Kai; Reidegeld, Kai A; Riederer, Peter; Rossier, Jean; Sanchez, Jean-Charles; Schrader, Michael; Stephan, Christian; Tagle, Danilo; Thiele, Herbert; Wang, Jing; Wiltfang, Jens; Yoo, Jong Shin; Zhang, Chenggang; Klose, Joachim; Meyer, Helmut E

    2006-09-01

    The Human Proteome Organisation (HUPO) initiated several projects focusing on the proteome analysis of distinct human organs. The Brain Proteome Project (BPP) is the initiative dedicated to the brain, its development and correlated diseases. Two pilot studies have been performed aiming at the comparison of techniques, laboratories and approaches. With the help of the results gained, objective data submission, storage and reprocessing workflow have been established. The biological relevance of the data will be drawn from the inter-laboratory comparisons as well as from the re-calculation of all data sets submitted by the different groups. In the following, results of the single groups as well as the centralised reprocessing effort will be summarised and compared, showing the added value of this concerted work.

  11. Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

    SciTech Connect

    Worrall, Andrew; Todosow, Michael

    2016-01-01

    Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include: increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance

  12. The Basis for Developing Samarium AMS for Fuel Cycle Analysis

    SciTech Connect

    Buchholz, B A; Biegalski, S R; Whitney, S M; Tumey, S J; Weaver, C J

    2008-10-13

    Modeling of nuclear reactor fuel burnup indicates that the production of samarium isotopes can vary significantly with reactor type and fuel cycle. The isotopic concentrations of {sup 146}Sm, {sup 149}Sm, and {sup 151}Sm are potential signatures of fuel reprocessing, if analytical techniques can overcome the inherent challenges of lanthanide chemistry, isobaric interferences, and mass/charge interferences. We review the current limitations in measurement of the target samarium isotopes and describe potential approaches for developing Sm-AMS. AMS sample form and preparation chemistry will be discussed as well as possible spectrometer operating conditions.

  13. Quantitative NDA measurements of advanced reprocessing product materials containing uranium, neptunium, plutonium, and americium

    NASA Astrophysics Data System (ADS)

    Goddard, Braden

    The ability of inspection agencies and facility operators to measure powders containing several actinides is increasingly necessary as new reprocessing techniques and fuel forms are being developed. These powders are difficult to measure with nondestructive assay (NDA) techniques because neutrons emitted from induced and spontaneous fission of different nuclides are very similar. A neutron multiplicity technique based on first principle methods was developed to measure these powders by exploiting isotope-specific nuclear properties, such as the energy-dependent fission cross sections and the neutron induced fission neutron multiplicity. This technique was tested through extensive simulations using the Monte Carlo N-Particle eXtended (MCNPX) code and by one measurement campaign using the Active Well Coincidence Counter (AWCC) and two measurement campaigns using the Epithermal Neutron Multiplicity Counter (ENMC) with various (alpha,n) sources and actinide materials. Four potential applications of this first principle technique have been identified: (1) quantitative measurement of uranium, neptunium, plutonium, and americium materials; (2) quantitative measurement of mixed oxide (MOX) materials; (3) quantitative measurement of uranium materials; and (4) weapons verification in arms control agreements. This technique still has several challenges which need to be overcome, the largest of these being the challenge of having high-precision active and passive measurements to produce results with acceptably small uncertainties.

  14. An Integrative Model for the Neural Mechanism of Eye Movement Desensitization and Reprocessing (EMDR)

    PubMed Central

    Coubard, Olivier A.

    2016-01-01

    Since the seminal report by Shapiro that bilateral stimulation induces cognitive and emotional changes, 26 years of basic and clinical research have examined the effects of Eye Movement Desensitization and Reprocessing (EMDR) in anxiety disorders, particularly in post-traumatic stress disorder (PTSD). The present article aims at better understanding EMDR neural mechanism. I first review procedural aspects of EMDR protocol and theoretical hypothesis about EMDR effects, and develop the reasons why the scientific community is still divided about EMDR. I then slide from psychology to physiology describing eye movements/emotion interaction from the physiological viewpoint, and introduce theoretical and technical tools used in movement research to re-examine EMDR neural mechanism. Using a recent physiological model for the neuropsychological architecture of motor and cognitive control, the Threshold Interval Modulation with Early Release-Rate of rIse Deviation with Early Release (TIMER-RIDER)—model, I explore how attentional control and bilateral stimulation may participate to EMDR effects. These effects may be obtained by two processes acting in parallel: (i) activity level enhancement of attentional control component; and (ii) bilateral stimulation in any sensorimotor modality, both resulting in lower inhibition enabling dysfunctional information to be processed and anxiety to be reduced. The TIMER-RIDER model offers quantitative predictions about EMDR effects for future research about its underlying physiological mechanisms. PMID:27092064

  15. Reprocessing of Archival Direct Imaging Data of Herbig Ae/Be Stars

    NASA Astrophysics Data System (ADS)

    Safsten, Emily; Stephens, Denise C.

    2017-01-01

    Herbig Ae/Be (HAeBe) stars are intermediate mass (2-10 solar mass) pre-main sequence stars with circumstellar disks. They are the higher mass analogs of the better-known T Tauri stars. Observing planets within these young disks would greatly aid in understanding planet formation processes and timescales, particularly around massive stars. So far, only one planet, HD 100546b, has been confirmed to orbit a HAeBe star. With over 250 HAeBe stars known, and several observed to have disks with structures thought to be related to planet formation, it seems likely that there are as yet undiscovered planetary companions within the circumstellar disks of some of these young stars.Direct detection of a low-luminosity companion near a star requires high contrast imaging, often with the use of a coronagraph, and the subtraction of the central star's point spread function (PSF). Several processing algorithms have been developed in recent years to improve PSF subtraction and enhance the signal-to-noise of sources close to the central star. However, many HAeBe stars were observed via direct imaging before these algorithms came out. We present here current work with the PSF subtraction program PynPoint, which employs a method of principal component analysis, to reprocess archival images of HAeBe stars to increase the likelihood of detecting a planet in their disks.

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

    PubMed

    Dijkstra, Arie; van Asten, Regine

    2014-01-01

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

  17. Multi-Isotope Process (MIP) Monitor: A Near-Real-Time Monitor For Reprocessing Facilities

    SciTech Connect

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

    2008-06-01

    INTRODUCTION The threat of protracted diversion of Pu from commercial reprocessing operations is perhaps the greatest concern to national and international agencies tasked with safeguarding these facilities. While it is generally understood that a method for direct monitoring of process on-line and in near-real time (NRT) would be the best defense against protracted diversion scenarios, an effective method with these qualities has yet to be developed. Here, we attempt to bridge this gap by proposing an on-line NRT process monitoring method that should be sensitive to minor alterations in process conditions and compatible with small, easily deployable, detection systems. This Approach is known as the Multi-Isotope Process (MIP) Monitor and involves the determination and recognition of the contaminant pattern within a process stream for a suite of indicator (radioactive) elements present in the spent fuel as a function of process variables. Utilization of a suite of radio-elements, including ones with multiple oxidation states, decreases the likelihood that attempts to divert Pu by altering the ReDox environment within the process would go undetected. In addition, by identifying gamma-emitting indicator isotopes, this Approach might eliminate the need for bulky neutron detection systems, relying instead on small, portable, high-resolution gamma detectors easily deployable throughout the facility.

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

    SciTech Connect

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

    1986-09-01

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

  19. The used nuclear fuel problem - can reprocessing and consolidated storage be complementary?

    SciTech Connect

    Phillips, C.; Thomas, I.

    2013-07-01

    This paper describes our CISF (Consolidated Interim Storage Facilities) and Reprocessing Facility concepts and show how they can be combined with a geologic repository to provide a comprehensive system for dealing with spent fuels in the USA. The performance of the CISF was logistically analyzed under six operational scenarios. A 3-stage plan has been developed to establish the CISF. Stage 1: the construction at the CISF site of only a rail receipt interface and storage pad large enough for the number of casks that will be received. The construction of the CISF Canister Handling Facility, the Storage Cask Fabrication Facility, the Cask Maintenance Facility and supporting infrastructure are performed during stage 2. The construction and placement into operation of a water-filled pool repackaging facility is completed for Stage 3. By using this staged approach, the capital cost of the CISF is spread over a number of years. It also allows more time for a final decision on the geologic repository to be made. A recycling facility will be built, this facility will used the NUEX recycling process that is based on the aqueous-based PUREX solvent extraction process, using a solvent of tri-N-butyl phosphate in a kerosene diluent. It is capable of processing spent fuels at a rate of 5 MT per day, at burn-ups up to 50 GWD per ton of spent fuels and a minimum of 5 years out-of-reactor cooling.

  20. Iodine-Xenon Dating: Sensitive Chronometer for Reprocessing in the Primitive Solar System

    NASA Technical Reports Server (NTRS)

    Pravdivtseva, O. V.; Hohenberg, C. M.

    1999-01-01

    The I-Xe chronometer is based upon decay of I-129 to Xe-129 in the early Solar System. Recent comparison of I-Xe system in individual mineral separates from twelve different meteorites with independent Pb-Pb data has demonstrated that I-Xe clock is a reliable sensitive chronometer when applied to a single mineral system. Since most iodine hosts are secondary minerals, the I-Xe clock generally records post-formational processing, providing the information on early meteorite evolution. Absolute I-Xe ages can be found by normalization using the measured I-Xe and Pb-Pb ages of Acapulco phosphate (4.557 plus or minus 0.002 Ga). Absolute ages for the I-Xe internal standards Shallow water and Bjurbole, 4.566 plus or minus 0.002 Ga and 4.565 plus or minus 0.003 Ga, respectively, provide absolute I-Xe ages for all other samples. The I-Xe age of bulk meteorite is meaningful and interpretable only when the carrier of primordial iodine is a major mineral phase (e. g., enstatite chondrites). Using the "monomineral" approach, separated phases from the Richardton H5 chondrite provide a case history of post-formational alteration in this object. This work applies the I-Xe chronometer to determine the times of reprocessing of selected minerals in single meteorite types. A preliminary account of this work was recently reported. Additional information is contained in the original extended abstract.