Science.gov

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. International safeguards for a light-water reactor fuels reprocessing plant: containment and surveillance concepts

    SciTech Connect

    Cameron, C.P.; Bleck, M.E.

    1980-12-01

    Concepts for containment/surveillance for reprocessing plants are described, conceptual designs are developed, and their effectiveness is evaluated. A technical approach to design of containment/surveillance systems is presented, and design considerations are discussed. This is the second in a series of reports. The first described the basis for the study of international safeguards for reprocessing plants. In this second report, only containment/surveillance is discussed. The third report will discuss the integration of concepts for containment/surveillance and material accountancy.

  5. Wastes from selected activities in two light-water reactor fuel cycles

    SciTech Connect

    Palmer, C.R.; Hill, O.F.

    1980-07-01

    This report presents projected volumes and radioactivities of wastes from the production of electrical energy using light-water reactors (LWR). The projections are based upon data developed for a recent environmental impact statement in which the transuranic wastes (i.e., those wastes containing certain long-lived alpha emitters at concentrations of at least 370 becquerels, or 10 nCi, per gram of waste) from fuel cycle activities were characterized. In addition, since the WG.7 assumed that all fuel cycle wastes except mill tailings are placed in a mined geologic repository, the nontransuranic wastes from several activities are included in the projections reported. The LWR fuel cycles considered are the LWR, once-through fuel cycle (Strategy 1), in which spent fuel is packaged in metal canisters and then isolated in geologic formations; and the LWR U/Pu recycle fuel cycle (Strategy 2), wherein spent fuel is reprocessed for recovery and recycle of uranium and plutonium in LWRs. The wastes projected for the two LWR fuel cycles are summarized. The reactor operations and decommissioning were found to dominate the rate of waste generation in each cycle. These activities account for at least 85% of the fuel cycle waste volume (not including head-end wastes) when normalized to per unit electrical energy generated. At 10 years out of reactor, however, spent fuel elements in Strategy 1 represent 98% of the fuel cycle activity but only 4% of the volume. Similarly, the packaged high-level waste, fuel hulls and hardware in Strategy 2 concentrate greater than 95% of the activity in 2% of the waste volume.

  6. 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.

  7. 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.

  8. 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

  9. 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

  10. 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.

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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.

  16. 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.

  17. Breeder Reprocessing Engineering Test

    SciTech Connect

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

    1984-01-01

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

  18. 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.

  19. 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

  20. 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.

  1. 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

  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. 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

  4. Nuclear fuel cycle analysis of the SABR fusion-fission hybrid transmutation reactor

    NASA Astrophysics Data System (ADS)

    Sommer, Chris; Stacey, Weston; Petrovic, Bojan

    2009-11-01

    Various fuel cycles have been designed and analyzed for the Subcritical Advanced Burner Reactor (SABR). SABR is a sodium cooled fast reactor fueled with transuranics (TRU) from spent fuel of light water reactors and driven by a tokamak fusion neutron source based on ITER physics and technology. SABR employs a four batch fuel cycle using an out-to-in shuffling pattern, with the fuel being reprocessed at the end of each cycle. The reprocessing method assumes recovery rates of 99.9% of the actinides and 0.1% of the fission products remain in the recycled fuel. The reprocessing fuel cycles were analyzed to find an optimal cycle length in terms of burn up, power distribution, and materials limitations. Fuel cycles are analyzed using CEA's ERANOS2.0 code, with fuel residence times limited by radiation damage at 100, 150 and 200 dpa.

  5. 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.

  6. 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.

  7. 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.

  8. 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%.

  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. Integrated international safeguards concepts for fuel reprocessing

    SciTech Connect

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

    1981-12-01

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

  11. MOPITT V5 reprocessing

    Atmospheric Science Data Center

    2013-08-06

    ... V5 products. The original L1 filenames included the text string 'L1V3.36' whereas the reprocessed L1 files include 'L1V3.37'. The original L2 filenames included the text string 'L2V10.0' whereas the reprocessed L2 files include 'L2V10.1'.   ...

  12. International Atomic Energy Agency (IAEA) Update on Spent Fuel Management Activities with Focus on Reprocessing

    SciTech Connect

    Lovasic, Z.

    2008-07-01

    level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. Innovative reprocessing methods would have to be developed for the treatment of fuel types that may be utilized in the future; these fuels may differ substantially from the UO{sub 2} or MOX ceramics used in current light water reactors. Continued research and development on these methods must continue in view of the expected evolution in fuel and reactor types. The design of advanced reprocessing methods must deal in a comprehensive manner with (1) safety, (2) the control and minimization of plant effluents, (3) minimization of the waste generation, (4) the production of stable and durable waste forms, and (5) economic competitiveness. International collaboration on the development of advanced reprocessing methods, considering the magnitude of the challenges, is essential to facilitate the future deployment of these technologies. Several organizations (like IAEA and OECD) have developed modelling tools for analyzing nuclear fuel cycles. There are a number of challenges to be addressed in development of new reprocessing strategies and technologies. Proliferation risk has to be reduced. One of the major obstacles to overcome is public acceptance of the advanced fuel strategies. (authors)

  13. Flexible Fuel Cycle Initiative for the Harmonized Deployment of Gen-IV Reactors

    NASA Astrophysics Data System (ADS)

    Fukasawa, Tetsuo; Yamashita, Junichi; Hoshino, Kuniyoshi; Fujimura, Koji; Sasahira, Akira

    Generation IV type fast reactors (FR) are expected to be commercially deployed instead of light water reactors (LWR) from around 2050. Replacement of LWR to FR needs flexibility due to uncertain factors such as FR deployment rate which affects the FR fuel (Pu) supply amount from LWR spent fuel reprocessing and the capacity of related facilities. If the FR deployment rate is as currently planned, more Pu must be prepared by expanding LWR reprocessing. If the FR deployment rate decreases, LWR reprocessing must be reduced to avoid excess Pu. To cope with this issue we proposed the innovative system called Flexible Fuel Cycle Initiative (FFCI) that has integral reprocessing for LWR and FR spent fuels. LWR reprocessing in FFCI only carries out about 90% U recovery and residual material with Pu, U (˜5%), minor actinides (MA) and fission products (FP) goes to FR reprocessing for the planned FR deployment rate. For any decrease in the FR deployment rate temporary storage will be used. Coexistence of Pu/U with MA and FP until just before Pu/U usage in the FR provides high proliferation resistance. Preliminary evaluation revealed that FFCI can reduce the LWR reprocessing capacity and LWR spent fuel storage amount compared with current plan (reference system) if the FR deployment rate decreases. Several FR deployment scenarios and countermeasures such as FFCI were investigated.

  14. 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.

  15. 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.

  16. Multi-Application Small Light Water Reactor

    SciTech Connect

    Pierre Babka

    2002-10-31

    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 objective was 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.

  17. 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.

  18. 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.

  19. 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)

  20. 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.

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

    SciTech Connect

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

    2010-01-01

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

  2. Reprocessing in Luminous Disks

    NASA Astrophysics Data System (ADS)

    Bell, K. R.

    1999-11-01

    We develop and investigate a procedure that accounts for disk reprocessing of photons that originate in the disk itself. Surface temperatures and simple, blackbody spectral energy distributions (SEDs) of protostellar disks are calculated. In disks that flare with radius, reprocessing of stellar photons results in temperature profiles that are not power-law at all radii but are consistently shallower than r-3/4. Including the disk as a radiation source (as in the case of active accretion) 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 that 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 have 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 inner disk. Reprocessing throughout the disk depends sensitively on the inner disk shape and emitted temperature profile. We show that the thermal instability outburst models of Bell & Lin reproduce trends in the observed SEDs of FU Ori systems with T~r-3/4 in the inner disk (r<~0.25 AU corresponding to λ<~10 μm) and T~r-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 flux, outer disk (Ṁ=10

  3. An 'In Situ' Calibration-Correction Procedure (KCICLO) Based on AOD Diurnal Cycle: Comparative Results Between AERONET and Reprocessed (KCICLO method) AOD-Alpha Data Series at El Arenosillo, Spain

    SciTech Connect

    Cachorro, V. E.; Toledano, C.; Sorribas, M.; Berjon, A.; de Frutos, A. M.; Laulainen, Nels S.

    2008-01-31

    A comparative evaluation is carried out for nearly 5 years (February 2000 to May 2004) of data of aerosol optical depth (AOD) measured at the Aerosol Robotic Network (AERONET) site El Arenosillo (Huelva, southwestern Spain). The AERONET database and the reprocessed data set using a new correction procedure, which we call the KCICLO method, are compared with respect to the aerosol local climatology. The cause and necessity of AOD reprocessing were due to the existence of an observed fictitious diurnal cycle (including negative values) because of a deficient calibration as explained in detail in the companion paper (V. E. Cachorro et al., submitted manuscript, 2007). The derived alpha Ångström coefficient is also compared, as it appears to be an excellent indicator of the AOD data quality, because of its sensitivity to AOD variations and errors. Some illustrative cases show the influence of this fictitious diurnal cycle on the shape and values of diurnal variations of the AOD (or alpha), reaching differences as high as 100%, and the improvement resulting from using the KCICLO method. Absolute and relative differences are evaluated from the overall average of AOD and alpha coefficient of AERONET and KCICLO data series, making an exhaustive analysis for each spectral channel and for every photometer separately. Although great variability is shown for each filter and each photometer, apart from photometer 114 data that did not reach level 2.0, the discrepancy in the AOD local climatology in the four filters varies as a whole from 2.3% to 8.5% (2.4% for alpha coefficient). These values show a considerable reduction because of the compensating effect between the different photometers (positive or negative bias), and several jumps that break the continuity of the data series are observed. When monthly and yearly averages are analyzed, the differences are considerably reduced in such a way that the local climatology is not substantially affected, but we must be cautious

  4. Effect of repeated tracheostomy tube reprocessing on biofilm formation

    PubMed Central

    Rodney, Jennifer; Ojano‐Dirain, Carolyn P.; Antonelli, Patrick J.

    2015-01-01

    Objectives/Hypothesis To determine the effect of repeated reprocessing of pediatric tracheostomy tubes (TTs) on biofilm formation. Study Design In vitro microbiological study. Methods Pediatric, uncuffed, polyvinyl chloride (PVC) TTs from two different manufacturers (Tracoe Mini and Shiley) were reprocessed mechanically with household detergent and soaked in sodium hypochlorite (bleach). Two TTs of each brand were reprocessed 0 (control), 10, or 20 times. Twenty 2‐mm coupons were then obtained from each TT, immersed in human mucus, and cultured with either Staphylococcus aureus or Pseudomonas aeruginosa. Biofilm formation was evaluated with bacterial counts. Results Bacterial counts of S. aureus for both brands were significantly higher on the TTs that were reprocessed 20 times compared to those that were not reprocessed (Tracoe: P = .040, Shiley: P  <  .0001) or those that were reprocessed 10 times (Tracoe: P = .022, Shiley: P = .0002). There was no difference between controls and TTs reprocessed 10 times (Tracoe: P = .76, Shiley: P = .24). P. aeuruginosa counts were not significantly different among the varying numbers of reprocessing cycles for either Tracoe or Shiley TTs (P = .08 and P = .97, respectively). Conclusions Repeated reprocessing of PVC TTs with detergent and bleach paradoxically promotes S. aureus biofilm development, possibly due to degradation of the tube surface that facilitates bacterial attachment. Further investigation is needed to determine the optimal technique and limits of reprocessing TTs in clinical practice. Level of Evidence NA Laryngoscope, 126:996–999, 2016 PMID:26267243

  5. Development of reprocessing technology for breeder recycle. [Consolidated fuel reprocessing

    SciTech Connect

    Groenier, W.S.; Crouse, D.J.; Vondra, B.L.

    1980-01-01

    This paper is an overview of some of the activities under the Consolidated Fuel Reprocessing Program. Current status of technology for portions of four reprocessing development tasks is summarized: voloxidation, dissolution, solvent extraction, and off-gas processing. The Hot Experimental Facility is alos mentioned. 6 figures. (DLC)

  6. 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.

  7. 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.

  8. Surface evaluation of cardiac angiographic catheters after simulated use and reprocessing

    NASA Astrophysics Data System (ADS)

    Lucas, Thabata Coaglio; Oréfice, Rodrigo Lambert; Pinotti, Marcos; Huebner, Rudolf

    2009-12-01

    Reprocessing of single-use intravascular catheters is a common practice in public health services and hospitals. The determination of safe number of reprocessing cycles before the catheter integrity becomes compromised has been a priority issue. The present paper addresses the evaluating molecular and micro-structural integrity of reprocessed cardiac angiographic catheters. The Fourier Transform Infrared Spectroscopy and Scanning Electron Microscopy were carried out to elucidate morphological changes. The tensile test was performed on catheters to examine changes in bulk characteristics. In this work, samples of catheters were reprocessed until nine times and sterilized by hydrogen peroxide plasma. It was observed that the number of hydrogen-bonded carbonyls groups increased in 0.05 u.a. ( p < 0.001) after each reprocessing cycle. The spectra indicated degradation products included acids, esters, alcohols, and small amounts of other products containing a carbonyl functional group. The micrographs revealed that only after the fourth reprocessing cycle the effect increased in the surface roughness was more pronounced. On the other hand, after each reprocessing cycle and as consequence of extensive aging of polyamide/polyurethane blends of the catheters surface, it was observed that the micro-fissures, micro-scratches and micro-pores increased in quantity and length. The mechanical test proved that the Young modulus increased in average 3.26 MPa ( p = 0.0003) at increasing number of reprocessing cycles, also suggestive of crosslinking in this material.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

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

    SciTech Connect

    Charlton, William S

    1999-09-01

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

  16. Current Practice of Duodenoscope Reprocessing.

    PubMed

    Kim, Stephen; Muthusamy, V Raman

    2016-10-01

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

  17. 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.

  18. Retrospective CMORPH Reprocessing Efforts

    NASA Astrophysics Data System (ADS)

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

    2008-05-01

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

  19. 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.

  20. 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.

  1. Issues concerned with future light-water-reactor designs

    SciTech Connect

    Tong, L.S.

    1982-03-01

    This article discusses some light-water-reactor (LWR) design issues that are based on operating experiences and the results of water-reactor safety research. The impacts of these issues on reactor safety are described, and new engineering concepts are identified to encourage further improvement in future light-water-reactor designs.

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. Accident Performance of Light Water Reactor Cladding Materials

    SciTech Connect

    Nelson, Andrew T.

    2012-07-24

    During a loss of coolant accident as experienced at Fukushima, inadequate cooling of the reactor core forces component temperatures ever higher where they must withstand aggressive chemical environments. Conventional zirconium cladding alloys will readily oxidize in the presence of water vapor at elevated temperatures, rapidly degrading and likely failing. A cladding breach removes the critical barrier between actinides and fission products and the coolant, greatly increasing the probability of the release of radioactivity in the event of a containment failure. These factors have driven renewed international interest in both study and improvement of the materials used in commercial light water reactors. Characterization of a candidate cladding alloy or oxidation mitigation technique requires understanding of both the oxidation kinetics and hydrogen production as a function of temperature and atmosphere conditions. Researchers in the MST division supported by the DOE-NE Fuel Cycle Research and Development program are working to evaluate and quantify these parameters across a wide range of proposed cladding materials. The primary instrument employed is a simultaneous thermal analyzer (STA) equipped with a specialized water vapor furnace capable of maintaining temperatures above 1200 C in a range of atmospheres and water vapor contents. The STA utilizes thermogravimetric analysis and a coupled mass spectrometer to measure in situ oxidation and hydrogen production of candidate materials. This capability is unprecedented in study of materials under consideration for reactor cladding use, and is currently being expanded to investigate proposed coating techniques as well as the effect of coating defects on corrosion resistance.

  7. NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL FROM PHWR'S IN A CLOSED THORIUM FUEL CYCLE

    SciTech Connect

    Sleaford, B W; Collins, B A; Ebbinghaus, B B; Bathke, C G; Prichard, A W; Wallace, R K; Smith, B W; Hase, K R; Bradley, K S; Robel, M; Jarvinen, G D; Ireland, J R; Johnson, M W

    2010-04-26

    This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that {sup 233}U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined to date need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of 'attractiveness levels' that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented.

  8. 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.

  9. Data validation and security for reprocessing.

    SciTech Connect

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

    2008-10-01

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

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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

  18. 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

  19. 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

  20. 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.

  1. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    SciTech Connect

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers 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 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  2. 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.

  3. 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.

  4. Risk management and decision rules for light water reactor

    SciTech Connect

    Griesmeyer, J. M.; Okrent, D.

    1981-01-01

    The process of developing and adopting safety objectives in quantitative terms can provide a basis for focusing societal decision making on the suitability of such objectives and upon questions of compliance with those objectives. A preliminary proposal for a light water reactor (LWR) risk management framework is presented as part of that process.

  5. 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.

  6. 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

  7. Analysis of long time series of reprocessed GPS total column water vapour estimates

    NASA Astrophysics Data System (ADS)

    Bock, O.; Garayt, B.; Bar-Sever, Y.; Byun, S.

    2012-04-01

    Reprocessed GPS data provide accurate and stable estimates of zenith tropospheric delay (ZTD) and total column water vapour (TCWV) estimates. Time series exceeding 15 years become progressively available over the globally distributed continuously-operating International GNSS Service (IGS) network and the European EUREF Permanent Network (EPN). This work aims at assessing the quality of such reprocessed ZTD solutions and using them for climate monitoring and model validation. First we assessed the quality of three ZTD solutions: (i) the reprocessed tropospheric solution produced at JPL for IGS (repro1, covering period 1995-2007), (ii) the operational IGS tropospheric solution (trop_new, covering period 2001-2010), and (iii) a reprocessed solution produced at IGN (sgn_repro1, covering period 2004-2010). All three solutions show a good overall agreement. Slight differences are due to use of different data processing procedures (e.g. antenna model, mapping function). In several cases, doubtful metadata (e.g. logfile not updated) seems responsible of discrepancies in the operational solution which were corrected during reprocessing. The reprocessed GPS ZTD estimates were converted into TCWV and analysed globally and for different regions, with a focus on timescales pertinent to climate (seasonal cycle, diurnal cycle, etc.). The GPS TCWV estimates were also compared to the ECMWF reanalysis ERA-Interim and overall good agreement is found.

  8. X-ray reprocessing in binaries

    NASA Astrophysics Data System (ADS)

    Paul, Biswajit

    2016-07-01

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

  9. Third-party reprocessing of endoscopic accessories.

    PubMed

    Furman, P J

    2000-04-01

    Third-party reprocessing of medical devices labeled for single use is a safe, FDA regulated practice that helps hospitals reduce costs without compromising patient care. Simply because a device is labeled as single use does not mean it cannot be safely reprocessed. To the contrary, the single use label is chosen by the manufacturer, sometimes for economic gain, as there are no formal FDA regulations or standards to distinguish between reusable and single use devices. The current FDA regulatory framework for third-party reprocessors, which emphasizes compliance with FDA quality assurance requirements, is presently under review, and the agency is in the process of developing a new regulatory scheme for reprocessing.

  10. Remote maintenance in nuclear fuel reprocessing

    SciTech Connect

    Herndon, J.N.

    1985-01-01

    Remote maintenance techniques applied in large-scale nuclear fuel reprocessing plants are reviewed with particular attention to the three major maintenance philosophy groupings: contact, remote crane canyon, and remote/contact. Examples are given, and the relative success of each type is discussed. Probable future directions for large-scale reprocessing plant maintenance are described along with advanced manipulation systems for application in the plants. The remote maintenance development program within the Consolidated Fuel Reprocessing Program at the Oak Ridge National Laboratory is also described. 19 refs., 19 figs.

  11. 10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... property. 2. A fuel reprocessing plant's inventory of high-level liquid radioactive wastes will be limited... liquid radioactive wastes” means those aqueous wastes resulting from the operation of the first cycle..., or equivalent, in a facility for reprocessing irradiated reactor fuels.) High-level...

  12. 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.

  13. Issues affecting advanced passive light-water reactor safety analysis

    SciTech Connect

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-08-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

  14. Issues affecting advanced passive light-water reactor safety analysis

    SciTech Connect

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-01-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

  15. CASL: The Consortium for Advanced Simulation of Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Kothe, Douglas B.

    2010-11-01

    Like the fusion community, the nuclear engineering community is embarking on a new computational effort to create integrated, multiphysics simulations. The Consortium for Advanced Simulation of Light Water Reactors (CASL), one of 3 newly-funded DOE Energy Innovation Hubs, brings together an exceptionally capable team 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 the Virtual Reactor (VR), will: 1) Enable the use of leadership-class computing for engineering design and analysis to improve reactor capabilities, 2) Promote an enhanced scientific basis and understanding by replacing empirically based design and analysis tools with predictive capabilities, 3) Develop a highly integrated multiphysics environment for engineering analysis through increased fidelity methods, and 4) Incorporate UQ as a basis for developing priorities and supporting, application of the VR tools for predictive simulation. In this presentation, we present the plans for CASL and comment on the similarity and differences with the proposed Fusion Simulation Project (FSP).

  16. 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.

  17. 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.

  18. Core design study of a supercritical light water reactor with double row fuel rods

    SciTech Connect

    Zhao, C.; Wu, H.; Cao, L.; Zheng, Y.; Yang, J.; Zhang, Y.

    2012-07-01

    An equilibrium core for supercritical light water reactor has been designed. A novel type of fuel assembly with dual rows of fuel rods between water rods is chosen and optimized to get more uniform assembly power distributions. Stainless steel is used for fuel rod cladding and structural material. Honeycomb structure filled with thermal isolation is introduced to reduce the usage of stainless steel and to keep moderator temperature below the pseudo critical temperature. Water flow scheme with ascending coolant flow in inner regions is carried out to achieve high outlet temperature. In order to enhance coolant outlet temperature, the radial power distributions needs to be as flat as possible through operation cycle. Fuel loading pattern and control rod pattern are optimized to flatten power distribution at inner regions. Axial fuel enrichment is divided into three parts to control axial power peak, which affects maximum cladding surface temperature. (authors)

  19. 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.

  20. Plasma techniques for reprocessing nuclear wastes

    SciTech Connect

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

    1993-06-01

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

  1. 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.

  2. Advanced light water reactor requirements document: Chapter 4, Reactor systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the following for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR): reactor pressure vessel, nozzles and safe-ends, reactor internals, in-vessel portions of fluid systems (including reactor internal pumps (Emergency Core Cooling System (ECCS) piping and spargers), nuclear fuel, and the control rods and control rod drive system (including hydraulic supply and accumulators). Special tools required for reactor system maintenance, inspection and testing are also covered.

  3. 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)

  4. 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.

  5. 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.

  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. 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.

  8. 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.

  9. Optimizing near real time accountability for reprocessing.

    SciTech Connect

    Cipiti, Benjamin B.

    2010-06-01

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

  10. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

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

  11. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

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

  12. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

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

  13. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

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

  14. 21 CFR 211.115 - Reprocessing.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

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

  15. 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

  16. 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

  17. 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

  18. Development of fluoride reprocessing technologies devoted to molten-salt reactor systems

    SciTech Connect

    Uhlir, Jan; Marecek, Martin; Tulackova, Radka; Chuchvalcova Bimova, Karolina

    2007-07-01

    Main fuel processing and reprocessing technologies proposed for Molten Salt Reactor fuel cycle are pyrochemical or pyrometallurgical, majority of them are fluoride technologies. It is based on the fact that Molten Salt Reactor fuel is in the chemical form of molten fluorides and the reprocessing technology is needed to be an 'on-line' process. The corresponding pyrochemical separation processes proposed for MSR fuel processing and reprocessing are mainly fluoride volatilization processes, molten salt / liquid metal extraction processes, electrochemical separation processes from the molten salt media and gas extraction from the molten salt medium. Techniques based on fluoride volatilization and on electrochemical separation from fluoride molten salt media are under development in the Czech Republic. Whereas the Fluoride Volatility Method is proposed to be the main 'Front-end' technology of the MSR used as the actinide burner (transmuter), the electro-separation methods should be dedicated to the 'on-line' reprocessing of the circulating MSR fuel and should be used as for MSR incinerating transuranium fuel as for MSR working within the {sup 232}Th - {sup 233}U fuel cycle. (authors)

  19. Silicon carbide composite for light water reactor fuel assembly applications

    NASA Astrophysics Data System (ADS)

    Yueh, Ken; Terrani, Kurt A.

    2014-05-01

    The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  20. 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.

  1. 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.

  2. 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.

  3. Seismic margin assessment of evolutionary light water reactors

    SciTech Connect

    Ali, S.A.; Bagchi, G.

    1996-12-01

    The objectives of the US Nuclear Regulatory Commission (NRC) staff`s review of the evolutionary light water reactors (ELWR) probabilistic risk assessment (PRA) are drawn from 10 CFR Part 52, the Commission`s Severe Reactor Accident Policy Statement regarding future designs and existing plants, the Commission`s Safety Goal Policy Statement, The Commission approved positions concerning the analyses of external and events contained in SECY-93-087, and NRC interest in the use of PRA to help improve future reactor designs. In general, these objectives have been achieved by the ELWR PRAs and the NRC staff`s review. The staff`s applicable regulation for the analysis of external events for the ELWR PRAs is as follows. The probabilistic risk assessment required by 10 CFR 52.47(a)(1)(v) must include an assessment of internal and external events. For external events, simplified probabilistic methods and margins methods may be used to assess the capacity of the standard design to withstand the effects of events such as fires and earthquakes. Traditional probabilistic techniques should be used to evaluate internal floods. For earthquakes, a seismic margin analysis must consider the effects of earthquakes with accelerations approximately one and two-thirds the acceleration of the safe-shutdown earthquake (SSE).

  4. 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

  5. Distribution of characteristics of LWR [light water reactor] spent fuel

    SciTech Connect

    Reich, W.J.; Notz, K.J.; Moore, R.S.

    1991-01-01

    The purpose of this report is to develop a collective description of the entire spent fuel inventory in terms of various fuel properties relevant to Approved Testing Materials (ATMs) using information available from the Characteristics Data Base (CBD), which is sponsored by the US Department of Energy`s (DOE`s) Office of Civilian Radioactive Waste Management. A number of light-water reactor (LWR) characteristics were analyzed including assembly class representation, fuel burnup, enrichment, fuel fabrication data, defective fuel quantities, and, at PNL`s specific request, linear heat generation rate (LHGR) and the utilization of burnable poisons. A quantitative relationships was developed between burnup and enrichment for BWRs and PWRs. The relationship shows that the existing BWR ATM is near the center of the burnup-enrichment distribution, while the four PWR ATMs bracket the center of the burnup range but are on the low side of the enrichment range. Fuel fabrication data are based on vendor specifications for new fuel. Defective fuel distributions were analyzed in terms of assembly class and vendor design. LHGR values were calculated from utility data on burnup and effective full-power days; these calculations incorporate some unavoidable assumptions which may compromise the value of the results. Only a limited amount of data are available on burnable poisons at this time. Based on this distribution study, suggestions for additional ATMs are made. These are based on the class and design concepts and include BWR/2,3 barrier fuel, and the WE 17 {times} 17 class with integral burnable poison. Both should be at relatively high burnups. 16 refs., 5 figs., 15 tabs.

  6. 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.

  7. 77 FR 15812 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-16

    ... COMMISSION Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors AGENCY: Nuclear...-1265, ``Initial Test Program of Condensate and Feedwater Systems for Light- Water Reactors.'' DG-1265... plant startup, and power ascension tests for the condensate and feedwater systems in all types of...

  8. 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.

  9. 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%.

  10. 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.

  11. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production Progress Report for Year 1, Quarter 2 (January - March 2002)

    SciTech Connect

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-03-01

    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.

  12. R and D of On-line Reprocessing Technology for Molten-Salt Reactor Systems

    SciTech Connect

    Uhlir, Jan; Tulackova, Radka; Chuchvalcova Bimova, Karolina

    2006-07-01

    The Molten Salt Reactor (MSR) represents one of promising future nuclear reactor concept included in the Generation IV reactors family. The reactor can be operated as the thorium breeder or as the actinide transmuter. However, the future deployment of Molten-Salt Reactors will be significantly dependent on the successful mastering of advanced reprocessing technologies dedicated to their fuel cycle. Here the on-line reprocessing technology connected with the fuel circuit of MSR is of special importance because the reactor cannot be operated for a long run without the fuel salt clean-up. Generally, main MSR reprocessing technologies are pyrochemical, majority of them are fluoride technologies. The proposed flow-sheets of MSR on-line reprocessing are based on a combination of molten-salt / liquid metal extraction and electro-separation processes, which can be added to the gas extraction process already verified during the MSRE project in ORNL. The crucial separation method proposed for partitioning of actinides from fission products is based on successive Anodic dissolution and Cathodic deposition processes in molten fluoride media. (authors)

  13. 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

  14. 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

  15. 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.

  16. Assessment of the use of extended burnup fuel in light water power reactors

    SciTech Connect

    Baker, D.A.; Bailey, W.J.; Beyer, C.E.; Bold, F.C.; Tawil, J.J.

    1988-02-01

    This study has been conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water power reactors. It has been proposed that current batch average burnup levels of 33 GWd/t uranium be increased to above 50 GWd/t. The environmental effects of extending fuel burnup during normal operations and during accident events and the economic effects of cost changes on the fuel cycle are discussed in this report. The physical effects of extended burnup on the fuel and the fuel assembly are also presented as a basis for the environmental and economic assessments. Environmentally, this burnup increase would have no significant impact over that of normal burnup. Economically, the increased burnup would have favorable effects, consisting primarily of a reduction: (1) total fuel requirements; (2) reactor downtime for fuel replacement; (3) the number of fuel shipments to and from reactor sites; and (4) repository storage requirements. 61 refs., 4 figs., 27 tabs.

  17. 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.

  18. 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.

  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. Current status and directions for fast reactor reprocessing

    SciTech Connect

    Burch, W.D.

    1983-01-01

    The development of fast breeder reactors (FBRs) for commercial electric power production has been under way in several countries for more than 20 years. In the United States as elsewhere, early work was centered on small reactors to prove the feasibility of concepts and later was followed by larger reactors to test engineering features and to develop fuel technology. In the early 1970s, with the perceived crisis in electrical generation expected late in this century, major efforts were mounted to plan and carry out comprehensive development programs to ensure the capability to develop and begin using this new form of nuclear power by the end of this century. This comprehensive effort included the first serious efforts directed toward the supporting fuel cycle activities. However, because of the effects of the oil price rise and resulting conservation, a slowdown of industrial growth, and cut-backs in energy needs, there has been a decline in program activities. Unlike the fuel cycle for light-water reactors (LWRs), where supply and the back-end recycle and/or waste disposal activities can largely be uncoupled, recovery and recycle of fissile materials in spent fuel must be accomplished in one or two years in a practical breeder system. 3 references.

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

    SciTech Connect

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

    2002-01-01

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

  2. 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)

  3. Simplified probabilistic risk assessment in fuel reprocessing

    SciTech Connect

    Solbrig, C.W.

    1993-03-01

    An evaluation was made to determine if a backup mass tracking computer would significantly reduce the probability of criticality in the fuel reprocessing of the Integral Fast Reactor. Often tradeoff studies, such as this, must be made that would greatly benefit from a Probably Risk Assessment (PRA). The major benefits of a complete PRA can often be accrued with a Simplified Probabilistic Risk Assessment (SPRA). An SPRA was performed by selecting a representative fuel reprocessing operation (moving a piece of fuel) for analysis. It showed that the benefit of adding parallel computers was small compared to the benefit which could be obtained by adding parallelism to two computer input steps and two of the weighing operations. The probability of an incorrect material moves with the basic process is estimated to be 4 out of 100 moves. The actual values of the probability numbers are considered accurate to within an order of magnitude. The most useful result of developing the fault trees accrue from the ability to determine where significant improvements in the process can be made. By including the above mentioned parallelism, the error move rate can be reduced to 1 out of 1000.

  4. Simplified probabilistic risk assessment in fuel reprocessing

    SciTech Connect

    Solbrig, C.W.

    1993-01-01

    An evaluation was made to determine if a backup mass tracking computer would significantly reduce the probability of criticality in the fuel reprocessing of the Integral Fast Reactor. Often tradeoff studies, such as this, must be made that would greatly benefit from a Probably Risk Assessment (PRA). The major benefits of a complete PRA can often be accrued with a Simplified Probabilistic Risk Assessment (SPRA). An SPRA was performed by selecting a representative fuel reprocessing operation (moving a piece of fuel) for analysis. It showed that the benefit of adding parallel computers was small compared to the benefit which could be obtained by adding parallelism to two computer input steps and two of the weighing operations. The probability of an incorrect material moves with the basic process is estimated to be 4 out of 100 moves. The actual values of the probability numbers are considered accurate to within an order of magnitude. The most useful result of developing the fault trees accrue from the ability to determine where significant improvements in the process can be made. By including the above mentioned parallelism, the error move rate can be reduced to 1 out of 1000.

  5. 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)

  6. 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

    ... COMMISSION Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY... Treatment of Non-Safety 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:...

  7. 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...

  8. 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... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  9. 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... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  10. 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... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  11. 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... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  12. 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... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  13. Comparative study of plutonium burning in heavy and light water reactors.

    SciTech Connect

    Taiwo, T. A.; Kim, T. K.; Szakaly, F. J.; Hill, R. N.; Yang, W. S.; Dyck, G. R.; Hyland, B.; Edwards, G. W. R.; Nuclear Engineering Division; Atomic Energy Canada Ltd.

    2008-01-01

    There is interest in the U.S. and world-wide in reducing the burden on geological nuclear fuel disposal sites. In some disposal scenarios, the decay heat loading of the surrounding rock limits the commercial spent fuel capacity of the sites. In the long term (100 to 1,500 years), this decay heat is generated primarily by actinides, particularly {sup 241}Am and {sup 241}Pu. One possible approach to reducing this decay-heat burden would be to reprocess commercial spent nuclear fuel and use intermediate-tier thermal reactors to 'burn' these actinides and other transuranics (plutonium and higher actinides). The viability of this approach is dependent on the detailed changes in chemical and isotopic compositions of actinide-bearing fuels after irradiation in thermal reactor spectra. The intermediate-tier thermal burners could bridge the commercial water-cooled reactors and fast reactors required for ultimate consumption of the transuranics generated in the commercial reactors. This would reduce the number of such fast reactors required to complete the mission of burning transuranics. If thermal systems are to be used for the transmutation mission, it is likely that they would be similar to or are advanced versions of the systems currently used for power generation. In both the U.S. and Canada, light- and heavy-water-cooled thermal reactors are used for power generation in the commercial nuclear sector. About 103 pressurized- and boiling- light water reactors (PWRs and BRWs) are deployed in the U.S. nuclear industry while about 18 CANDU (heavy-water-cooled) reactors are used in the Canadian industry. There are substantial differences between light and heavy water-cooled reactors that might affect transmutation potential. These arise from differences in neutron balance of the reactors, in neutron energy spectra, in operational approaches (e.g., continuous refueling enhancing fuel burnup), and so on. A systematic study has been conducted to compare the transmutation

  14. 9 CFR 114.18 - Reprocessing of biological products.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ..., DEPARTMENT OF AGRICULTURE VIRUSES, SERUMS, TOXINS, AND ANALOGOUS PRODUCTS; ORGANISMS AND VECTORS PRODUCTION... authorize a licensee to reprocess a serial of completed product subject to the conditions prescribed in this... conducted on the reprocessed product. A serial found unsatisfactory by a required test shall not be...

  15. 9 CFR 114.18 - Reprocessing of biological products.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ..., DEPARTMENT OF AGRICULTURE VIRUSES, SERUMS, TOXINS, AND ANALOGOUS PRODUCTS; ORGANISMS AND VECTORS PRODUCTION... authorize a licensee to reprocess a serial of completed product subject to the conditions prescribed in this... conducted on the reprocessed product. A serial found unsatisfactory by a required test shall not be...

  16. 9 CFR 114.18 - Reprocessing of biological products.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 9 Animals and Animal Products 1 2013-01-01 2013-01-01 false Reprocessing of biological products. 114.18 Section 114.18 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION SERVICE... REQUIREMENTS FOR BIOLOGICAL PRODUCTS § 114.18 Reprocessing of biological products. The Administrator...

  17. 9 CFR 114.18 - Reprocessing of biological products.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 9 Animals and Animal Products 1 2012-01-01 2012-01-01 false Reprocessing of biological products. 114.18 Section 114.18 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION SERVICE... REQUIREMENTS FOR BIOLOGICAL PRODUCTS § 114.18 Reprocessing of biological products. The Administrator...

  18. Safeguards instruments for Large-Scale Reprocessing Plants

    SciTech Connect

    Hakkila, E.A.; Case, R.S.; Sonnier, C.

    1993-06-01

    Between 1987 and 1992 a multi-national forum known as LASCAR (Large Scale Reprocessing Plant Safeguards) met to assist the IAEA in development of effective and efficient safeguards for large-scale reprocessing plants. The US provided considerable input for safeguards approaches and instrumentation. This paper reviews and updates instrumentation of importance in measuring plutonium and uranium in these facilities.

  19. Data reprocessing on worldwide distributed systems

    SciTech Connect

    Diesburg, Mike; Wicke, Daniel; /Fermilab

    2004-12-01

    The D0 experiment faces many challenges enabling access to large datasets for physicists on 4 continents. The strategy followed is to make use of worldwide distributed computing clusters. Since the begin of Tevatron Run-II (March 2001) all Monte-Carlo simulations are produced outside of Fermilab on remote systems. For analyses a system of regional analysis centres (RACs) was established which supply the associated institutes with the data. This structure which is similar to the Tier structure foreseen for LHC was used in autumn 2003 to reprocess all D0-data with the up-to-date and much improved reconstruction software. With these achievements D0 is the first high energy physics experiment to have implemented and operated all essential computing tasks on non-dedicated, worldwide distributed systems. The experiences gained in D0 can be applied to evaluate the LHC computing model.

  20. Data reprocessing on worldwide distributed systems

    SciTech Connect

    Wicke, Daniel; /Fermilab

    2004-12-01

    The D0 experiment faces many challenges in terms of enabling access to large datasets for physicists on four continents. The strategy for solving these problems on worldwide distributed computing clusters is presented. Since the beginning of Run II of the Tevatron (March 2001) all Monte-Carlo simulations for the experiment have been produced at remote systems. For data analysis, a system of regional analysis centers (RACs) was established which supply the associated institutes with the data. This structure, which is similar to the tiered structure foreseen for the LHC was used in Fall 2003 to reprocess all D0 data with a much improved version of the reconstruction software. This makes D0 the first running experiment that has implemented and operated all important computing tasks of a high energy physics experiment on systems distributed worldwide.

  1. 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.

  2. 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) PMID:2896720

  3. Towards the reanalysis of void coefficients measurements at proteus for high conversion light water reactor lattices

    SciTech Connect

    Hursin, M.; Koeberl, O.; Perret, G.

    2012-07-01

    High Conversion Light Water Reactors (HCLWR) allows a better usage of fuel resources thanks to a higher breeding ratio than standard LWR. Their uses together with the current fleet of LWR constitute a fuel cycle thoroughly studied in Japan and the US today. However, one of the issues related to HCLWR is their void reactivity coefficient (VRC), which can be positive. Accurate predictions of void reactivity coefficient in HCLWR conditions and their comparisons with representative experiments are therefore required. In this paper an inter comparison of modern codes and cross-section libraries is performed for a former Benchmark on Void Reactivity Effect in PWRs conducted by the OECD/NEA. It shows an overview of the k-inf values and their associated VRC obtained for infinite lattice calculations with UO{sub 2} and highly enriched MOX fuel cells. The codes MCNPX2.5, TRIPOLI4.4 and CASMO-5 in conjunction with the libraries ENDF/B-VI.8, -VII.0, JEF-2.2 and JEFF-3.1 are used. A non-negligible spread of results for voided conditions is found for the high content MOX fuel. The spread of eigenvalues for the moderated and voided UO{sub 2} fuel are about 200 pcm and 700 pcm, respectively. The standard deviation for the VRCs for the UO{sub 2} fuel is about 0.7% while the one for the MOX fuel is about 13%. This work shows that an appropriate treatment of the unresolved resonance energy range is an important issue for the accurate determination of the void reactivity effect for HCLWR. A comparison to experimental results is needed to resolve the presented discrepancies. (authors)

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

    SciTech Connect

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

    1986-06-01

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

  5. 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.

  6. 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.

  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. Analysis of the light-water flooding of the HFBR thimble tubes

    SciTech Connect

    Carew, J.F.; Aronson, A.L.; Cokinos, D.M.; Prince, A.; Todosow, M.; Tichler, P.R.; Cheng, L.Y.; Karol, R.C. )

    1991-01-01

    The fuel elements surrounding the central vertical thimble tubes in the Brookhaven National Laboratory High-Flux Beam Reactor (HFBR) are highly undermoderated, and light-water flooding of these irradiation thimbles results in a positive core reactivity insertion. The light-water contamination of the D{sub 2}O thimble tube coolant is the result of a postulated double-ended guillotine break of a U tube in the experimental facilities heat exchanger during the HFBR light-water flooding (LWF) event. While this event has a low probability (1.3 x 10{sup {minus}4}/yr), the HFBR protection system must ensure adequate thermal margin during the power transient. This paper summarizes the analysis of the HFBR thimble-tube LWF event.

  9. 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.

  10. 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.

  11. Residual protein levels on reprocessed dental instruments.

    PubMed

    Smith, A; Letters, S; Lange, A; Perrett, D; McHugh, S; Bagg, J

    2005-11-01

    Reduction of the initial bioburden on instruments, prior to sterilization, is believed to reduce transmission risks of iatrogenic Creutzfeldt-Jakob disease. Endodontic files are used in the preparation of root canals and are likely to have close contact and become contaminated with neural material from branches of the maxillary and mandibular cranial nerves. This study examined methods used by 22 dental practices to clean endodontic files, and scored visible debris and residual protein levels adhering to 220 dental endodontic files that had been used, cleaned, autoclaved and were deemed ready for re-use. Visible debris was scored after examination under a dissecting light microscope. Residual protein was quantified using a fluorescent assay based on reaction of proteins with o-phthaldialdehyde/N-acetyl cysteine. There was wide variation in the methods used by practices to clean endodontic files. The cleaning process varied from a wipe with an alcohol-impregnated cloth to hand scrubbing and/or use of an ultrasonic bath. Surface debris was visually detected on 98% of files. Residual protein was detected on all the files examined (median amount: 5.4 microg; range: 0.5-63.2 microg). These results demonstrate that the cleaning of some instruments reprocessed routinely in primary care is incomplete, and such instruments cannot be excluded as a potential source of cross-infection.

  12. Dust Reprocessing of Stellar Tidal Disruption Flares

    NASA Astrophysics Data System (ADS)

    van Velzen, Sjoert; Gorjian, Varoujan; Krolik, Julian; Mendez, Alexander

    2015-10-01

    A stellar tidal disruption flare (TDF) occurs when a star gets too close to a supermassive black hole and is shredded into streams that are accreted. Traditionally, TDFs are observed at optical to soft X-ray wavelengths. We have recently made a discovery that opens a new (and unexpected) wavelength regime for the study of these flares: transient emission at 3.4 micron in WISE multi-epoch imaging. This dust reprocessing signal was not previously predicted, but will likely be of great importance to further our (limited) understanding the TDF emission mechanism. Since the radius of the IR-emitting shell is determined by the dust sublimation temperature, the break in the IR light curve can be used to measure the bolometric luminosity of the tidal flare. With the low-cadence WISE observations as a proof-of-concept, the time is ripe to use warm Spitzer observations to make a major breakthrough: we wish to obtain the first well-sampled light curve of dust reverberation by a stellar tidal flare. If successful, these observations will have lasting impact; near-future synoptic surveys (ZTF, LSST) will find thousands of TDFs per year, which can be followed-up by IR missions (JWST, WFIRST) to obtain a census of dust within the sphere of influence of quiescent supermassive black holes.

  13. 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.

  14. Evaluation of radioactivity release at Rokkasho reprocessing plant

    SciTech Connect

    Sugiyama, Hiroshi; Ishihara, Noriyuki; Maki, Akira

    2007-07-01

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

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

    PubMed

    Dunn, Debra

    2002-06-01

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

  16. 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)

  17. Containment/surveillance concepts for international safeguards in reprocessing plants

    SciTech Connect

    Bleck, M. E.; Cameron, C. P.; Camp, A. L.

    1980-01-01

    This paper examines the potential role of advanced containment/surveillance instrumentation systems for international safeguards in reprocessing plants. Several conceptual systems for the surveillance of containment boundary penetrations in a reference reprocessing plant are described and evaluated. The results of the evaluation aid in understanding the potential capabilities and limitations of containment/surveillance as an international safeguards concept in this type of facility.

  18. CODE Contribution to IGS Reprocessing: Status and Perspectives

    NASA Astrophysics Data System (ADS)

    Steigenberger, P.; Schaer, S.; Lutz, S.; Dach, R.; Ostini, L.; Hugentobler, U.; Bock, H.; Jäggi, A.; Meindl, M.; Thaller, D.

    2009-04-01

    Since 1992, the Center for Orbit Determination in Europe (CODE) is operated by Astronomisches Institut, Universität Bern (AIUB, Switzerland) in cooperation with the Bundesamt für Landestopographie (swisstopo, Switzerland), the Bundesamt für Kartographie und Geodäsie (BKG, Germany), and the Institut für Astronomische und Physikalische Geodäsie (IAPG) of Technische Universität München (TUM, Germany). Since the very beginning, CODE is an analyisis center of the International GNSS Service (IGS). The operational CODE processing is a rigorous GNSS analysis including GPS and GLONASS for all product lines of the IGS. The first CODE reprocessing run covers the time period January 1994 till December 2008. About 240 stations are included in the processing. Following the IGS guidelines for the current reprocessing effort, the analysis is limited to GPS. An extension to GLONASS is planned for the near future. Although these reprocessing activities are mainly performed by IAPG, full consistency with the currrent modeling setup of the operational CODE processing is realized. We will describe the processing strategy of the CODE GPS reprocessing and present selected results. The benefits of the reprocessing are demonstrated by comparisons with the operational CODE products and the results from the reprocessing efforts of other groups, e.g. the Potsdam Dresden Reprocesing (PDR). A special focus is put on the combination and realignment of P1-C1 differential code biases (DCBs).

  19. 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 .

  20. 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.

  1. 77 FR 55877 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-09-11

    ... Federal Register on March 16, 2012 (77 FR 15812), for a 60-day public comment period. The public comment... COMMISSION Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors AGENCY: Nuclear... plant startup, and power ascension tests for the condensate and feedwater systems in all types of...

  2. Light Water Reactor Safety Research Program. Semiannual report, April-September 1982

    SciTech Connect

    Berman, M.

    1983-10-01

    This report documents progress made in Light Water Reactor Safety research conducted by Division 6441 in the period from April 1982 to September 1982. The programs conducted under investigation include Core Concrete Interactions, Core Melt-Coolant Interactions, Containment Emergency Sump Performance, the Hydrogen Program, and Combustible Gas in Containment Program. 50 references.

  3. 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.

  4. 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)

  5. Disk reprocessing in three dimensions - emerging spectra and polarization

    NASA Astrophysics Data System (ADS)

    Goosmann, R.; Tamborra, F.; Marin, F.; Mouchet, M.; Dumont, A.; Dovčiak, M.

    2015-07-01

    We present the latest results from our ongoing computation of a new X-ray reprocessing model for irradiated, ionized accretion disks observed e.g. in active galactic nuclei and X-ray binaries. Combining a semi-analytical radiative transfer method (TITAN) with Monte-Carlo simulations (STOKES), we obtain a grid of reprocessed intensity and polarization spectra across an energy range of 0.1--100 keV with sufficient energy resolution to be applied to NuSTAR and even Astro-H observations. The model includes all polar and, for the first time, also azimuthal dependencies of the local incident and emission angles. As it was shown previously, these dependencies matter once the observed spectrum at infinity is computed by a relativistic ray-tracing routine. The Monte-Carlo treatment of the disk reprocessing also allows us to investigate in detail the impact of thermal and turbulent velocity on the line transfer, which can affect the observed amount of soft X-ray emission. Finally, we explore the impact of the high and low energy cutoff of the irradiating spectrum on the reprocessing and we determine the timing response of the disk spectrum as a function of photon energy. We conduct a comparison of our results with existing X-ray reprocessing models.

  6. 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.

  7. Greater-than-Class C low-level waste characterization. Appendix F: Greater-than-Class C low-level radioactive waste light water reactor projections

    SciTech Connect

    Tuite, P.; Tuite, K.; Levin, A.; O`Kelley, M.

    1991-08-01

    This study characterizes potential greater-than-Class C low-level radioactive waste streams, estimates the amounts of waste generated, and estimates their radionuclide content and distribution. Several types of low-level radioactive wastes produced by light water reactors were identified in an earlier study as being potential greater-than-Class C low-level waste, including specific activated metal components and certain process wastes in the form of cartridge filters and decontamination resins. Light water reactor operating parameters and current management practices at operating plants were reviewed and used to estimate the amounts of potential greater-than-Class C low-level waste generated per fuel cycle. The amounts of routinely generated activated metal components and process waste were estimated as a function of fuel cycle. Component-specific radionuclide content and distribution was calculated for activated metals components. Empirical data from actual low-level radioactive waste streams were used to estimate radionuclide content and distribution for process wastes. The greater-than-Class C low-level waste volumes that could be generated through plant closure were also estimated, along with volumes and activities for potential greater-than-Class C activated metals generated at decommissioning.

  8. 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.

  9. Testing of the Multi-Application Small Light Water Reactor (MASLWR) Passive Safety Systems

    SciTech Connect

    Reyes, Jose N.; Groome, John; Woods, Brian G.; Young, Eric; Abel, Kent; Yao, You; Yeon Jong Yoo

    2006-07-01

    Experimental thermal hydraulic research has been conducted at Oregon State University for the purpose of assessing the performance of a new reactor design concept, the Multi-application Small Light Water Reactor (MASLWR). MASLWR is a pressurized light water reactor that uses natural circulation in both normal and transient operation. The purpose of the OSU MASLWR Test Facility is to assess the operation of the MASLWR under normal full pressure and full temperature conditions and to assess the passive safety systems under transient conditions. The data generated by the testing program will be used to assess computer code calculations and to provide a better understanding of the thermal-hydraulic phenomena in the design of the MASLWR NSSS. During this testing program, four tests were conducted at the OSU MASLWR Test Facility. These tests included one design basis accident and one beyond design basis accident. Plant start up, normal operation and shut down evolutions were also examined. (authors)

  10. Flow-induced vibration for light water reactors. Progress report, October 1980-December 1980

    SciTech Connect

    Torres, M.R.

    1981-09-01

    Flow-Induced Vibration for Light Water Reactors (FIV for LWRs) is a four-year program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, general scaling laws to improve the accuracy of reduced-scale tests, and the identification of high FIV risk areas. The program is managed by the General Electric Nuclear Power Systems Engineering Department and has three major contributors: General Electric Nuclear Power Systems Engineering Department (NPSED), General Electric Corporate Research and Development (CR and D) and Argonne National Laboratory (ANL). The program commenced December 1, 1976. This progress report summarizes the accomplishments achieved during the period from October 1980 to December 1980.

  11. Flow-induced vibration for light-water reactors. Progress report, April-June 1981

    SciTech Connect

    Torres, M. R.

    1981-10-01

    Flow-Induced Vibration for Light Water Reactors (FIV for LWRs) is a program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, and general scaling laws to improve the accuracy of reduced-scale tests, and through the identification of high FIV risk areas. The program is managed by the General Electric Nuclear Power Systems Engineering Department and has three major contributors: General Electric Nuclear Power Systems Engineering Department (NPSED), General Electric Corporate Research and Development (CR and D) and Argonne National Laboratory (ANL). The program commenced December 1, 1976. This progress report summarizes the accomplishments achieved during the period from April 1981 to June 1981.

  12. Flow-induced vibration for light water reactors. Final progress report, July 1981-September 1981

    SciTech Connect

    Torres, M.R.

    1981-11-01

    Flow-Induced Vibration for Light Water Reactors (FIV for LWRs) is a program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, and general scaling laws to improve the accuracy of reduced-scale tests, and through the identification of high FIV risk areas. The program is managed by the General Electric Nuclear Power Systems Engineering Department and has three major contributors: General Electric Nuclear Power Systems Engineering Department (NPSED), General Electric Corporate Research and Development (CR and D) and Argonne National Laboratory (ANL). The program commenced December 1, 1976. This progress report summarizes the accomplishments achieved during the final period from July 1981 to September 1981. This is the last quarterly progress report to be issued for this program.

  13. End-of-life nondestructive examination of Light Water Breeder Reactor fuel rods (LWBR Development Program)

    SciTech Connect

    Gorscak, D.A.; Campbell, W.R.; Clayton, J.C.

    1987-10-01

    In-bundle and out-of-bundle (single rod) nondestructive examinations of Light Water Breeder Reactor fuel rods were performed. In-bundle examinations included visual examination and measurement of rod bow, rod-to-rod gaps, and rod removal forces. Out-of-bundle examinations included rod visuals and measurement of fuel rod length, diameter and ovality, cladding oxide and crud thickness, support grid induced cladding wear mark depth and volume, and fuel rod free hanging bow. The out-of-bundle examination also included ultrasonic inspection for cladding defects, neutron radiography for pellet integrity and plenum gap measurements, and gamma scans for instack axial gap screening and binary fuel stack length measurements. The measurements confirmed design predictions of fuel rod performance and provided evidence of excellent fuel rod performance for operation of Light Water Breeder Reactor to 29,047 effective full power hours (EFPH).

  14. 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.

  15. 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.

  16. 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

  17. Survey of Remote Area Monitoring Systems at U.S. Light-Water-Cooled Power Reactors

    SciTech Connect

    Kathren, R. L.; Mileham, A. P.

    1982-04-01

    A study was made of the capabilities and operating practices, including calibration, of remote area monitoring (RAM) systems at light-water-cooled power reactors in the United States. The information was obtained by mail questionaire. Specific design capabilities, including range, readout and alarm features are documented along with the numbers and location of detectors, calibration and operational procedures. Comments of respondents regarding RAM systems are also included.

  18. Zirconium alloy performance in light water reactors: A review of UK and Scandinavian experience

    SciTech Connect

    Pickman, D.O.

    1994-12-31

    Various aspects of zirconium alloy development for light water reactors in the UK and Scandinavia are reviewed, including the contribution made by some unique nuclear testing facilities. Among the problems encountered were the irradiation enhancement of corrosion and hydrogen pickup, crud deposition, iodine-induced stress-corrosion cracking on power ramping, and severe cladding deformation in loss-of-coolant accident conditions. The causes and behavior of defects, including hydride defects and fretting corrosion, are discussed.

  19. 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.

  20. Nondestructive verification with minimal movement of irradiated light-water-reactor fuel assemblies

    SciTech Connect

    Phillips, J.R.; Bosler, G.E.; Halbig, J.K.; Klosterbuer, S.F.; Menlove, H.O.

    1982-10-01

    Nondestructive verification of irradiated light-water reactor fuel assemblies can be performed rapidly and precisely by measuring their gross gamma-ray and neutron signatures. A portable system measured fuel assemblies with exposures ranging from 18.4 to 40.6 GWd/tU and with cooling times ranging from 1575 to 2638 days. Differences in the measured results for side or corner measurements are discussed. 25 figures, 20 tables.

  1. 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.

  2. 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.

  3. 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.

  4. End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

    SciTech Connect

    Richardson, K.D.

    1987-10-01

    Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580/sup 0/F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs.

  5. 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.

  6. Evaluation of fuel cycle scenarios on MOX fuel recycling in PWRs and SFRs

    SciTech Connect

    Carlier, B.; Caron-Charles, M.; Van Den Durpel, L.; Senentz, G.; Serpantie, J.P.

    2013-07-01

    Prospects on advanced fuel cycle scenario are considered for achieving a progressive integration of Sodium Fast Reactor (SFR) technology within the current French Pressurized Water Reactor (PWR) nuclear fleet, in a view to benefit from fissile material multi-recycling capability. A step by step process is envisioned, and emphasis is put on its potential implementation through the nuclear mass inventory calculations with the COSAC code. The overall time scale is not optimized. The first step, already implemented in several countries, the plutonium coming from the reprocessing of used Light Water Reactor (LWR) fuels is recycled into a small number of LWRs. The second step is the progressive introduction of the first SFRs, in parallel with the continuation of step 1. This second step lets to prepare the optimized multi recycling of MOX fuel which is considered in step 3. Step 3 is characterized by the introduction of a greater number of SFR and MOX management between EPR reactors and SFRs. In the final step 4, all the fleet is formed with SFRs. This study assesses the viability of each step of the overall scenario. The switch from one step to the other one could result from different constrains related to issues such as resources, waste, experience feedback, public acceptance, country policy, etc.

  7. 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.

  8. 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.

  9. 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)

  10. Research and Development of High Temperature Light Water Cooled Reactor Operating at Supercritical-Pressure in Japan

    SciTech Connect

    Yoshiaki Oka; Katsumi Yamada

    2004-07-01

    This paper summarizes the status and future plans of research and development of the high temperature light water cooled reactor operating at supercritical-pressure in Japan. It includes; the concept development; material for the fuel cladding; water chemistry under supercritical pressure; thermal hydraulics of supercritical fluid; and the conceptual design of core and plant system. Elements of concept development of the once-through coolant cycle reactor are described, which consists of fuel, core, reactor and plant system, stability and safety. Material studies include corrosion tests with supercritical water loops and simulated irradiation tests using a high-energy transmission electron microscope. Possibilities of oxide dispersion strengthening steels for the cladding material are studied. The water chemistry research includes radiolysis and kinetics of supercritical pressure water, influence of radiolysis and radiation damage on corrosion and behavior on the interface between water and material. The thermal hydraulic research includes heat transfer tests of single tube, single rod and three-rod bundles with a supercritical Freon loop and numerical simulations. The conceptual designs include core design with a three-dimensional core simulator and sub-channel analysis, and balance of plant. (authors)

  11. 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.

  12. 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.

  13. 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

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

    SciTech Connect

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

    1991-01-01

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

  15. Reprocessing of lithium titanate pebbles by graphite bed method

    NASA Astrophysics Data System (ADS)

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

    2015-04-01

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

  16. 76 FR 45268 - Reprocessing of Reusable Medical Devices

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-28

    ... methodologies, validation methodologies, and health care facility best practices. This is part of an ongoing..., reprocessing methodologies, validation methodologies, and health care facility best practices. As part of this... Silver Spring, MD. In the Federal Register of May 2, 2011 (76 FR 24495), FDA announced the workshop...

  17. 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…

  18. Assessment of photochemical applications to specific stages in Savannah River Plant actinide reprocessing streams

    SciTech Connect

    Toth, L.M.; Bell, J.T.; Mailen, J.C.; Dodson, K.E.

    1986-01-01

    The application of photochemical redox methods has been evaluated as a means of separating actinides in Purex reprocessing streams. This chemistry promises to eliminate many of the chemical reagents which are otherwise necessary to effect valence control of such actinides as plutonium and neptunium. The most promising processing stages of the Savannah River Plant reprocessing facility for feasibility testing of the photochemical method appear to be those in which the concentrations of the actinides in question are the lowest, thus minimizing the required amount of absorbed light. Although neptunium valence control appears to be feasible through the photochemical generation of nitrite ion, the urgency for its control is secondary to that of plutonium. Therefore, the Purex ''2nd-U-cycle feed'' stage has been selected as the top priority for testing the photochemical technique. The chemistry initially involves the photoreduction of U(VI) to U(IV), which in turn reduces Pu(IV) to Pu(III). The results of preliminary experiments on nitrate ion and uranyl ion photoreduction are also given. With less than 100 W of absorbed power, the rates of NO/sub 2//sup -/ and U/sup 4 +/ generation, respectively, appear ample to handle process requirements. 14 refs., 3 figs., 5 tabs.

  19. 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.

  20. 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

  1. 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

  2. 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.

  3. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    SciTech Connect

    Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

    2003-04-01

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

  4. An assessment of the effect on Olkiluoto repository capacity achievable with advanced fuel cycles

    SciTech Connect

    Juutilainen, P.; Viitanen, T.

    2013-07-01

    Previously a few scenarios have been simulated for transition from thermal to fast reactor fleet in Finland in order to determine how much the transuranic inventory could be reduced with the partitioning and transmutation (P-T) technologies. Those calculations, performed with COSI6 code developed by CEA, are extended in the present study, in which the effect of P-T on the capacity of the planned final disposal repository at Olkiluoto (Finland) is evaluated by taking into account the created fission products and transuranic residuals from the reprocessing operations. The decay heat is assumed to be the most restrictive factor in defining the waste disposal packing density. The repository capacity evaluation of this study is based on the comparison of the decay heats produced by the deposited waste in various scenarios. The reference scenario of this article involves only Light Water Reactors (LWR) in an open fuel cycle. The capacity requirement of the geological repository is estimated in a few closed fuel cycle scenarios, all including actinide transmutation with Fast Reactors (FR). The comparison between the P-T scenarios and reference is based on the decay heat production of the deposited waste. The COSI6 code is used for simulations to provide the repository decay heat curves. Applying the closed fuel cycle would change the disposal concept and schedule, because of which it is not quite straightforward to assess the impact of P-T on the capacity. However, it can be concluded that recycling the transuranic nuclides probably decreases the required volume for the disposal, but thermal dimensioning analysis is needed for more specific conclusions.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Generation-IV Multi-Application Small Light Water Reactor (MASLWR)

    SciTech Connect

    Modro, Slawomir Michael; Fisher, James Ebberly; Weaver, Kevan Dean; Babka, P.; Reyes, Johnny Paul; Groome, J.; Wilson, Gary Edward

    2002-04-01

    The Idaho National Engineering and Environmental Laboratory (INEEL), Nexant Inc. and the Oregon State University (OSU) have developed a Multi-Application Small Light Water Reactor (MASLWR) concept. The MASLWR is a small, safe and economic natural circulation pressurized light water reactor. MASLWR reactor module consists of an integral reactor/steam generator located in a steel cylindrical containment. The entire module is to be entirely shop fabricated and transported to site on most railways or roads. Two or more modules are located in a reactor building, each being submersed in a common, below grade cavity filled with water. For the most severe postulated accident, the volume of water in the cavity provides a passive ultimate heat sink for 3 or more days allowing the restoration of lost normal active heat removal systems. MASLWR thermal power of a single module is 150 MWt, primary system pressure 10.5 MPa, steam pressure1.52 MPa and the net electrical output is 35 - 50 MWe.

  12. Generation-IV Multi-Application Small Light Water Reactor (MASLWR)

    SciTech Connect

    Modro, S. Michael; Fisher, James; Weaver, Kevan; Babka, Pierre; Reyes, Jose; Groome, John

    2002-07-01

    The Idaho National Engineering and Environmental Laboratory (INEEL), Nexant Inc. and the Oregon State University (OSU) have developed a Multi-Application Small Light Water Reactor (MASLWR) concept. The MASLWR is a small, safe and economic natural circulation pressurized light water reactor. MASLWR reactor module consists of an integral reactor/steam generator located in a steel cylindrical containment. The entire module is to be entirely shop fabricated and transported to site on most railways or roads. Two or more modules are located in a reactor building, each being submersed in a common, below grade cavity filled with water. For the most severe postulated accident, the volume of water in the cavity provides a passive ultimate heat sink for 3 or more days allowing the restoration of lost normal active heat removal systems. MASLWR thermal power of a single module is 150 MWt, primary system pressure 10.5 MPa, steam pressure 1.52 MPa and the net electrical output is 35 - 50 MWe. (authors)

  13. 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)

  14. 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)

  15. 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

  16. 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.

  17. Echo Tomography of Reprocessing Sites in X-Ray Binaries

    NASA Technical Reports Server (NTRS)

    Patterson, Joseph; Haswell, Carole

    1998-01-01

    We discovered correlated rapid variability between the optical/UV and X-ray emission for the first time in a soft X-ray transient, GRO J1655-40. Hubble Space Telescope light curves show features similar to those seen by the Rossi X-ray Timing Explorer, but with a mean delay of up to 10 - 20 s. We interpret the correlation as the result of reprocessing of X-rays into optical and UV emission, with a delay owing to finite light travel time; this assumption enables us to perform echo mapping of the system. The time-delay distribution has a mean of 14.6 +/-1.4 s and a dispersion of 10.5+/-1.9 s at binary phase 0.4. This establishes that the reprocessing region is the accretion disk around the compact star, rather than the mass-donating secondary. These results have been published.

  18. Remotex and servomanipulator needs in nuclear fuel reprocessing plants

    SciTech Connect

    Garin, J.

    1981-01-01

    Work on the conceptual design of a pilot-scale plant for reprocessing breeder reactor fuels is being performed at Oak Ridge National Laboratory. The plant design will meet all current federal regulations for repocessing plants and will serve as prototype for future production plants. A unique future of the concept is the incorporation of totally remote operation and maintenance of the process equipment within a large barn-like hot cell. This approach, caled Remotex, utilizes servomanipulators coupled with television viewing to extend man's capabilities into the hostile cell environment. The Remotex concept provides significant improvements for fuel reprocessing plants and other nuclear facilities in the areas of safeguarding nuclear materials, reducing radiation exposure, improving plant availability, recovering from unplanned events, and plant decommissioning.

  19. 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.

  20. 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.

  1. Potential for meeting IAEA safeguards goals for reprocessing

    SciTech Connect

    Bambas, K.J.

    1980-09-01

    The author concludes that: Safeguards which allow the meeting of IAEA uranium and plutonium diversion detection goals for large-scale reprocessing plants are achievable. Concepts exists for ways in which IAEA inspectors could verify data received from operations of such facilities. Additional work is needed, particularly in the futher development of on-line instruments. Computerized near real-time accounting systems may provide significant operational benefits in addition to being required for safeguards purposes.

  2. 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.

  3. High-temperature reprocessing of petroleum oily sludges

    SciTech Connect

    Hahn, W.J. )

    1994-08-01

    Crude oil tank bottoms and other petroleum oily sludges and emulsions containing paraffins and volatile hydrocarbons can be economically reprocessed with heavy-oil dehydration facilities to recover residual hydrocarbons and to achieve volume reductions. The main factors affecting the use of this alternative are (1) the characteristics of the sludges requiring treatment, (2) the availability of waste heat or existing high temperature (> 350 F) dehydration facilities, (3) air emissions from the process, and (4) effluent criteria for treated residues. This paper discusses operational variables that affect high-temperature reprocessing (HTR) and illustrates an application of the process. The example pilot project evaluated the feasibility of high-temperature reprocessing for tank-bottom sludges and skim oils from Kern County, CA, light- (>30[degree] API) oil-producing leases. The process performance was quantified in terms of general operating parameters (flash point and paraffin, oil, water, and solids content); specific constituent analyses for benzene, toluene, ethylbenzene, and xylene (BTEX); and analyses for total petroleum hydrocarbons (TPH) content. Information on the percent removal of these parameters, characteristics of the treated residues, and the hydrocarbon recovery efficiency of the process are presented.

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

    SciTech Connect

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

    2013-08-01

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

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

    PubMed

    Pioro, L S; Pioro, I L

    2004-01-01

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

  6. 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.

  7. 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.

  8. Dual phase magnesia-zirconia ceramics for light water reactor inert matrix fuel.

    SciTech Connect

    Medvedev, P.; Meyer, M. K.; Nuclear Technology

    2005-01-01

    To address the low thermal conductivity of the ZrO{sub 2}-based inert matrix fuel and the instability in water of the MgO-based inert matrix fuel, the dual-phase MgO-ZrO{sub 2} ceramics are proposed as a matrix for light water reactor fuel for actinide transmutation and Pu burning. It is envisioned that in a dual-phase system MgO will act as efficient heat conductor while ZrO{sub 2} will provide protection from the coolant attack. This paper describes results of fabrication, characterization and hydration testing of MgO-ZrO{sub 2} ceramics containing 30-70 wt% of MgO.

  9. Criticality benchmark guide for light-water-reactor fuel in transportation and storage packages

    SciTech Connect

    Lichtenwalter, J.J.; Bowman, S.M.; DeHart, M.D.; Hopper, C.M.

    1997-03-01

    This report is designed as a guide for performing criticality benchmark calculations for light-water-reactor (LWR) fuel applications. The guide provides documentation of 180 criticality experiments with geometries, materials, and neutron interaction characteristics representative of transportation packages containing LWR fuel or uranium oxide pellets or powder. These experiments should benefit the U.S. Nuclear Regulatory Commission (NRC) staff and licensees in validation of computational methods used in LWR fuel storage and transportation concerns. The experiments are classified by key parameters such as enrichment, water/fuel volume, hydrogen-to-fissile ratio (H/X), and lattice pitch. Groups of experiments with common features such as separator plates, shielding walls, and soluble boron are also identified. In addition, a sample validation using these experiments and a statistical analysis of the results are provided. Recommendations for selecting suitable experiments and determination of calculational bias and uncertainty are presented as part of this benchmark guide.

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. Technical support for the hydrogen control requirement for the EPRI advanced light water reactor requirements document

    SciTech Connect

    Plys, M.G.

    1988-01-01

    Hydrogen could be a significant contributor to severe-accident risk if hydrogen generation and combustion were to lead to containment failure and resulting release of fission products. To eliminate hydrogen as a significant risk contributor for advanced light water reactors (ALWRs), the Electric Power Research Institute (EPRI) ALWR requirements document has imposed a hydrogen control requirement. This requirement specifies an upper limit of 13 dry vol% for the allowable hydrogen concentration in containment. The requirement also considers hydrogen generation during severe accidents and states an upper bound on the hydrogen source equivalent to that generated by oxidizing 75% of the active cladding (commonly stated as 75% metal/water reaction (MWR)). The purpose of this paper is to technically support and substantiate the EPRI ALWR hydrogen requirement. The current understanding of hydrogen generation and combustion is evaluated as it applies to reactor systems, and it is concluded that both experimental results and analytical methods provide a sound technical basis for the requirement.

  15. 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.

  16. 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.

  17. 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.

  18. Modeling of microstructure evolution in austenitic stainless steels irradiated under light water reactor condition

    NASA Astrophysics Data System (ADS)

    Gan, J.; Was, G. S.; Stoller, R. E.

    2001-10-01

    A model for microstructure development in austenitic alloys under light water reactor irradiation conditions is described. The model is derived from the model developed by Stoller and Odette to describe microstructural evolution under fast neutron or fusion reactor irradiation conditions. The model is benchmarked against microstructure measurements in 304 and 316 SS irradiated in a boiling water reactor core using one material-dependent and three irradiation-based parameters. The model is also adapted for proton irradiation at higher dose rate and higher temperature and is calibrated against microstructure measurements for proton irradiation. The model calculations show that for both neutron and proton irradiations, in-cascade interstitial clustering is the driving mechanism for loop nucleation. The loss of interstitial clusters to sinks by interstitial cluster diffusion was found to be an important factor in determining the loop density. The model also explains how proton irradiation can produce an irradiated dislocation microstructure similar to that in neutron irradiation.

  19. Thermodynamic analysis of cesium and iodine behavior in severe light water reactor accidents

    NASA Astrophysics Data System (ADS)

    Minato, Kazuo

    1991-11-01

    In order to understand the release and transport behavior of cesium (Cs) and iodine (I) in severe light water reactor accidents, chemical forms of Cs and I in steam-hydrogen mixtures were analyzed thermodynamically. In the calculations reactions of boron (B) with Cs were taken into consideration. The analysis showed that B plays an important role in determining chemical forms of Cs. The main Cs-containing species are CsBO 2(g) and CsBO 2(l), depending on temperature. The contribution of CsOH(g) is minor. The main I-containing species are HI(g) and CsI(g) over the wide ranges of the parameters considered. Calculations were also carried out under the conditions of the Three Mile Island Unit 2 accident.

  20. Nanostructure of Metallic Particles in Light Water Reactor Used Nuclear Fuel

    SciTech Connect

    Buck, Edgar C.; Mausolf, Edward J.; Mcnamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2015-03-11

    The extraordinary nano-structure of metallic particles in light water reactor fuels points to possible high reactivity through increased surface area and a high concentration of high energy defect sites. We have analyzed the metallic epsilon particles from a high burn-up fuel from a boiling water reactor using transmission electron microscopy and have observed a much finer nanostructure in these particles than has been reported previously. The individual round particles that varying in size between ~20 and ~50 nm appear to consist of individual crystallites on the order of 2-3 nm in diameter. It is likely that in-reactor irradiation induce displacement cascades results in the formation of the nano-structure. The composition of these metallic phases is variable yet the structure of the material is consistent with the hexagonal close packed structure of epsilon-ruthenium. These findings suggest that unusual catalytic behavior of these materials might be expected, particularly under accident conditions.

  1. Estimation of fatigue strain-life curves for austenitic stainless steels in light water reactor environments.

    SciTech Connect

    Chopra, O. K.; Smith, J. L.

    1998-02-12

    The ASME Boiler and Pressure Vessel Code design fatigue curves for structural materials do not explicitly address the effects of reactor coolant environments on fatigue life. Recent test data indicate a significant decrease in fatigue lives of austenitic stainless steels (SSs) in light water reactor (LWR) environments. Unlike those of carbon and low-alloy steels, environmental effects on fatigue lives of SSs are more pronounced in low-dissolved-oxygen (low-DO) water than in high-DO water, This paper summarizes available fatigue strain vs. life data on the effects of various material and loading variables such as steel type, DO level, strain range, and strain rate on the fatigue lives of wrought and cast austenitic SSs. Statistical models for estimating the fatigue lives of these steels in LWR environments have been updated with a larger data base. The significance of the effect of environment on the current Code design curve has been evaluated.

  2. 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.

  3. Validation of scale-4 for LWR (Light Water Reactor) fuel in transportation and storage cask conditions

    SciTech Connect

    Bowman, S.M.; Parks, C.V. ); Bierman, S.R. )

    1990-01-01

    This paper presents the results of criticality calculations performed to validate the recently released SCALE-4 modular code system for light water reactor (LWR) fuel under various conditions typical of transportation and storage casks. The modifications in SCALE-4 include NITAWL-II, an updated version of the NITAWL code that performs resonance self-shielding calculations using the Nordeim Integral Treatment. In order to validate SCALE-4 with the new resonance self-shielding treatment, the CSAS4 control module was used to calculate the effective neutron multiplication factor (k{sub eff}) via the BONAMI{sup 3}, NITAWL-II, and KENO V.a{sup 4} codes. The cross section library used was the 27 group ENDF/B-IV library, which has been updated for use with NITAWL-II. 12 refs., 1 tab.

  4. Ethical and hygiene aspects of the reprocessing of medical devices in Germany.

    PubMed

    Kramer, Axel; Assadian, Ojan

    2008-01-01

    Based on safety and quality principles, for each medical device (MD), regardless of its declared status as single- or multi-use device, careful considerations must be made. This includes assessment whether reprocessing is economical and ecological meaningful, and technical feasible. So far, however, in Germany reprocessing of declared single use MD is legally allowed, provided that the above aspects are well covered. The purpose of this paper is to elucidate, when circumstances allow reprocessing of declared single-use MD. For reprocessing of single use MD the following preconditions must be fulfilled:The security level of the reprocessed MD must be equivalent to the status of the newly delivered item; this means that a patient is not exposed to a higher risk through a reprocessed disposable MD than through the new, i.e. un-processed product.The reprocessing must be based on a detailed risk assessment and risk analysis, and must be described in detail regarding selection of the reprocessing method. Additionally, all necessary safety- and quality assurance measures must be stated.The reprocessing measure needs to be accompanied with a quality management system which determines and documents the responsibility of all stages of reprocessing; where the corresponding reprocessing procedures are well defined; and the efficacy of the procedure is proven by product-specific or product-group-specific tests and reports. The process must be validated according to recognised methods of science and technology, taking into account potential negative influences of the reprocessing on the properties of the material and the technical and functional safety. For reprocessing of MDs of the category Critical C the quality assurance must be certified by an accredited certifying body. PMID:20204097

  5. 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.

  6. 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.

  7. 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. PMID:23716682

  8. 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

  9. 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

  10. 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.

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

    SciTech Connect

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

    1985-07-01

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

  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. High-level disinfection of gastrointestinal endoscope reprocessing

    PubMed Central

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

    2015-01-01

    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. PMID:25699232

  14. 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.

  15. 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. PMID:25699232

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

    SciTech Connect

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

    1980-12-01

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

  17. 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.

  18. 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.

  19. 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.

  20. Seasonal signals in the reprocessed GPS coordinate time series

    NASA Astrophysics Data System (ADS)

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

    2008-12-01

    The global (IGS) and regional (EPN) CGPS time series have already been studied in detail by several authors to analyze the periodic signals and noise present in the long term displacement series. The comparisons indicated that the amplitude and phase of the CGPS derived seasonal signals mostly disagree with the surface mass redistribution models. The CGPS results are highly overestimating the seasonal term, only about 40% of the observed annual amplitude can be explained with the joint contribution of the geophysical models (Dong et al. 2002). Additionally the estimated amplitudes or phases are poorly coherent with the models, especially at sites close to coastal areas (van Dam et al, 2007). The conclusion of the studies was that the GPS results are distorted by analysis artifacts (e.g. ocean tide loading, aliasing of unmodeled short periodic tidal signals, antenna PCV models), monument thermal effects and multipath. Additionally, the GPS series available so far are inhomogeneous in terms of processing strategy, applied models and reference frames. The introduction of the absolute phase center variation (PCV) models for the satellite and ground antennae in 2006 and the related reprocessing of the GPS precise orbits made a perfect ground and strong argument for the complete re-analysis of the GPS observations from global to local level of networks. This enormous work is in progress within the IGS and a pilot analysis was already done for the complete EPN observations from 1996 to 2007 by the MUT group (Military University of Warsaw). The quick analysis of the results proved the expectations and the superiority of the reprocessed data. The noise level (weekly coordinate repeatability) was highly reduced making ground for the later analysis on the daily solution level. We also observed the significant decrease of the seasonal term in the residual coordinate time series, which called our attention to perform a repeated comparison of the GPS derived annual periodicity

  1. 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)

  2. 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.

  3. 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. PMID:12963330

  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. 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

  6. 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

  7. Deactivating a major nuclear fuels reprocessing facility cost effectively

    SciTech Connect

    LeBaron, G.J.

    1997-08-15

    This paper describes three key processes used in deactivating the Plutonium Uranium Extraction (PUREX) Facility, a large, complex nuclear reprocessing facility, 15 months ahead of schedule and $77 million under budget. The organization was reengineered to refine its business processes and more effectively organize around the deactivation work scope. Multi-disciplined work teams were formed to be self-sufficient and empowered to make decisions and perform work. A number of benefits were realized by reengineering. A comprehensive process to develop end points which clearly identified specific results and the post-project facility configuration was developed so all areas of a facility were addressed. Clear and specific end points allowed teams to focus on completing deactivation activities and helped ensure there were no unfulfilled end-of-project expectations. The RCRA regulations require closure of permitted facilities within 180 days after cessation of operations which may essentially necessitate decommissioning. A more cost effective approach was adopted which significantly reduced risk to human health and the environment by taking the facility to a passive, safe, inexpensive-to-maintain surveillance and maintenance condition (deactivation) prior to disposition. PUREX thus became the first large reprocessing facility with active TSD [treatment, storage, and disposal] units to be deactivated under the RCRA regulations.

  8. 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)

  9. 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)

  10. 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.; Xu, Kevin G.; Wachs, Daniel M.

    2016-09-28

    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

  11. 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.

  12. Accident source terms for Light-Water Nuclear Power Plants. Final report

    SciTech Connect

    Soffer, L.; Burson, S.B.; Ferrell, C.M.; Lee, R.Y.; Ridgely, J.N.

    1995-02-01

    In 1962 tile US Atomic Energy Commission published TID-14844, ``Calculation of Distance Factors for Power and Test Reactors`` which specified a release of fission products from the core to the reactor containment for a postulated accident involving ``substantial meltdown of the core``. This ``source term``, tile basis for tile NRC`s Regulatory Guides 1.3 and 1.4, has been used to determine compliance with tile NRC`s reactor site criteria, 10 CFR Part 100, and to evaluate other important plant performance requirements. During the past 30 years substantial additional information on fission product releases has been developed based on significant severe accident research. This document utilizes this research by providing more realistic estimates of the ``source term`` release into containment, in terms of timing, nuclide types, quantities and chemical form, given a severe core-melt accident. This revised ``source term`` is to be applied to the design of future light water reactors (LWRs). Current LWR licensees may voluntarily propose applications based upon it.

  13. Nanostructure of metallic particles in light water reactor used nuclear fuel

    NASA Astrophysics Data System (ADS)

    Buck, Edgar C.; Mausolf, Edward J.; McNamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2015-06-01

    An extraordinary nano-structure has been observed in the metallic (Mo-Tc-Ru-Rh-Pd) particles that are known to form during irradiated in light water nuclear reactor fuels. This structure points possible high catalytic reactivity through the occurrence of a very high surface area as well as defect sites. We have analyzed separated metallic particles from dissolved high burn-up spent nuclear fuel using scanning and transmission electron microscopy. The larger particles vary in diameter between ∼10 and ∼300 nm and possess a hexagonally close packed epsilon-ruthenium structure. These particles are not always single crystals but often consist of much smaller crystallites on the order of 1-3 nm in diameter with evidence suggesting the occurrence of some amorphous regions. It is possible that neutron irradiation and fission product recoils generated the unusual small crystallite size. The composition of the metallic particles was variable with low levels of uranium present in some of the particles. We hypothesize that the uranium may have induced the formation of the amorphous (or frustrated) metal structure. This unique nano-structure may play an important role in the environmental behavior of nuclear fuels.

  14. 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.

  15. 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.

  16. Safeguards and security requirements for weapons plutonium disposition in light water reactors

    SciTech Connect

    Thomas, L.L.; Strait, R.S.

    1994-10-01

    This paper explores the issues surrounding the safeguarding of the plutonium disposition process in support of the United States nuclear weapons dismantlement program. It focuses on the disposition of the plutonium by burning mixed oxide fuel in light water reactors (LWR) and addresses physical protection, material control and accountability, personnel security and international safeguards. The S and S system needs to meet the requirements of the DOE Orders, NRC Regulations and international safeguards agreements. Experience has shown that incorporating S and S measures into early facility designs and integrating them into operations provides S and S that is more effective, more economical, and less intrusive. The plutonium disposition safeguards requirements with which the US has the least experience are the implementation of international safeguards on plutonium metal; the large scale commercialization of the mixed oxide fuel fabrication; and the transportation to and loading in the LWRs of fresh mixed oxide fuel. It is in these areas where the effort needs to be concentrated if the US is to develop safeguards and security systems that are effective and efficient.

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

    SciTech Connect

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

    1999-05-01

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

  18. Nuclide Composition Benchmark Data Set for Verifying Burnup Codes on Spent Light Water Reactor Fuels

    SciTech Connect

    Nakahara, Yoshinori; Suyama, Kenya; Inagawa, Jun; Nagaishi, Ryuji; Kurosawa, Setsumi; Kohno, Nobuaki; Onuki, Mamoru; Mochizuki, Hiroki

    2002-02-15

    To establish a nuclide composition benchmark data set for the verification of burnup codes, destructive analyses of light water reactor spent-fuel samples, which were cut out from several heights of spent-fuel rods, were carried out at the analytical laboratory at the Japan Atomic Energy Research Institute. The 16 samples from three kinds of pressurized water reactor (PWR) fuel rods and the 18 samples from two boiling water reactor (BWR) fuel rods were examined. Their initial {sup 235}U enrichments and burnups were from 2.6 to 4.1% and from 4 to 50 GWd/t, respectively. One PWR fuel rod and one BWR fuel rod contained gadolinia as a burnable poison. The measurements for more than 40 nuclides of uranium, transuranium, and fission product elements were performed by destructive analysis using mass spectrometry, and alpha-ray and gamma-ray spectrometry. Burnup for each sample was determined by the {sup 148}Nd method. The analytical methods and the results as well as the related irradiation condition data are compiled as a complete benchmark data set.

  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. EPRI advanced light water reactor requirements document to establish standardization principles

    SciTech Connect

    Stahlkopf, K.E.; Noble, D.M.

    1986-01-01

    The essential element of the EPRI Advanced Light Water Reactor (ALWR) Program is its Utility Requirements Document. The requirements document consists of utility-authorized specifications for a future nuclear plant. The document will actually consist of 13 chapters. Each of these chapters will contain requirements for several plant systems so that all plant systems are addressed with requirements for a future LWR (either a BWR or a PWR). In addition, a chapter on plant arrangement and another on instrumentation and control cover overall aspects of plant design and construction requirements. The requirements document is intended to specify the features utilities consider essential for a future plant. An ALWR Utility Steering Committee has been established to provide for review and direction to the production of the utility requirements. The committee is formed by executives of EPRI member utilities with significant nuclear design, construction, operation, and maintenance experience. Completing the requirements document will involve obtaining favorable safety evaluation reports from the NRC on each chapter and the complete document. When completed, the ALWR Utility Requirements Document will be an NRC-endorsed expression of utility needs for future plants.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 76 FR 13605 - Notice of Availability of Draft Waste Incidental to Reprocessing Evaluation for the Vitrification...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-14

    ... materials. It was used to solidify high-level waste which had been generated by commercial reprocessing of... vitrified the waste (combined it at a high temperature with borosilicate glass) and transferred the molten... waste from reprocessing of spent nuclear fuel and certain treatment material) at the West...

  7. 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

    ... WVDP Act. To solidify the waste, DOE vitrified the waste (combined it at a high temperature with... vitrifying waste from reprocessing of spent nuclear fuel and certain treatment material at the West Valley... incidental to reprocessing and thus are not high-level radioactive waste (HLW) and may be managed...

  8. 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…

  9. 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)

  10. Recommendations to resolve inconsistent guidelines for the reprocessing of sheathed and unsheathed rigid laryngoscopes.

    PubMed

    Muscarella, Lawrence F

    2007-04-01

    Neither a consensus statement nor a formal set of step-by-step guidelines for reprocessing rigid laryngoscopes have been published or endorsed by professional organizations. Several published guidelines, standards, and clinical reports were reviewed to evaluate the risk of nosocomial infection associated with the use of rigid laryngoscopes, to determine their minimum reprocessing requirements. This review found that the recommendations of some guidelines and standards for reprocessing rigid laryngoscopes are incomplete, inadequate, and inconsistent with one another, and that current practices for reprocessing rigid laryngoscopes are reported to be inadequate and lack standardization. It is recommended that a consensus statement be developed that standardizes the reprocessing of rigid laryngoscopes and requires cleaning followed by high-level disinfection (or sterilization) and drying of the rigid laryngoscope's blade and handle to prevent nosocomial infection, regardless of whether a protective barrier or sheath is used during the procedure.

  11. 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.

  12. 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

  13. Modeling the behavior of a light-water production reactor target rod

    SciTech Connect

    Sherwood, D.J.

    1992-03-01

    Pacific Northwest Laboratory has been conducting a series of in-reactor experiments in the Idaho National Engineering Laboratory (INEL) Advanced Test Reactor (ATR) to determine the amount of tritium released by permeation from a target rod under neutron irradiation. The model discussed in this report was developed from first principles to model the behavior of the first target rod irradiated in the ATR. The model can be used to determine predictive relationships for the amount of tritium that permeates through the target rod cladding during irradiation. The model consists of terms and equations for tritium production, gettering, partial pressure, and permeation, all of which are described in this report. The model addressed only the condition of steady state and features only a single adjustable parameter. The target rod design for producing tritium in a light-water reactor was tested first in the WC-1 in-reactor experiment. During irradiation, tritium is generated in the target rod within the ceramic lithium target material. The target rod has been engineered to limit the release of tritium to the reactor coolant during normal operations. The engineered features are a nickel-plated Zircaloy-4 getter and a barrier coating on the cladding surfaces. The ceramic target is wrapped with the getter material and the resulting ``pencils`` are inserted into the barrier coated cladding. These features of the rod are described in the report, along with the release of tritium from the ceramic target. The steady-state model could be useful for the design procedure of target rod components.

  14. Modeling the behavior of a light-water production reactor target rod

    SciTech Connect

    Sherwood, D.J.

    1992-03-01

    Pacific Northwest Laboratory has been conducting a series of in-reactor experiments in the Idaho National Engineering Laboratory (INEL) Advanced Test Reactor (ATR) to determine the amount of tritium released by permeation from a target rod under neutron irradiation. The model discussed in this report was developed from first principles to model the behavior of the first target rod irradiated in the ATR. The model can be used to determine predictive relationships for the amount of tritium that permeates through the target rod cladding during irradiation. The model consists of terms and equations for tritium production, gettering, partial pressure, and permeation, all of which are described in this report. The model addressed only the condition of steady state and features only a single adjustable parameter. The target rod design for producing tritium in a light-water reactor was tested first in the WC-1 in-reactor experiment. During irradiation, tritium is generated in the target rod within the ceramic lithium target material. The target rod has been engineered to limit the release of tritium to the reactor coolant during normal operations. The engineered features are a nickel-plated Zircaloy-4 getter and a barrier coating on the cladding surfaces. The ceramic target is wrapped with the getter material and the resulting pencils'' are inserted into the barrier coated cladding. These features of the rod are described in the report, along with the release of tritium from the ceramic target. The steady-state model could be useful for the design procedure of target rod components.

  15. 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.

  16. 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.

  17. 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.

  18. Adaptation of gas tagging for failed fuel identification in light water reactors

    SciTech Connect

    Lambert, J.D.B.; Gross, K.C.; Depiante, E.V.; Callis, E.L.; Egebrecht, P.M.

    1996-03-01

    This paper discusses experience with noble gas tagging and its adaptation to commercial reactors. It reviews the recent incidence of fuel failures in light water reactors, and methods used to identify failures, and concludes that the on-line technique of gas tagging could significantly augment present flux tilting, sipping and ultrasonic testing of assemblies. The paper describes calculations on tag gas stability in-reactor, and tag injection tests that were carried out collaboratively with Commonwealth Edison Company in the Byron-2 pressurized water reactor (P%a) and with Duke Power Company and Babcock and Wilcox Fuel Company in the Oconee-2 PWM. The tests gave information on: (a) noble gas concentration dynamics as the tag gases were dissolved in and eventually removed from subsystems of the RCS; and (b) the suitability of candidate Ar, Ne, Kr and Xe isotopes for tagging PWR fuel. It was found that the activity of Xe{sup 125} (the activation product of the tag isotope Xe{sup 124}) acted as a ``tag of a tag`` and tracked gas through the reactor; measured activities are being used to model gas movement in the RCS. Several interference molecules (trace contaminants normally present at sub-ppM concentrations in RCS samples) and entrained air in the RCS were found to affect mass spectrometer sensitivity for tag isotopes. In all instances the contaminants could be differentiated from the tag isotopes by operating the mass spectrometer at high resolution (2500). Similarly, it was possible to distinguish all the candidate tag gases against a high background of air. The test results suggested, however, that for routine analysis a high resolution static mass spectrometer will be preferable to the dynamic instrument used for the present analyses.

  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. 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.

  1. 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.

  2. Detection and characterization of flaws in segments of light water reactor pressure vessels

    SciTech Connect

    Cook, K.V.; Cunningham, R.A. Jr.; McClung, R.W.

    1987-01-01

    Studies have been conducted to determine flaw density in segments cut from light water reactor (LWR) pressure vessels as part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology (HSST) Program. Segments from the Hope Creek Unit 2 vessil and the Pilgrim Unit 2 Vessel were purchased from salvage dealers. Hope Creek was a boiling water reactor (BWR) design and Pilgrim was a pressurized water reactor (PWR) design. Neither were ever placed in service. Objectives were to evaluate these LWR segments for flaws with ultrasonic and liquid penetrant techniques. Both objectives were successfully completed. One significant indication was detected in a Hope Creek seam weld by ultrasonic techniques and characterized by further analyses terminating with destructive correlation. This indication (with a through-wall dimension of approx.6 mm (approx.0.24 in.)) was detected in only 3 m (10 ft) of weldment and offers extremely limited data when compared to the extent of welding even in a single pressure vessel. However, the detection and confirmation of the flaw in the arbitrarily selected sections implies the Marshall report estimates (and others) are nonconservative for such small flaws. No significant indications were detected in the Pilgrim material by ultrasonic techniques. Unfortunately, the Pilgrim segments contained relatively little weldment; thus, we limited our ultrasonic examinations to the cladding and subcladding regions. Fluorescent liquid penetrant inspection of the cladding surfaces for both LWR segments detected no significant indications (i.e., for a total of approximately 6.8 m/sup 2/ (72 ft/sup 2/) of cladding surface).

  3. Development and Deployment Strategy for a Small Advanced Light Water Reactor

    SciTech Connect

    Modro, S. Michael; Reith, Raymond; Babka, Pierre

    2002-07-01

    This paper discusses development and deployment strategies for the modular Multi-Application Small Light Water Reactor (MASLWR). Modularity, small size, capability to transport whole modules including containment on road or by rail, simplicity and safety of this reactor allows innovative deployment strategies for a variety of applications. A larger plant may be constructed of many independent power generation units. The multi-module plant is intended to be operated as a base-load plant. Each reactor is to be operated at full load. However, in response to changes in power demand individual units can brought on line or shut down. A larger plant can be built in small increments to match the power demand balancing capital commitments with revenues from sales of electricity. Also, an unplanned shutdown of a reactor only affects a relatively small portion of the total plant capacity. Simplification of MASLWR design and extensive use of modularization coupled with factory fabrication will result in improved productivity of fieldwork and improved quality achieved in a factory environment. The initial MASLWR design concept development has been completed under the U.S. DOE (Department of Energy) Nuclear Energy Research Initiative (NERI) project. This paper discusses a strategy for developing and deploying a MASLWR plant by 2015. This schedule is realistic because the plant design relies on existing industrial experience and manufacturing capabilities. The development strategy consists of the following elements: concept confirmation through testing (under the NERI program a scaled integral test facility has been constructed and initial testing performed), design concept optimization, and design certification based on prototype testing. (authors)

  4. 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

  5. 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.

  6. 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

  7. Design and analysis of a nuclear reactor core for innovative small light water reactors

    NASA Astrophysics Data System (ADS)

    Soldatov, Alexey I.

    In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

  8. 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.

  9. 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.

  10. 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

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

    SciTech Connect

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

    1980-01-01

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

  12. 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

  13. 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).

  14. 76 FR 24494 - Draft Guidance for Industry and FDA Staff: Processing/Reprocessing Medical Devices in Health Care...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-02

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

  15. 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

  16. 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.

    2016-09-01

    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.

  17. 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.

  18. Waste Management Planned for the Advanced Fuel Cycle Facility

    SciTech Connect

    Soelberg

    2007-09-01

    The U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) program has been proposed to develop and employ advanced technologies to increase the proliferation resistance of spent nuclear fuels, recover and reuse nuclear fuel resources, and reduce the amount of wastes requiring permanent geological disposal. In the initial GNEP fuel cycle concept, spent nuclear fuel is to be reprocessed to separate re-useable transuranic elements and uranium from waste fission products, for fabricating new fuel for fast reactors. The separated wastes would be converted to robust waste forms for disposal. The Advanced Fuel Cycle Facility (AFCF) is proposed by DOE for developing and demonstrating spent nuclear fuel recycling technologies and systems. The AFCF will include capabilities for receiving and reprocessing spent fuel and fabricating new nuclear fuel from the reprocessed spent fuel. Reprocessing and fuel fabrication activities will generate a variety of radioactive and mixed waste streams. Some of these waste streams are unique and unprecedented. The GNEP vision challenges traditional U.S. radioactive waste policies and regulations. Product and waste streams have been identified during conceptual design. Waste treatment technologies have been proposed based on the characteristics of the waste streams and the expected requirements for the final waste forms. Results of AFCF operations will advance new technologies that will contribute to safe and economical commercial spent fuel reprocessing facilities needed to meet the GNEP vision. As conceptual design work and research and design continues, the waste management strategies for the AFCF are expected to also evolve.

  19. 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.

  20. 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)

  1. 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.

  2. 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.

  3. 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

  4. Licensing for tritium production in a commercial light water reactor: A utility view

    SciTech Connect

    Chardos, J.S.; Sorensen, G.C.; Erickson, L.W.

    2000-07-01

    In a December 1995 Record of Decision for the Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling, the US Department of Energy (DOE) decided to pursue a dual-track approach to determine the preferred option for future production of tritium for the nuclear weapons stockpile. The two options to be pursued were (a) the Accelerator Production of Tritium and (b) the use of commercial light water reactors (CLWRs). DOE committed to select one of these two options as the primary means of tritium production by the end of 1998. The other option would continue to be pursued as a backup to the primary option. The Tennessee Valley Authority (TVA) became involved in the tritium program in early 1996, in response to an inquiry from Pacific Northwest National Laboratory (PNNL) for an expression of interest by utilities operating nuclear power plants (NPPs). In June 1996, TVA was one of two utilities to respond to a request for proposals to irradiate lead test assemblies (LTAs) containing tritium-producing burnable absorber rods (TPBARs). TVA proposed that the LTAs be placed in Watts Bar NPP Unit 1 (WBN). TVA participated with DOE (the Defense Programs Office of CLWR Tritium Production), PNNL, and Westinghouse Electric Company (Westinghouse) in the design process to ensure that the TPBARs would be compatible with safe operation of WBN. Following US Nuclear Regulatory Commission (NRC) issuance of a Safety Evaluation Report (SER) (NUREG-1607), TVA submitted a license amendment request to the NRC for approval to place four LTAs, containing eight TPBARs each, in WBN during the September 1997 refueling outage. In December 1998, DOE announced the selection of the CLWR program as the primary option for tritium production and identified the TVA WBN and Sequoyah NPP (SQN) Units 1 and 2 (SQN-1 and SQN-2, respectively) reactors as the preferred locations to perform tritium production. TVA will prepare license amendment requests for the three plants (WBN, SQN-1

  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. 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

  7. 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

  8. 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

  9. 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.

  10. 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

  11. Analysis of the options - rationale for servomanipulator maintenance in future reprocessing plants

    SciTech Connect

    Herndon, J.N.; Hamel, W.R.

    1984-04-01

    The currently available maintenance systems which can be applied, in various combinations, to large-volume in-cell maintenance operations are: (1) contact, (2) overhead cranes, (3) power manipulators, (4) mechanical master/slave manipulators, and (5) servomanipulators. The requirements for reprocessing plant maintenance are reviewed, and the capabilities of remote maintenance systems are described. A basic trade-off analysis of these remote maintenance systems considering reprocessing plant requirements is given. Justification is given for selecting the overhead crane/servomanipulator-based maintenance concept as the option most desirable for future large reprocessing plant maintenance. 23 references, 6 tables.

  12. 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.

  13. 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.

  14. 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...

  15. 10 CFR 51.51 - Uranium fuel cycle environmental data-Table S-3.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 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...

  16. 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.

  17. 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

  18. On the reprocessing of gamma-rays produced by jets

    NASA Astrophysics Data System (ADS)

    Orellana, M.; Pellizza, L. J.; Romero, G. E.

    2011-02-01

    Systems of two very different sizescales are known to produce very high-energy (VHE) radiation in their jets: AGNs and microquasars. The produced VHE photons (Eγ ~ 1 TeV) can be absorbed by the intense environmental soft photon fields, coming from the companion star (in high mass binaries) or from the accreting material (disk+corona in AGNs), as these are the dominant sources at energies around ~(mec2)2/Eγ. Energetic pairs are created by the photon-photon annihilation, and, depending on how efficient are the competing cooling channels, the absorption can lead to a reprocessing by Inverse Compton pair-cascade development. A self-consistent modeling of these systems as gamma-ray sources should then include, along with the emission and absorption processes, a thorough treatment of the pair cascades. We discuss here on this issue, focusing on our (preliminary) results of numerical simulations devoted to a study case similar to the high-mass microquasar candidate LS 5039.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  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. 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.

  5. Consolidated Fuel Reprocessing Program. Progress report, October 1-December 31, 1984

    SciTech Connect

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

    1985-02-01

    Improved processes and components for the Breeder Reprocessing Engineering Test (BRET) were identifed and developed as well as the design, procurement and development of prototypic equipment. The integrated testing of process equipment and flowsheets prototypical of a pilot-scale full reprocessing plant, and also for testing prototypical remote features of specific complex componets in the system are provided. Information to guide the long-range activities of the Consolidated Fuel Reprocessing Program (CFRP), a focal point for foreign exchange activities, and support in specialized technical areas are described. Research and development activities in HTGR fuel treatment technology are being conducted. Head-end process and laboratory-scale development efforts, as well as studies specific to HTGR fuel, are reported. The development of off-gas treatment processes has generic application to fuel reprocessing; progress in this work is also reported.

  6. Current issues result in a paradigm shift in reprocessing medical and surgical instruments.

    PubMed

    Alfa, Michelle J

    2016-05-01

    The objective of this report is to review the available scientific data on reprocessing of medical and surgical instruments and discuss the current issues related to cleaning and disinfection of flexible endoscopes and intracavitary ultrasound probes.

  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. Thermochemical cycles

    NASA Technical Reports Server (NTRS)

    Funk, J. E.; Soliman, M. A.; Carty, R. H.; Conger, W. L.; Cox, K. E.; Lawson, D.

    1975-01-01

    The thermochemical production of hydrogen is described along with the HYDRGN computer program which attempts to rate the various thermochemical cycles. Specific thermochemical cycles discussed include: iron sulfur cycle; iron chloride cycle; and hybrid sulfuric acid cycle.

  9. A Multi-techniques Approach to assess reprocessing of Single-use electrosurgical pencils.

    PubMed

    Tessarolo, Francesco; Torres, Sebastian; Ballesteros, Luis; Rigoni, Marta; Piccoli, Federico; Caola, Iole; Montoya, Yesid; Nollo, Giandomenico

    2015-01-01

    Reprocessing and reuse of single-use electrosurgical pencils (EPs) is a diffused practice in countries with limited access to healthcare subvention system and in developing countries. However, safety and efficacy issues are associated to this practice, requiring specific methods for checking the reprocessed device before clinical re-use. This study aimed at defining a set of testing methods for assessing thermal and surface characteristics of reprocessed single-use EPs and evaluating the suitability of these techniques for revealing modifications between brand new and reprocessed single-use EPs. We reported a multi-technique approach based on optical and electron microscopy, X-rays spectroscopy and thermal analysis. The assessment of a total of 30 variables of interest on both brand new and reprocessed devices, allowed to identify the most informative ones. Seven of the evaluated variables were found to differentiate the reprocessed device from the new ones in a significant way. The presented methods deserve potential for tracking modifications during the device lifecycle. PMID:26737873

  10. 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)

  11. Fuel cycle cost uncertainty from nuclear fuel cycle comparison

    SciTech Connect

    Li, J.; McNelis, D.; Yim, M.S.

    2013-07-01

    This paper examined the uncertainty in fuel cycle cost (FCC) calculation by considering both model and parameter uncertainty. Four different fuel cycle options were compared in the analysis including the once-through cycle (OT), the DUPIC cycle, the MOX cycle and a closed fuel cycle with fast reactors (FR). The model uncertainty was addressed by using three different FCC modeling approaches with and without the time value of money consideration. The relative ratios of FCC in comparison to OT did not change much by using different modeling approaches. This observation was consistent with the results of the sensitivity study for the discount rate. Two different sets of data with uncertainty range of unit costs were used to address the parameter uncertainty of the FCC calculation. The sensitivity study showed that the dominating contributor to the total variance of FCC is the uranium price. In general, the FCC of OT was found to be the lowest followed by FR, MOX, and DUPIC. But depending on the uranium price, the FR cycle was found to have lower FCC over OT. The reprocessing cost was also found to have a major impact on FCC.

  12. Closing nuclear fuel cycle with fast reactors: problems and prospects

    SciTech Connect

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V.

    2013-07-01

    The closed nuclear fuel cycle (CNFC) with fast reactors (FR) is the most promising way of nuclear energetics development because it prevents spent nuclear fuel (SNF) accumulation and minimizes radwaste volume due to minor actinides (MA) transmutation. CNFC with FR requires the elaboration of safety, environmentally acceptable and economically effective methods of treatment of SNF with high burn-up and low cooling time. The up-to-date industrially implemented SNF reprocessing technologies based on hydrometallurgical methods are not suitable for the reprocessing of SNF with high burn-up and low cooling time. The alternative dry methods (such as electrorefining in molten salts or fluoride technologies) applicable for such SNF reprocessing have not found implementation at industrial scale. So the cost of SNF reprocessing by means of dry technologies can hardly be estimated. Another problem of dry technologies is the recovery of fissionable materials pure enough for dense fuel fabrication. A combination of technical solutions performed with hydrometallurgical and dry technologies (pyro-technology) is proposed and it appears to be a promising way for the elaboration of economically, ecologically and socially accepted technology of FR SNF management. This paper deals with discussion of main principle of dry and aqueous operations combination that probably would provide safety and economic efficiency of the FR SNF reprocessing. (authors)

  13. Advanced Process Monitoring Techniques for Safeguarding Reprocessing Facilities

    SciTech Connect

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

    2010-11-30

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

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. Incorporation of Passive Safety Systems in the Generation-IV Multi-Application Small Light Water Reactor (MASLWR)

    SciTech Connect

    Modro, S. Michael; Ougouag, Abderrafi M.; Fisher, James; Weaver, Kevan

    2002-07-01

    The Idaho National Engineering and Environmental Laboratory (INEEL), Nexant Inc. and the Oregon State University (OSU) developed an innovative Multi-Application Small Light Water Reactor (MASLWR) concept. The MASLWR is a small, modular, safe, and economic natural circulation light water reactor developed with the primary goal of producing electric power, but with the flexibility to be used for water desalination or district heating with deployment in a variety of locations. The MASLWR was developed, by design, to be a safe and economic reactor concept that can be deployed in the near term by utilizing current experience and capabilities of the industry. The key features of the MASLWR concept are the extreme simplicity of the design and its passive safety systems. This paper provides an overview of safety analyses performed for the MASLWR concept and explores potential for the increase in passive safety via the implementation of new features. The results of these safety studies demonstrate that the reactor core will be provided with a stable cooling source adequate to remove decay heat without significant cladding heatup under all credible scenarios. The response of the system to accident conditions is a controlled depressurization, whereby most of the primary system blowdown occurs via the submerged ADS blowdown pathway. (authors)

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. 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.”

  7. 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.

  8. Material specification for alloy X-750 for use in LWR (light water reactor) internal components (Revision 1)

    SciTech Connect

    Stein, A.A.; Gennaro, M.S. )

    1990-11-01

    This report presents a review of the stress corrosion cracking behavior of alloy X-750, the alloy chemistry, metallurgy, and mechanical properties as well as an interim material specification developed from these data. The material specification addresses requirements for heat treating, metallurgy, mechanical tests, microstructure, and quality assurance to optimize the stress corrosion resistance of alloy X-750 in light water reactor environments. The specification can be used to effectively identify and screen stress corrosion cracking resistant material. A justification for these specification requirements also is included. This information will enable the utility industry to purchase alloy X-750 in a manner which optimizes its corrosion performance with no degradation of its other properties. 26 refs.

  9. 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.

  10. 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. PMID:27225394

  11. A reprocessing model for the ultraviolet and optical light from 4U 1820-30

    NASA Technical Reports Server (NTRS)

    Arons, Jonathan; King, Ivan R.

    1993-01-01

    We show that the recently discovered optical and ultraviolet light from the X-ray burst source 4U 1820-30 in the globular cluster NGC 6624 is due to reprocessing of the X-rays in the outer regions of an optically thick, geometrically thin accretion disk. We suggest that observation of orbital modulation of the reprocessed light, due to the variable contribution made as the heated face of the companion turns toward and away from the observer, would provide constraints on the inclination of the binary orbit, and we suggest that detection in the reprocessed flux of the 'red noise' already observed in the X-rays would provide useful constraints on the geometry and physics of the accretion disk.

  12. 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.

  13. Development of Nitrogen Oxide Closed System in the Future Reprocessing Process

    SciTech Connect

    Takaoku, Y.; Hattori, I.; Watanabe, T.; Moriya, N.; Sumida, Y.; Araya, S.; Homma, S.; Suzuki, Y.; Akai, Y.

    2007-07-01

    An aqueous reprocessing for spent fuels generates much wastes mainly including sodium nitrate as secondary waste, which has some kinds of difficulties in disposal. A process with salt-free reagent and complete recycle of nitric acid would resolve the problem, but development for such process is not easy. We propose the treatment system of sodium nitrate waste, which are termed 'Nitrogen Oxide Closed System' (NCS) as mentioned below. The system decomposes nitrate ion, and enables reuse of sodium in sodium nitrate with no generation of sodium nitrate waste. Accordingly, the NCS system allows the use of sodium salt reagents, and generation of excess acid in a reprocessing process. (authors)

  14. 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.

  15. 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.

  16. 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.

  17. 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.

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

    SciTech Connect

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

    1982-07-01

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

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

    SciTech Connect

    Pokhitonov, Y.

    2007-07-01

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

  20. 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)

  1. 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.

  2. Reprocessing 'single use' devices--why is there so much debate?

    PubMed

    Stoermer, W B

    1999-01-01

    Reducing costs without compromising standards of patient care challenges health care professionals at all levels. Reprocessing of selected disposable devices can result in major savings when issues are properly addressed. No published standards differentiate quality, function, or length of useful life between reusable and single-use devices.

  3. 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.

  4. 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…

  5. 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…

  6. 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.

  7. 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. PMID:26897699

  8. 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.

  9. 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)

  10. 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

  11. 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.

  12. 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.; 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 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

  13. 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.

  14. Advanced light water reactor requirements document: Chapter 3, Reactor coolant system and reactor non-safety auxiliary systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Coolant System and the Reactor Non-safety Auxiliary Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the reactor coolant system and reactor non-safety auxiliary systems for Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Non-safety auxiliaries include systems which are required for normal operation of the plant but are not required to operate for accident mitigation or to bring the plant to a safe shutdown condition. For PWRs, the reactor coolant system, steam generator system, chemical and volume control system and boron recycle system are included. For BWRs, the reactor coolant system and reactor water cleanup system are included. The chapter also includes requirements for the above systems which are common to BWRs and PWRs and requirements for process sampling for BWRs and PWRs.

  15. 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.

  16. 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.

  17. 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.

  18. Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining

    SciTech Connect

    S. D. Herrmann; S. X. Li

    2010-09-01

    A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl – 1 wt% Li2O at 650 °C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 °C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

  19. 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.

  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. 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

  2. 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.

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

    PubMed

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

    2014-12-01

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

  4. 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.

  5. VISION - Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics

    SciTech Connect

    Steven J. Piet; A. M. Yacout; J. J. Jacobson; C. Laws; G. E. Matthern; D. E. Shropshire

    2006-02-01

    The U.S. DOE Advanced Fuel Cycle Initiative’s (AFCI) fundamental objective is to provide technology options that - if implemented - would enable long-term growth of nuclear power while improving sustainability and energy security. The AFCI organization structure consists of four areas; Systems Analysis, Fuels, Separations and Transmutations. The Systems Analysis Working Group is tasked with bridging the program technical areas and providing the models, tools, and analyses required to assess the feasibility of design and deployment options and inform key decision makers. An integral part of the Systems Analysis tool set is the development of a system level model that can be used to examine the implications of the different mixes of reactors, implications of fuel reprocessing, impact of deployment technologies, as well as potential "exit" or "off ramp" approaches to phase out technologies, waste management issues and long-term repository needs. The Verifiable Fuel Cycle Simulation Model (VISION) is a computer-based simulation model that allows performing dynamic simulations of fuel cycles to quantify infrastructure requirements and identify key trade-offs between alternatives. It is based on the current AFCI system analysis tool "DYMOND-US" functionalities in addition to economics, isotopic decay, and other new functionalities. VISION is intended to serve as a broad systems analysis and study tool applicable to work conducted as part of the AFCI and Generation IV reactor development studies.

  6. 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.

  7. 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.

  8. 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.

  9. Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

    SciTech Connect

    Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E.

    1995-07-01

    The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab.

  10. 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.

  11. Menstrual Cycle

    MedlinePlus

    ... Pregnancy This information in Spanish ( en español ) 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 ...

  12. Biogeochemical Cycling

    NASA Technical Reports Server (NTRS)

    Bebout, Brad; Fonda, Mark (Technical Monitor)

    2002-01-01

    This lecture will introduce the concept of biogeochemical cycling. The roles of microbes in the cycling of nutrients, production and consumption of trace gases, and mineralization will be briefly introduced.

  13. 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.

  14. Designing and Operating for Safeguards: Lessons Learned From the Rokkasho Reprocessing Plant (RRP)

    SciTech Connect

    Johnson, Shirley J.; Ehinger, Michael

    2010-08-07

    This paper will address the lessons learned during the implementation of International Atomic Energy Agency (IAEA) safeguards at the Rokkasho Reprocessing Plant (RRP) which are relevant to the issue of ‘safeguards by design’. However, those lessons are a result of a cumulative history of international safeguards experiences starting with the West Valley reprocessing plant in 1969, continuing with the Barnwell plant, and then with the implementation of international safeguards at WAK in Germany and TRP in Japan. The design and implementation of safeguards at RRP in Japan is the latest and most challenging that the IAEA has faced. This paper will discuss the work leading up to the development of a safeguards approach, the design and operating features that were introduced to improve or aid in implementing the safeguards approach, and the resulting recommendations for future facilities. It will provide an overview of how ‘safeguardability’ was introduced into RRP.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

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

    SciTech Connect

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

    2009-11-10

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

  20. Quality Assessment and Collection V1.1 Reprocessing of the Suomi NPP VIIRS LAND Records

    NASA Astrophysics Data System (ADS)

    Devadiga, S.; Davidson, C. C.; Sarkar, S.; Ye, G.; Hattori, M.; Praderas, C.; Kalb, V.; Nguyen, A.; Hamilton, C.; Kuyper, J.; Roman, M. O.; Mauoka, E.

    2014-12-01

    The Land Product Evaluation and Algorithm Testing Element (PEATE) is an important element of the Suomi National Polar-orbiting Partnership (SNPP) at NASA's Goddard Space Flight Center. The primary goals of NASA's Land PEATE are to assess the quality of the Visible Infrared Imaging Radiometer Suite (VIIRS) Land operational products made by the Interface Data Processing System (IDPS), and to recommend improvements to the operational algorithms to meet NASA's Land science needs. The Land PEATE uses a version of the MODIS Adaptive Processing System (MODAPS), NPPDAPS, which has been modified to run the IDPS operational algorithms, as well as software provided by the NASA SNPP Land Science Team. Since the early pre-launch period (c. 2009) the Land PEATE has used the MODIS Land Data Operational Product Evaluation (LDOPE) team for evaluation of the data records generated by NPPDAPS.In June 2014, Land PEATE completed Collection V1.1 reprocessing of the SNPP VIIRS Land products from the beginning (Jan 19, 2012) of the SNPP mission to current day using the best of the IDPS operational and NASA Land science team provided algorithms. The processing used the refined LUTs provided by the NASA VIIRS Calibration Support Team (VCST) for the L1B Sensor Data Records (SDR), including LUTs for calibration and stray light correction of the VIIRS Day/Night Band. In addition to generating the operational SDRs, Intermediate Products (IPs), and Environmental Data Records (EDRs) this reprocessing also produced Diagnostics Data Records, MODIS heritage L3 gridded products using the VIIRS observations. This paper describes approaches used to assess the quality of the products from operational processing and reprocessing of VIIRS records at Land PEATE. The paper also presents results from inter-comparison of records from this reprocessing with the MODIS heritage products. Our analysis verified that MODIS quality data records can be produced using the VIIRS observations, however with additional

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

    SciTech Connect

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

    1984-08-01

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

  2. Reprocessing MSU/AMSU/SSU long-term satellite instruments for climate monitoring and reanalyses

    NASA Astrophysics Data System (ADS)

    Zou, C.

    2009-12-01

    Climate monitoring requires generation of long-term satellite data products using satellite retrieval algorithms. Alternatively, satellite radiance data can be assimilated into numerical climate models to generate comprehensive reanalysis data products for climate monitoring and research. However, calibration and data consistency have been a major issue in producing reliable satellite and reanalysis climate products. Many long-term satellite climate products and earlier-generation reanalyses suffer from spurious climate jumps and variability induced by satellite transition and calibration-related instrument changes. Future satellite and reanalysis climate products require satellite data to be well calibrated so that instrument-related time-varying intersatellite biases are removed before they are used for climate analysis and data assimilation. Instrument characteristics such as frequency response functions are also required to be well described for accurately simulating radiances by radiative transfer models in reanalyses. For these purposes, NOAA/NESDIS is reprocessing and recalibrating radiance data from the NOAA and DMSP polar-orbiting satellite series. The instruments being investigated include MSU/AMSU/ATMS, SSU, HIRS, AVHRR, and SSM/I, etc. Recently, the NESDIS MSU/AMSU/SSU reprocessing effort has been funded by the NOAA Scientific Data Stewardship Program. In this presentation, we will review the current status on the NESDIS recalibration/reprocessing of the 30-year MSU/AMSU radiance data using simultaneous nadir overpass (SNO) method. We present results of the merged MSU/AMSU long-term atmospheric temperature climate data records derived from the recalibration. We also present preliminary results on bias correction and reprocessing of the SSU long-term time series using the SNO method.

  3. 75 FR 45167 - Notice of Public Workshop on a Potential Rulemaking for Spent Nuclear Fuel Reprocessing Facilities

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-02

    ... Advisory Committee on Nuclear Waste and Materials (ACNW&M) titled ``Background, Status and Issues Related... Register on August 31, 1984 (49 FR 34658) discusses waste from reprocessing facilities in the first...

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

    SciTech Connect

    Maxwell, S.L. III.

    1991-01-01

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

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

    SciTech Connect

    Maxwell, S.L. III

    1991-12-31

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

  6. Comparing SeaWiFS Reprocessing Versions (R3 vs. R4)

    NASA Technical Reports Server (NTRS)

    Casey, Nancy W.; Gregg, Watson W.

    2003-01-01

    Satellite observations of global ocean chlorophyll from SeaWiFS were recently reprocessed to incorporate calibration and algorithm improvements. Here, comparisons are made between the newly reprocessed SeaWiFS Level-3 chlorophyll product and the previous version using in situ measurements. The results show that the newly reprocessed SeaWiFS data matches up better with the surface measurements than the previous version did. Globally, the slope of the match-ups improves to 0.85 from 0.78 in log-log scale. A significant trend that contributed to this improvement was the overall decrease in SeaWiFS chlorophyll levels less than 1.0 mg m(exp -3). Regional analyses reveal that the matchups improve in every oceanic basin, except the Antarctic. However, SeaWiFS continues to exhibit poor correspondence with in situ data in the North Atlantic where the match-ups have a slope of 0.54. Also, an examination of monthly images for May 1999 revealed that the number and magnitude of high-value chlorophyll pixels had increased in the high-latitude open ocean of the South Pacific.

  7. Assessment of the orbits from the 1st IGS reprocessing campaign (Invited)

    NASA Astrophysics Data System (ADS)

    Griffiths, J.; Gendt, G.; Nischan, T.; Ray, J.

    2009-12-01

    All eight Finals Analysis Centers (ACs) of the International GNSS Service (IGS), plus three allied groups, for a total of eleven contributing ACs, are using the latest models and methodology to reanalyze the history of Global Positioning System (GPS) data collected since 1994 by a global network of tracking stations. The solutions from the contributing ACs will be used to form a fully consistent, long-term set of official IGS products, specifically Earth orientation parameters, clocks, orbits and station coordinates. This is the first such reprocessing campaign by the IGS, which is expected to be complete by mid-November, 2009. A second reprocessing campaign will follow as further updates and improvements are made to the models and methodology. This paper, then, focuses on assessing the reliability and quality of the combined orbits from this first campaign, and making recommendations to be considered in preparation for the second campaign. In our preliminary assessment of the reliability of the combined orbits, we have examined the reprocessed AC orbit estimates from the 1025 days of results until the end of 2007. Each daily AC satellite ephemeris for each pair of consecutive days has been fitted to the extended CODE orbit model, extrapolated to the mid-point epoch between the days, and the geocentric satellite position differences computed to give time series of orbit repeatabilities. Occasional data gaps have been filled by linear interpolation, FFT power spectra computed, and the spectra stacked over the full GPS constellation and lightly smoothed. This preliminary analysis of the midnight discontinuities reveals considerable diversity among AC orbits. Several show broad semi-annual and fortnightly spectral peaks, as well as even harmonics of the GPS draconitic year (1.040 cpy) with varying amplitudes. High-frequency white noise floors can be detected in most AC orbit spectra, with an average sigma of 14 mm and larger. The statistics from preliminary orbit

  8. Safeguards operations in the integral fast reactor fuel cycle

    SciTech Connect

    Goff, K.M.; Benedict, R.W.; Brumbach, S.B.; Dickerman, C.E.; Tompot, R.W.

    1994-08-01

    Argonne National Laboratory is currently demonstrating the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The safeguards aspects of the fuel cycle demonstration must be approved by the United States Department of Energy, but a further goal of the program is to develop a safeguards system that could gain acceptance from the Nuclear Regulatory Commission and International Atomic Energy Agency. This fuel cycle is described with emphasis on aspects that differ from aqueous reprocessing and on its improved safeguardability due to decreased attractiveness and diversion potential of all process streams, including the fuel product.

  9. FES cycling.

    PubMed

    Newham, D J; Donaldson, N de N

    2007-01-01

    Spinal cord injury (SCI) leads to a partial or complete disruption of motor, sensory, and autonomic nerve pathways below the level of the lesion. In paraplegic patients, functional electrical stimulation (FES) was originally widely considered as a means to restore walking function but this was proved technically very difficult because of the numerous degrees of freedom involved in walking. FES cycling was developed for people with SCI and has the advantages that cycling can be maintained for reasonably long periods in trained muscles and the risk of falls is low. In the article, we review research findings relevant to the successful application of FES cycling including the effects on muscle size, strength and function, and the cardiovascular and bone changes. We also describe important practical considerations in FES cycling regarding the application of surface electrodes, training and setting up the stimulator limitations, implanted stimulators and FES cycling including FES cycling in groups and other FES exercises such as FES rowing.

  10. 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

  11. 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.

  12. 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.).

  13. Toward mechanistic understanding of nuclear reprocessing chemistries by quantifying lanthanide solvent extraction kinetics via microfluidics with constant interfacial area and rapid mixing.

    PubMed

    Nichols, Kevin P; Pompano, Rebecca R; Li, Liang; Gelis, Artem V; Ismagilov, Rustem F

    2011-10-01

    The closing of the nuclear fuel cycle is an unsolved problem of great importance. Separating radionuclides produced in a nuclear reactor is useful both for the storage of nuclear waste and for recycling of nuclear fuel. These separations can be performed by designing appropriate chelation chemistries and liquid-liquid extraction schemes, such as in the TALSPEAK process (Trivalent Actinide-Lanthanide Separation by Phosphorus reagent Extraction from Aqueous Komplexes). However, there are no approved methods for the industrial scale reprocessing of civilian nuclear fuel in the United States. One bottleneck in the design of next-generation solvent extraction-based nuclear fuel reprocessing schemes is a lack of interfacial mass transfer rate constants obtained under well-controlled conditions for lanthanide and actinide ligand complexes; such rate constants are a prerequisite for mechanistic understanding of the extraction chemistries involved and are of great assistance in the design of new chemistries. In addition, rate constants obtained under conditions of known interfacial area have immediate, practical utility in models required for the scaling-up of laboratory-scale demonstrations to industrial-scale solutions. Existing experimental techniques for determining these rate constants suffer from two key drawbacks: either slow mixing or unknown interfacial area. The volume of waste produced by traditional methods is an additional, practical concern in experiments involving radioactive elements, both from disposal cost and experimenter safety standpoints. In this paper, we test a plug-based microfluidic system that uses flowing plugs (droplets) in microfluidic channels to determine absolute interfacial mass transfer rate constants under conditions of both rapid mixing and controlled interfacial area. We utilize this system to determine, for the first time, the rate constants for interfacial transfer of all lanthanides, minus promethium, plus yttrium, under TALSPEAK

  14. Toward mechanistic understanding of nuclear reprocessing chemistries by quantifying lanthanide solvent extraction kinetics via microfluidics with constant interfacial area and rapid mixing.

    PubMed

    Nichols, Kevin P; Pompano, Rebecca R; Li, Liang; Gelis, Artem V; Ismagilov, Rustem F

    2011-10-01

    The closing of the nuclear fuel cycle is an unsolved problem of great importance. Separating radionuclides produced in a nuclear reactor is useful both for the storage of nuclear waste and for recycling of nuclear fuel. These separations can be performed by designing appropriate chelation chemistries and liquid-liquid extraction schemes, such as in the TALSPEAK process (Trivalent Actinide-Lanthanide Separation by Phosphorus reagent Extraction from Aqueous Komplexes). However, there are no approved methods for the industrial scale reprocessing of civilian nuclear fuel in the United States. One bottleneck in the design of next-generation solvent extraction-based nuclear fuel reprocessing schemes is a lack of interfacial mass transfer rate constants obtained under well-controlled conditions for lanthanide and actinide ligand complexes; such rate constants are a prerequisite for mechanistic understanding of the extraction chemistries involved and are of great assistance in the design of new chemistries. In addition, rate constants obtained under conditions of known interfacial area have immediate, practical utility in models required for the scaling-up of laboratory-scale demonstrations to industrial-scale solutions. Existing experimental techniques for determining these rate constants suffer from two key drawbacks: either slow mixing or unknown interfacial area. The volume of waste produced by traditional methods is an additional, practical concern in experiments involving radioactive elements, both from disposal cost and experimenter safety standpoints. In this paper, we test a plug-based microfluidic system that uses flowing plugs (droplets) in microfluidic channels to determine absolute interfacial mass transfer rate constants under conditions of both rapid mixing and controlled interfacial area. We utilize this system to determine, for the first time, the rate constants for interfacial transfer of all lanthanides, minus promethium, plus yttrium, under TALSPEAK

  15. 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.

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

    SciTech Connect

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

    2009-10-06

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

  17. 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.

  18. 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.

  19. 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.

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

    DOEpatents

    Forsberg, Charles W.

    1992-01-01

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

  1. 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).

  2. Report on the NGS3 Working Group on Safeguards by Design For Aqueous Reprocessing Plants

    SciTech Connect

    Johnson, Shirley J.; Ehinger, Michael; Schanfein, Mark

    2011-02-01

    The objective of the Working Group on SBD for Aqueous Reprocessing Facilities was to provide recommendations, for facility operators and designers, which would aid in the coordination and integration of nuclear material accountancy and the safeguards requirements of all concerned parties - operators, state/regional authorities, and the IAEA. The recommendations, which are to be provided to the IAEA, are intended to assist in optimizing facility design and operating parameters to ensure the safeguardability of the facility while minimizing impact on the operations. The one day Working Group session addressed a wide range of design and operating topics.

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. Cycle Analysis

    SciTech Connect

    Wright, Steven A.

    2012-03-20

    1. The Cycle Analysis code is an Microsoft Excel code that performs many different types of thermodynamic cycle analysis for power producing systems. The code will calculate the temperature and pressure and all other thermodynamic properties at the inlet and outlet of each component. The code also calculates the power that is produced, the efficiency, and the heat transported in the heater, gas chiller and recuperators. The code provides a schematic of the loop and provides the temperature and pressure at each location in the loop. The code also provides a T-S (temperature-entropy) diagram of the loop and often it provides an pressure enthalpy plot as well. 2. This version of the code concentrates on supercritical CO2 power cycles, but by simply changing the name of the working fluid many other types of fluids can be analyzed. The Cycle Analysis code provided here contains 18 different types of power cycles. Each cycle is contained in one worksheet or tab that the user can select. The user can change the yellow highlighted regions to perform different thermodynamic cycle analysis.

  8. Cycle Analysis

    2012-03-20

    1. The Cycle Analysis code is an Microsoft Excel code that performs many different types of thermodynamic cycle analysis for power producing systems. The code will calculate the temperature and pressure and all other thermodynamic properties at the inlet and outlet of each component. The code also calculates the power that is produced, the efficiency, and the heat transported in the heater, gas chiller and recuperators. The code provides a schematic of the loop andmore » provides the temperature and pressure at each location in the loop. The code also provides a T-S (temperature-entropy) diagram of the loop and often it provides an pressure enthalpy plot as well. 2. This version of the code concentrates on supercritical CO2 power cycles, but by simply changing the name of the working fluid many other types of fluids can be analyzed. The Cycle Analysis code provided here contains 18 different types of power cycles. Each cycle is contained in one worksheet or tab that the user can select. The user can change the yellow highlighted regions to perform different thermodynamic cycle analysis.« less

  9. Understanding the Atomic-Level Chemistry and Structure of Oxide Deposits on Fuel Rods in Light Water Nuclear Reactors Using First Principles Methods

    NASA Astrophysics Data System (ADS)

    Rak, Zs.; O'Brien, C. J.; Brenner, D. W.; Andersson, D. A.; Stanek, C. R.

    2016-09-01

    The results of recent studies are discussed in which first principles calculations at the atomic level have been used to expand the thermodynamic database for science-based predictive modeling of the chemistry, composition and structure of unwanted oxides that deposit on the fuel rods in pressurized light water nuclear reactors. Issues discussed include the origin of the particles that make up deposits, the structure and properties of the deposits, and the forms by which boron uptake into the deposits can occur. These first principles approaches have implications for other research areas, such as hydrothermal synthesis and the stability and corrosion resistance of other materials under other extreme conditions.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. A spectrophotometric study of cerium IV and chromium VI species in nuclear fuel reprocessing process streams

    NASA Astrophysics Data System (ADS)

    Nickson, I. D.; Boxall, C.; Jackson, A.; Whillock, G. O. H.

    2010-03-01

    Nuclear fuel reprocessing schemes such as PUREX and UREX utilise HNO3 media. An understanding of the corrosion of process engineering materials such as stainless steel in such media is a major concern for the nuclear industry. Two key species are cerium and chromium which, as Ce(IV), Cr(VI), may act as corrosion accelerants. An on-line analytical technique for these quantities would be useful for determining the relationship between corrosion rate and [Ce(IV)] and [Cr(VI)]. Consequently, a strategy for simultaneous quantification of Ce(IV), Cr(VI) and Cr(III) in the presence of other ions found in average burn-up Magnox / PWR fuel reprocessing stream (Fe, Mg, Nd, Al) is being developed. This involves simultaneous UV-vis absorbance measurement at 620, 540, 450 nm, wavelengths where Ce and Cr absorb but other ions do not. Mixed solutions of Cr(VI) and Ce(IV) are found to present higher absorbance values at 540 nm than those predicted from absorbances recorded from single component solutions of those ions. This is attributed to the formation of a 3:1 Cr(VI)-Ce(IV) complex and we report on the complexation and UV-visible spectrophotometric characteristics of this species. To the best of our knowledge this is the first experimental study of this complex in aqueous nitric acid solution systems.

  15. 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.

  16. 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.

  17. Vapor Compression Cycle Design Program (CYCLE_D)

    National Institute of Standards and Technology Data Gateway

    SRD 49 NIST Vapor Compression Cycle Design Program (CYCLE_D) (PC database for purchase)   The CYCLE_D database package simulates the vapor compression refrigeration cycles. It is fully compatible with REFPROP 9.0 and covers the 62 single-compound refrigerants . Fluids can be used in mixtures comprising up to five components.

  18. The concept of the use of recycled uranium for increasing the degree of security of export deliveries of fuel for light-water reactors

    NASA Astrophysics Data System (ADS)

    Alekseev, P. N.; Ivanov, E. A.; Nevinitsa, V. A.; Ponomarev-Stepnoi, N. N.; Rumyantsev, A. N.; Shmelev, V. M.; Borisevich, V. D.; Smirnov, A. Yu.; Sulaberidze, G. A.

    2010-12-01

    The present paper deals with investigation of the possibilities for reducing the risk of proliferation of fissionable materials by means of increasing the degree of protection of fresh fuel intended for light-water reactors against unsanctioned use in the case of withdrawal of a recipient country of deliveries from IAEA safeguards. It is shown that the use of recycled uranium for manufacturing export nuclear fuel makes transfer of nuclear material removed from the fuel assemblies for weapons purposes difficult because of the presence of isotope 232U, whose content increases when one attempts to enrich uranium extracted from fresh fuel. In combination with restricted access to technologies for isotope separation by means of establishing international centers for uranium enrichment, this technical measure can significantly reduce the risk of proliferation associated with export deliveries of fuel made of low-enriched uranium. The assessment of a maximum level of contamination of nuclear material being transferred by isotope 232U for the given isotope composition of the initial fuel is obtained. The concept of further investigations of the degree of security of export deliveries of fuel assemblies with recycled uranium intended for light-water reactors is suggested.

  19. Concept for a small, colocated fuel cycle facility for oxide breeder fuels

    SciTech Connect

    Burch, W.D.; Stradley, J.G.; Lerch, R.E.

    1987-01-01

    As part of a United States Department of Energy (USDOE) program to examine innovative liquid-metal reactor (LMR) system designs over the past three years, the Oak Ridge National Laboratory (ORNL) and the Westinghouse Hanford Company (WHC) collaborated on studies of mixed oxide fuel cycle options. A principal effort was an advanced concept for a small integrated fuel cycle colocated with a 1300-MW(e) reactor station. The study provided a scoping design and a basis on which to proceed with implementation of such a facility if future plans so dictate. The facility integrated reprocessing, waste management, and refabrication functions in a single facility of nominal 35-t/year capacity utilizing the latest technology developed in fabrication programs at WHC and in reprocessing at ORNL. The concept was based on many years of work at both sites and extensive design studies of prior years.

  20. Concept for a small, colocated fuel cycle facility for oxide breeder fuels

    SciTech Connect

    Burch, W.D.; Lerch, R.E.; Stradley, J.G.

    1987-01-01

    As part of a United States Department of Energy (USDOE) program to examine innovative liquid-metal reactor (LMR) system designs over the past three years, the Oak Ridge National Laboratory (ORNL) and the Westinghouse Hanford Company (WHC) collaborated on studies of mixed oxide fuel cycle options. A principal effort was an advanced concept for a small integrated fuel cycle colocated with a 1300-MW(e) reactor station. The study provided a scoping design, capital and operating cost estimates, and a basis on which to proceed with implementation of such a facility if future plans so dictate. The facility integrated reprocessing, waste management, and refabrication functions in a single facility of nominal 35-t/year capacity utilizing the latest technology developed in fabrication programs at WHC and in reprocessing at ORNL. The concept was based on many years of work at both sites and extensive design studies of prior years.

  1. Multiple reheat helium Brayton cycles for sodium fast reactors

    SciTech Connect

    Haihua Zhao; Per F. Peterson

    2008-07-01

    Sodium fast reactors (SFR) traditionally adopt the steam Rankine cycle for power conversion. The resulting potential for water-sodium reaction remains a continuing concern which at least partly delays the SFR technology commercialization and is a contributor to higher capital cost. Supercritical CO2 provides an alternative, but is also capable of sustaining energetic chemical reactions with sodium. Recent development on advanced inert-gas Brayton cycles could potentially solve this compatibility issue, increase thermal efficiency, and bring down the capital cost close to light water reactors. In this paper, helium Brayton cycles with multiple reheat and intercooling states are presented for SFRs with reactor outlet temperatures in the range of 510°C to 650°C. The resulting thermal efficiencies range from 39% and 47%, which is comparable with supercritical recompression CO2 cycles (SCO2 cycle). A systematic comparison between multiple reheat helium Brayton cycle and the SCO2 cycle is given, considering compatibility issues, plant site cooling temperature effect on plant efficiency, full plant cost optimization, and other important factors. The study indicates that the multiple reheat helium cycle is the preferred choice over SCO2 cycle for sodium fast reactors.

  2. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

    SciTech Connect

    Philip E. MacDonald

    2005-01-01

    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% versus 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 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 are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was

  3. 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

  4. An assessment of the NGS’ contribution to the reprocessed IGS products

    NASA Astrophysics Data System (ADS)

    Griffiths, J.; Ray, J.; Rohde, J. R.; Kass, W. G.; Dulaney, R. L.; Cline, M.; Hilla, S.; Snay, R. A.

    2009-12-01

    The National Geodetic Survey (NGS) has reanalyzed the history of Global Positioning System (GPS) data collected by a global network of tracking stations using the latest version of the PAGES software, resulting in a fully consistent set of station coordinates, satellite orbits and Earth orientation parameters (EOPs). The reprocessed solutions have been submitted for inclusion in the official reprocessed IGS combinations. For this paper, we attempt to assess the quality of our contribution to the reprocessed IGS products. For the station coordinates, we have stacked our full time series of SINEX files using the CATREF software from Institut Géographique National. The stacking consists of aligning each weekly SINEX file to the current IGS terrestrial frame (e.g., IGS05) and solving for a set of regularized coordinates and secular velocities for each site in the global network. Differences between the secular velocities from CATREF and those for IGS05 provide a quality measure of NGS’ realization of the terrestrial frame. According to these velocity differences, our weekly solutions appear to be internally consistent, and any biases or distortions introduced during the data reduction process are removable by a Helmert alignment to the datum. For the orbits, we have examined positional discontinuities between successive 24-hour SP3 files from the 1025 days of results until the end of 2007. Daily satellite ephemerides for each pair of consecutive days have been fitted to the extended CODE orbit model, extrapolated to the mid-point epoch between the days, and the geocentric satellite position differences computed to give the time series of orbit repeatabilities. Occasional data gaps have been filled by linear interpolation, power spectra computed, and the spectra stacked over the full GPS constellation and lightly smoothed. This preliminary analysis shows fortnightly spectral peaks, as well as even harmonics of the GPS draconitic year (1.040 cpy). A high

  5. Progress and experiences from the decommissioning of the Eurochemic reprocessing plant

    SciTech Connect

    Gills, R.; Lewandowski, P.; Ooms, B.; Reusen, N.; Van Laer, W.; Walthery, R.

    2007-07-01

    Belgoprocess started the industrial decommissioning of the main process building of the former EUROCHEMIC reprocessing plant in 1990, after completion of a pilot project in which two buildings were emptied and decontaminated to background levels. The remaining structures were demolished and the concrete debris was disposed of as industrial waste and green field conditions restored. The Eurochemic reprocessing plant operated from 1966 to 1974 to process fuel from power reactors and research reactors. The main building is a large concrete structure, comprising a surface area of 55,000 m{sup 2}, concrete volume 12,500 m{sup 3}, and 1,500 Mg of metal components. The building is divided into multiple cells. About 106 individual cell structures have to be dismantled, involving the removal and decontamination of equipment from each cell, the decontamination of the cell walls, ceilings and floors, the dismantling of the ventilation system. Most of the work involves hands-on operations under protective clothing tailored to each specific task. Tool automation and automatic positioning systems are successfully applied. In view of the final demolition of the main process building, the main process building is divided into three parts - each part is isolated from the others. In the middle of 2008, after the removal of the NDA-IPAN/GEA installation, the eastern part will be demolished. The paper presents a status overview of the decommissioning and decontamination activities at the main process building of the former Eurochemic reprocessing plant on the nuclear site of Dessel in Belgium. The specific BELGOPROCESS approach will be highlighted, in which the decommissioning activities are carried out on an industrial scale with special emphasis on cost minimisation, the use of technology on an industrial representative scale and the specific alpha contamination of equipment and building surfaces, requiring that the decommissioning work is done with adequate protective clothing

  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. Intensive HST, RXTE, and ASCA Monitoring of NGC 3516: Evidence against Thermal Reprocessing

    NASA Technical Reports Server (NTRS)

    Edelson, Rick; Koratkar, Anuradha; Nandra, Kirpal; Goad, Michael; Peterson, Bradley M.; Collier, Stefan; Krolik, Julian; Malkan, Matthew; Maoz, Dan; OBrien, Paul

    2000-01-01

    During 1998 April 1316, the bright, strongly variable Seyfert 1 galaxy NGC 3516 was monitored almost continuously with HST for 10.3 hr at ultraviolet wavelengths and 2.8 days at optical wavelengths, and simultaneous RXTE and ASCA monitoring covered the same period. The X-ray fluxes were strongly variable with the soft (0.5-2 keV) X-rays showing stronger variations (approx. 65% peak to peak) than the hard (2-10 keV) X-rays (approx. 50% peak to peak). The optical continuum showed much smaller but still highly significant variations: a slow approx. 2.5% rise followed by a faster approx. 3.5% decline. The short ultraviolet observation did not show significant variability. The soft and hard X-ray light curves were strongly correlated, with no evidence for a significant interband lag. Likewise, the optical continuum bands (3590 and 5510 A) were also strongly correlated, with no measurable lag, to 3(sigma) limits of approx. less than 0.15 day. However, the optical and X-ray light curves showed very different behavior, and no significant correlation or simple relationship could be found. These results appear difficult to reconcile with previous reports of correlations between X-ray and optical variations and of measurable lags within the optical band for some other Seyfert 1 galaxies. These results also present serious problems for "reprocessing" models in which the X-ray source heats a stratified accretion disk, which then reemits in the optical/ultraviolet : the synchronous variations within the optical would suggest that the emitting region is approx. less than 0.3 It-day across, while the lack of correlation between X-ray and optical variations would indicate, in the context of this model, that any reprocessing region must be approx. greater than 1 It-day in size. It may be possible to resolve this conflict by invoking anisotropic emission or special geometry, but the most natural explanation appears to be that the bulk of the optical luminosity is generated by some

  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. 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.

  10. Menu Cycles.

    ERIC Educational Resources Information Center

    Clayton, Alfred; Almony, John

    The curriculum guide for commercial foods instruction is designed to aid the teacher in communicating the importance of menu cycles in commercial food production. It also provides information about the necessary steps in getting food from the raw form to the finished product, and then to the consumer. In addition to providing information on how to…

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

    NASA Astrophysics Data System (ADS)

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

    2005-02-01

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

  12. 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)

  13. 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.

  14. Cadmium transport through molten salts in the reprocessing of spent fuel for the integral fast reactor

    SciTech Connect

    Goff, K.M.; Schneider, A. ); Battles, J.E. )

    1993-06-01

    The reprocessing of spent fuel from the Integral Fast Reactor is to be accomplished with a pyrochemical process employing molten LiCl-KCl salt covering a pool of cadmium. An examination of this system demonstrates that cadmium metal is soluble to a small extent in this salt and that it diffuses through the salt covering and vaporizes at the surface. The cadmium is soluble in the salt because of either chemical or physical solubility, both of which are dependent on the salt's surface tension. Mixing increases the vaporization rate of the cadmium by increasing its transport to the salt surface. The cadmium vapors can therefore be reduced by decreasing the mixing conditions, by choosing a salt with a higher surface tension so that the cadmium is less soluble, or by decreasing the temperature of the system, thereby lowering the vapor pressure of the cadmium.

  15. Mindfulness meditation training combined with eye movement desensitization and reprocessing in psychotherapy of an elderly patient.

    PubMed

    Sun, Tzan-Fu; Wu, Ching-Kuan; Chiu, Nien-Mu

    2004-06-01

    We present our experiences with an elderly patient with depression that was attributed to a surge of physical ailments who also had trauma-derived fear of having to undergo a tracheotomy. He refused pharmacotherapy and was offered intensive training in Mindfulness Meditation (MM) plus Eye Movement Desensitization and Reprocessing (EMDR) therapy during the 2 weeks of hospitalization. This treatment combination had not been used previously. We suggest that EMDR eliminated his fear of surgery, whereas MM relieved his depression and attendant anxiety. However, the two techniques appeared to work synergistically. Following his discharge, he continued to practice MM, which prevented the recurrence of emotional distress, and even helped to reduce its causative physical symptoms. We offer an explanation for the success of our combined treatments and discuss the potential usefulness in specific psychotherapeutic situations. We also propose a place for MM within general geriatric care, and point out the reluctance to consider the therapeutic value of meditation.

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

    SciTech Connect

    Levitskaia, Tatiana G.; Bryan, Samuel A.

    2008-05-23

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

  17. Idaho Nuclear Technology and Engineering Center (INTEC) Sodium Bearing Waste - Waste Incidental to Reprocessing Determination

    SciTech Connect

    Jacobson, Victor Levon

    2002-08-01

    U.S. Department of Energy Manual 435.1-1, Radioactive Waste Management, Section I.1.C, requires that all radioactive waste subject to Department of Energy Order 435.1 be managed as high-level radioactive waste, transuranic waste, or low-level radioactive waste. Determining the radiological classification of the sodium-bearing waste currently in the Idaho Nuclear Technology and Engineering Center Tank Farm Facility inventory is important to its proper treatment and disposition. This report presents the technical basis for making the determination that the sodium-bearing waste is waste incidental to spent fuel reprocessing and should be managed as mixed transuranic waste. This report focuses on the radiological characteristics of the sodiumbearing waste. The report does not address characterization of the nonradiological, hazardous constituents of the waste in accordance with Resource Conservation and Recovery Act requirements.

  18. 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.

  19. 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.

  20. 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.