National Library of Energy BETA

Sample records for nuclear fuel cladding

  1. Nuclear fuel elements having a composite cladding

    DOE Patents [OSTI]

    Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

    1983-09-20

    An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

  2. Double-clad nuclear fuel safety rod

    DOE Patents [OSTI]

    McCarthy, William H. (Los Altos, CA); Atcheson, Donald B. (Cupertino, CA); Vaidyanathan, Swaminathan (San Jose, CA)

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  3. Double-clad nuclear-fuel safety rod

    DOE Patents [OSTI]

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  4. Welding fixture for nuclear fuel pin cladding assemblies

    DOE Patents [OSTI]

    Oakley, David J. (Richland, WA); Feld, Sam H. (West Richland, WA)

    1986-01-01

    A welding fixture for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

  5. Welding fixture for nuclear fuel pin cladding assemblies

    DOE Patents [OSTI]

    Oakley, D.J.; Feld, S.H.

    1984-02-22

    A welding fixture is described for locating a driver sleeve about the open end of a nuclear fuel pin cladding. The welding fixture includes a holder provided with an open cavity having shoulders for properly positioning the driver sleeve, the end cap, and a soft, high temperature resistant plastic protective sleeve that surrounds a portion of the end cap stem. Ejected contaminant particles spewed forth by closure of the cladding by pulsed magnetic welding techniques are captured within a contamination trap formed in the holder for ultimate removal and disposal of contaminating particles along with the holder.

  6. Advanced LWR Nuclear Fuel Cladding System Development Trade-off Study |

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

    Department of Energy LWR Nuclear Fuel Cladding System Development Trade-off Study Advanced LWR Nuclear Fuel Cladding System Development Trade-off Study The LWR Sustainability (LWRS) Program activities must support the timeline dictated by utility life extension decisions to demonstrate a lead test rod in a commercial reactor within 10 years. In order to maintain the demanding development schedule that must accompany this aggressive timeline, the LWRS Program focuses on advanced fuel cladding

  7. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

    SciTech Connect (OSTI)

    Lu, Hongbing; Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

    2014-01-09

    Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear fuels are critical to understand the burnup, and thus the fuel efficiency.

  8. U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development

    Broader source: Energy.gov [DOE]

    A significant effort is being placed on silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding by Light Water Reactor Sustainability (LWRS) Advanced Light Water Reactor Nuclear...

  9. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

    SciTech Connect (OSTI)

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an accident tolerant fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

  10. SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

    SciTech Connect (OSTI)

    IJ van Rooyen; WR Lloyd; TL Trowbridge; SR Novascone; KM Wendt; SM Bragg-Sitton

    2013-09-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designs being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points and clad configurations. The 2-ply sleeve samples show a higher bend momentum compared to those of the 1-ply sleeve samples. This is applicable to both the hybrid mock-up and bare SiC-CMC sleeve samples. Comparatively both the 1- and 2-ply hybrid mock-up samples showed a higher bend stiffness and strength compared with the standard Zr-4 mock-up sample. The characterization of the hybrid mock-up samples showed signs of distress and preliminary signs of fraying at the protective Zr-4 sleeve areas for the 1-ply SiC-CMC sleeve. In addition, the microstructure of the SiC matrix near the cracks at the region of highest compressive bending strain shows significant cracking and flaking. The 2-ply SiC-CMC sleeve samples showed a more bonded, cohesive SiC matrix structure. This cracking and fraying causes concern for increased fretting during the actual use of the design. Tomography was proven as a successful tool to identify open porosity during pre-test characterization. Although there is currently insufficient data to make conclusive statements regarding the overall merit of the hybrid cladding design, preliminary characterization of this novel design has been demonstrated.

  11. U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development

    SciTech Connect (OSTI)

    George W. Griffith

    2011-10-01

    A significant effort is being placed on silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding by Light Water Reactor Sustainability (LWRS) Advanced Light Water Reactor Nuclear Fuels Pathway. The intent of this work is to invest in a high-risk, high-reward technology that can be introduced in a relatively short time. The LWRS goal is to demonstrate successful advanced fuels technology that suitable for commercial development to support nuclear relicensing. Ceramic matrix composites are an established non-nuclear technology that utilizes ceramic fibers embedded in a ceramic matrix. A thin interfacial layer between the fibers and the matrix allows for ductile behavior. The SiC CMC has relatively high strength at high reactor accident temperatures when compared to metallic cladding. SiC also has a very low chemical reactivity and doesn't react exothermically with the reactor cooling water. The radiation behavior of SiC has also been studied extensively as structural fusion system components. The SiC CMC technology is in the early stages of development and will need to mature before confidence in the developed designs can created. The advanced SiC CMC materials do offer the potential for greatly improved safety because of their high temperature strength, chemical stability and reduced hydrogen generation.

  12. Fuel pin cladding

    DOE Patents [OSTI]

    Vaidyanathan, S.; Adamson, M.G.

    1983-12-16

    An improved fuel pin cladding, particularly adapted for use in breeder reactors, is described which consist of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel an/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

  13. Fuel pin cladding

    DOE Patents [OSTI]

    Vaidyanathan, Swaminathan; Adamson, Martyn G.

    1986-01-01

    An improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

  14. Fuel pin cladding

    DOE Patents [OSTI]

    Vaidyanathan, S.; Adamson, M.G.

    1986-01-28

    Disclosed is an improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients. 2 figs.

  15. Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

    SciTech Connect (OSTI)

    Mickalonis, J. I.

    2015-08-01

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  16. Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

    SciTech Connect (OSTI)

    Mickalonis, J. I.

    2015-08-31

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  17. CORROSION OF ALUMINUM CLAD SPENT NUCLEAR FUEL IN THE 70 TON CASK DURING TRANSFER FROM L AREA TO H-CANYON

    SciTech Connect (OSTI)

    Mickalonis, J.

    2014-06-01

    Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 260 {degrees}C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 {degrees}C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

  18. Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Digby Macdonald; Mirna Urquidi-Macdonald; Yingzi Chen; Jiahe Ai; Pilyeon Park; Han-Sang Kim

    2006-12-12

    Our work involved the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important parameters, and the development of a code that can be used by reactor operators to actively manage the accumulation of corrosion damage to the fuel cladding and other components in the heat transport circuits in both BWRs and PWRs. In addition, the modified boiling crevice model has been further developed to describe the accumulation of solutes in porous deposits (CRUD) on fuel under boiling (BWRs) and nucleate boiling (PWRs) conditions, in order to accurately describe the environment that is contact with the Zircaloy cladding. In the current report, we have derived expressions for the total steady-state current density and the partial anodic and cathodic current densities to establish a deterministic basis for describing Zircaloy oxidation. The models are deterministic because the relevant natural laws are satisfied explicitly, most importantly the conversation of mass and charge and the equivalence of mass and charge (Faradays law). Cathodic reactions (oxygen reduction and hydrogen evolution) are also included in the models, because there is evidence that they control the rate of the overall passive film formation process. Under open circuit conditions, the cathodic reactions, which must occur at the same rate as the zirconium oxidation reaction, are instrumental in determining the corrosion potential and hence the thickness of the barrier and outer layers of the passive film. Controlled hydrodynamic methods have been used to measure important parameters in the modified Point Defect Model (PDM), which is now being used to describe the growth and breakdown of the passive film on zirconium and on Zircaloy fuel sheathing in BWRs and PWRs coolant environments. The modified PDMs recognize the existence of a thick oxide outer layer over a thin barrier layer. From thermodynamic analysis, it is postulated that a hydride barrier layer forms under PWR coolant conditions whereas an oxide barrier layer forms under BWR primary coolant conditions. Thus, the introduction of hydrogen into the solution lowers the corrosion potential of zirconium to the extent that the formation of ZrH2 is predicted to be spontaneous rather than the ZrO2. Mott-Schottky analysis shows that the passive film formed on zirconium is n-type, which is consistent with the PDM, corresponding to a preponderance of oxygen/hydrogen vacancies and/or zirconium interstitials in the barrier layer. The model parameter values were extracted from electrochemical impedance spectroscopic data for zirconium in high temperature, de-aerated and hydrogenated environments by optimization. The results indicate that the corrosion resistance of zirconium is dominated by the porosity and thickness of the outer layer for both cases. The impedance model based on the PDM provides a good account of the growth of the bi-layer passive films described above, and the extracted model parameter values might be used, for example, for predicting the accumulation of general corrosion damage to Zircaloy fuel sheath in BWR and PWR operating environments. Transients in current density and film thickness for passive film formation on zirconium in dearated and hydrogenated coolant conditions have confirmed that the rate law afforded by the Point Defect Model (PDM) adequately describes the growth and thinning of the passive film. The experimental results demonstrate that the kinetics of oxygen or hydrogen vacancy generation at the metal/film interface control the rate of film growth, when the potential is displaced in the positive direction, whereas the kinetics of dissolution of the barrier layer at the barrier layer/solution interface control the rate of passive film thinning when the potential is stepped in the negative direction. In addition, the effects of second phase particles (SPPs) on the electrochemistry of passive zirconium in the

  19. Advanced Fuels Campaign Cladding & Coatings Meeting Summary

    SciTech Connect (OSTI)

    Not Listed

    2013-03-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) organized a Cladding and Coatings operational meeting February 12-13, 2013, at Oak Ridge National Laboratory (ORNL). Representatives from the U.S. Department of Energy (DOE), national laboratories, industry, and universities attended the two-day meeting. The purpose of the meeting was to discuss advanced cladding and cladding coating research and development (R&D); review experimental testing capabilities for assessing accident tolerant fuels; and review industry/university plans and experience in light water reactor (LWR) cladding and coating R&D.

  20. Uranium and cesium diffusion in fuel cladding of electrogenerating channel

    SciTech Connect (OSTI)

    Vasilev, I. V. Ivanov, A. S.; Churin, V. A.

    2014-12-15

    The results of reactor tests of a carbonitride fuel in a single-crystal cladding from a molybdenum-based alloy can be used in substantiating the operational reliability of fuels in developing a project of a megawatt space nuclear power plant. The results of experimental studies of uranium and cesium penetration into the single-crystal cladding of fuel elements with a carbonitride fuel are interpreted. Those fuel elements passed nuclear power tests in the Ya-82 pilot plant for 8300 h at a temperature of about 1500C. It is shown that the diffusion coefficients for uranium diffusion into the cladding are virtually coincident with the diffusion coefficients measured earlier for uranium diffusion into polycrystalline molybdenum. It is found that the penetration of uranium into the cladding is likely to occur only in the case of a direct contact between the cladding and fuel. The experimentally observed nonmonotonic uranium-concentration profiles are explained in terms of predominant uranium diffusion along grain boundaries. It is shown that a substantially nonmonotonic behavior observed in our experiment for the uranium-concentration profile may be explained by the presence of a polycrystalline structure of the cladding in the surface region from its inner side. The diffusion coefficient is estimated for the grain-boundary diffusion of uranium. The diffusion coefficients for cesium are estimated on the basis of experimental data obtained in the present study.

  1. Impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants. Case study: PWR during routine operations

    SciTech Connect (OSTI)

    Moeller, M.P.; Martin, G.F.; Haggard, D.L.

    1986-01-01

    The purpose of this report is to present data in support of evaluating the impact of fuel cladding failure events on occupational radiation exposure. To determine quantitatively whether fuel cladding failure contributes significantly to occupational radiation exposure, radiation exposure measurements were taken at comparable locations in two mirror-image pressurized-water reactors (PWRs) and their common auxiliary building. One reactor, Unit B, was experiencing degraded fuel characterized as 0.125% fuel pin-hole leakers and was operating at approximately 55% of the reactor's licensed maximum core power, while the other reactor, Unit A, was operating under normal conditions with less than 0.01% fuel pin-hole leakers at 100% of the reactor's licensed maximum core power. Measurements consisted of gamma spectral analyses, radiation exposure rates and airborne radionuclide concentrations. In addition, data from primary coolant sample results for the previous 20 months on both reactor coolant systems were analyzed. The results of the measurements and coolant sample analyses suggest that a 3560-megawatt-thermal (1100 MWe) PWR operating at full power with 0.125% failed fuel can experience an increase of 540% in radiation exposure rates as compared to a PWR operating with normal fuel. In specific plant areas, the degraded fuel may elevate radiation exposure rates even more.

  2. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    SciTech Connect (OSTI)

    Rudisill, T; John Mickalonis, J

    2006-09-27

    The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO{sub 2}) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO{sub 2} layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH{sub 4}F)/ammonium nitrate (NH{sub 4}NO{sub 3}) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO{sub 2} layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH{sub 4}){sub 2}ZrF{sub 6}) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of contamination. The thermal decomposition of this material is also undesirable if the cladding hulls are melted for volume reduction or to produce waste forms. Handling and disposal of the corrosive off-gas stream and ZrO{sub 2}-containing dross must be addressed. The stability of Zr{sup 4+} in the NHF{sub 4}/NH{sub 4}NO{sub 3} solution is also a concern. Precipitation of ammonium zirconium fluorides upon cooling of the dissolving solution was observed in the feasibility experiments. Precipitation of the solids was attributed to the high fluoride to Zr ratios used in the experiments. The solubility of Zr{sup 4+} in NH{sub 4}F solutions decreases as the free fluoride concentration increases. The removal of the ZrO{sub 2} layer from Zircaloy-4 coupons with HF showed a strong dependence on both the concentration and temperature. Very rapid dissolution of the oxide layer and significant amounts of metal was observed in experiments using HF concentrations {ge} 2.5 M. Treatment of the coupons using HF concentrations {le} 1.0 M was very effective in removing the oxide layer. The most effective conditions resulted in dissolution rates which were less than approximately 2 mg/cm{sup 2}-min. With dissolution rates in this range, uniform removal of the oxide layer was obtained and a minimal amount of Zircaloy metal was dissolved. Future HF dissolution studies should focus on the decontamination of actual spent fuel cladding hulls to determine if the treated hulls meet criteria for disposal as a LLW.

  3. Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

    SciTech Connect (OSTI)

    Tome, Carlos N; Caro, J A; Lebensohn, R A; Unal, Cetin; Arsenlis, A; Marian, J; Pasamehmetoglu, K

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems to develop predictive tools is critical. Not only are fabrication and performance models needed to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.

  4. The impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants: Case study, PWR (pressurized-water reactor) during an outage

    SciTech Connect (OSTI)

    Moeller, M.P.; Martin, G.F.; Kenoyer, J.L.

    1987-08-01

    This report is the second in a series of case studies designed to evaluate the magnitude of increase in occupational radiation exposures at commercial US nuclear power plants resulting from small incidents or abnormal events. The event evaluated is fuel cladding failure, which can result in elevated primary coolant activity and increased radiation exposure rates within a plant. For this case study, radiation measurements were made at a pressurized-water reactor (PWR) during a maintenance and refueling outage. The PWR had been operating for 22 months with fuel cladding failure characterized as 105 pin-hole leakers, the equivalent of 0.21% failed fuel. Gamma spectroscopy measurements, radiation exposure rate determinations, thermoluminescent dosimeter (TLD) assessments, and air sample analyses were made in the plant's radwaste, pipe penetration, and containment buildings. Based on the data collected, evaluations indicate that the relative contributions of activation products and fission products to the total exposure rates were constant over the duration of the outage. This constancy is due to the significant contribution from the longer-lived isotopes of cesium (a fission product) and cobalt (an activation product). For this reason, fuel cladding failure events remain as significant to occupational radiation exposure during an outage as during routine operations. As documented in the previous case study (NUREG/CR-4485 Vol. 1), fuel cladding failure events increased radiation exposure rates an estimated 540% at some locations of the plant during routine operations. Consequently, such events can result in significantly greater radiation exposure rates in many areas of the plant during the maintenance and refueling outages than would have been present under normal fuel conditions.

  5. Nuclear fuel element

    DOE Patents [OSTI]

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1983-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

  6. Nuclear reactor fuel element

    DOE Patents [OSTI]

    Johnson, Carl E. (Elk Grove, IL); Crouthamel, Carl E. (Richland, WA)

    1980-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

  7. Nuclear fuel element

    DOE Patents [OSTI]

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

    A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

  8. Nuclear fuel element

    DOE Patents [OSTI]

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1980-04-29

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

  9. Characterization of Fuel-Cladding Bond Strength Using Laser Shock

    SciTech Connect (OSTI)

    James A. Smith; David L. Cottle; Barry H. Rabin

    2014-04-01

    This paper describes new laser-based capabilities for characterization of fuel-cladding bond strength in nuclear fuels, and presents preliminary results obtained from studies on as-fabricated monolithic fuel consisting of uranium-10 wt.% molybdenum alloys clad in 6061 aluminum by hot isostatic pressing. Two complementary experimental methods are employed, laser-shock testing and laser-ultrasonic imaging. Measurements are spatially localized, non-contacting and require minimum specimen preparation, and are therefore ideally suited for applications involving radioactive materials, including irradiated materials. The theoretical principles and experimental approaches employed in characterization of nuclear fuel plates are described. The ability to measure layer thicknesses, elastic properties of the constituents, and the location and nature of laser-shock induced debonds is demonstrated, and preliminary bond strength measurement results are discussed.

  10. Evaluation of hoop creep behaviors in long-term dry storage condition of pre-hydrided and high burn-up nuclear fuel cladding

    SciTech Connect (OSTI)

    Kim, Sun-Ki; Bang, J.G.; Kim, D.H.; Yang, Y.S.

    2007-07-01

    Related to the degradation of the mechanical properties of Zr-based nuclear fuel cladding tubes under long term dry storage condition, the mechanical tests which can simulate the degradation of the mechanical properties properly are needed. Especially, the degradation of the mechanical properties by creep mechanism seems to be dominant under long term dry storage condition. Accordingly, in this paper, ring creep tests were performed in order to evaluate the creep behaviors of high burn-up fuel cladding under a hoop loading condition in a hot cell. The tests are performed with Zircaloy-4 fuel cladding whose burn-up is approximately {approx}60,000 MWd/tU in the temperature range from 350 deg. to 550 deg.. The tests are also performed with pre-hydrided Zircaloy-4 and ZIRLO up to 1,000 ppm. First of all, the hoop loading grip for the ring creep test was designed in order that a constant curvature of the specimen was maintained during the creep deformation, and the graphite lubricant was used to minimize the friction between the outer surface of the die insert and the inner surface of the ring specimen. The specimen for the ring creep test was designed to limit the deformation within the gauge section and to maximize the uniformity of the strain distribution. It was confirmed that the mechanical properties under a hoop loading condition can be correctly evaluated by using this test technique. In this paper, secondary creep rate with increasing hydrogen content are drawn, and then kinetic data such as pre-exponential factor and activation energy for creep process are also drawn. In addition, creep life are predicted by obtaining LMP (Larson-Miller parameter) correlation in the function of hydrogen content and applied stress to yield stress ratio. (authors)

  11. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented. 6 references, 4 figures.

  12. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented.

  13. Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

    SciTech Connect (OSTI)

    Robinson, Sharon M.; Chattin, Marc Rhea; Giaquinto, Joseph; Jubin, Robert Thomas

    2015-09-01

    It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritium labelled water (THO) or T2O. In a standard processing flowsheet, tritium management would be accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste for Zircaloy clad fuel by recovering the zirconium from the cladding for reuse. These recycle processes could release the tritium in the cladding. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 096%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. Samples of Surry-2 and H.B. Robinson pressurized water reactor cladding were heated to 11001200C to oxidize the zirconium and release all of the tritium in the cladding sample. Cladding samples were also heated within the temperature range of 480600C expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700C) to determine the impact of tritium pretreatment on tritium release from the cladding. The tritium content of the Surry-2 and H.B. Robinson cladding was measured to be ~234 and ~500 Ci/g, respectively. Heating the Surry-2 cladding at 500C for 24 h removed ~0.2% of the tritium from the cladding, and heating at 700C for 24 h removed ~9%. Heating the H.B. Robinson cladding at 700C for 24 h removed ~11% of the tritium. When samples of the Surry-2 and H.B. Robinson claddings were heated at 700C for 96 h, essentially all of the tritium in the cladding was removed. However, only ~3% of the tritium was removed when a sample of Surry-2 cladding was heated at 600C for 96 h. These data indicate that the amount of tritium released from tritium pretreatment systems will be dependent on both the operating temperature and length of time in the system. Under certain conditions, a significant fraction of the tritium could remain bound in the cladding and would need to be considered in operations involving cladding recycle.

  14. Structure of high-burnup-fuel Zircaloy cladding. [PWR; BWR

    SciTech Connect (OSTI)

    Chung, H.M.

    1983-06-01

    Zircaloy cladding from high-burnup (> 20 MWd/kg U) fuel rods in light-water reactors is characterized by a high density of irradiation-induced defects (RID), compositional changes (e.g., oxygen and hydrogen uptake) associated with in-service corrosion, and geometrical changes produced by creepdown, bowing, and irradiation-induced growth. During a reactor power transient, the cladding is subject to localized stress imposed by thermal expansion of the cracked fuel pellets and to mechanical constraints imposed by pellet-cladding friction. As part of a program to provide a better understanding of brittle-type failure of Zircaloy fuel cladding by pellet-cladding interaction (PCI) phenomenon, the stress-rupture properties and microstructural characteristics of high-burnup spent fuel cladding have been under investigation. This paper reports the results of the microstructural examinations by optical microscopy, scanning (SEM), 100-keV transmission (TEM), and 1 MeV high-voltage (HVEM) electron microscopies of the fractured spent fuel cladding with a specific empahsis on a correlation of the structural characteristics with the fracture behavior.

  15. Surface Modification of Fuel Cladding Materials with Integral Fuel BUrnable Absorber Boron

    SciTech Connect (OSTI)

    Dr. Kumar Sridharan; Dr. Todd Allen; Jesse Gudmundson; Benjamin Maier

    2008-11-03

    Integral fuel burnable absorgers (IFBA) are added to some rods in the fuel assembly to counteract excessive reactivity. These IFBA elements (usually boron or gadolinium) are presently incorporated in the U)2 pellets either by mixing in the pellets or as coatings on the pellet surface. In either case, the incorporation of ifba into the fuel has to be performed in a nuclear-regulated facility that is physically separated from the main plant. These operations tend to be costly and can add from 20 to 30% to the manufacturing cost of the fuel. The goal of this NEER research project was to develop an alternative approach that involves incorporation of IFBA element boron at the surface of the fuel cladding material.

  16. Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding

    SciTech Connect (OSTI)

    Jason D. Hales; Various

    2014-09-01

    As a result of corrosion during normal operation in nuclear reactors, hydrogen can enter the zirconium-alloy fuel cladding and precipitate as brittle hydride platelets, which can severely degrade the cladding ductility. Under a heterogeneous temperature distribution, hydrides tend to accumulate in the colder areas, creating local spots of degraded cladding that can favor crack initiation. Therefore, an estimation of the local hydride distribution is necessary to help predict the risk of cladding failure. The hydride distribution is governed by three competing phenomena. Hydrogen in solid solution diffuses under a concentration gradient due to Ficks law and under a temperature gradient due to the Soret effect. Precipitation of the hydride platelets occurs once the hydrogen solubility limit is reached. A model of these phenomena was implemented in the 3D fuel performance code BISON in order to calculate the hydrogen distribution for arbitrary geometries, such as a nuclear fuel rod, and is now available for BISON users. Simulations have been performed on simple geometries to validate the model and its implementation. The simulations predict that before precipitation occurs, hydrogen tends to accumulate in the colder spots due to the Soret effect. Once the solubility limit is reached, hydrogen precipitates and forms a rim close to the outer edge of the cladding. The simulations also predict that the reactor shut down has little effect on already precipitated hydrides but causes the remaining hydrogen to precipitate homogeneously into hydrides.

  17. Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

    SciTech Connect (OSTI)

    Isabella J van Rooyen

    2012-09-01

    The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

  18. Advanced surface plasma nitriding for development of corrosion resistant and accident tolerant fuel cladding

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

    surface plasma nitriding for development of corrosion resistant and accident tolerant fuel cladding PI: Lin Shao, Texas A&M University Collaborators: Don A. Lucca - Oklahoma State University, Michael P. Short - Massachusetts Institute of Technology, Frank Garner - Texas A&M University Program: Nuclear Energy Enabling Technologies Presented by Elizabeth Castanon, Ph.D candidate in Nuclear Eng., Texas A&M University This project aims to develop a new plasma nitriding technique which is

  19. Evolution of pores in the fuel cladding of the electrogenerating channel

    SciTech Connect (OSTI)

    Vasilev, I. V. Ivanov, A. S.

    2014-12-15

    The results of reactor tests of carbonitride fuel in a monocrystalline cladding from a molybdenum-based alloy that were carried out earlier on an experimental setup Ya-82 for 8 to 300 h at a temperature of ?1500C can be used in order to prove the operational reliability of fuel elements in the design of a megawatt nuclear power plant for a spacecraft. A raster image of the surface of a sample of cladding shows that the interfaces between the layers are decorated by pores. This result is explained in this work by the theory of coalescence. The mechanisms responsible for the evolution of pores taking place at the parameters of a Ya-82 setup are considered. The effect of decoration by pores of the interfaces between the layers of a sample of cladding of the electrogenerating channel by carrying out reactor tests is explained. The dependence of the average radius of pores on the duration of the experiment is obtained. An evaluation of the average sizes of pores arising under the conditions of the experiment gives a value of ?2 ?m, which is in agreement with the experimental data. A computational study of swelling of the cladding material in the process of irradiation is performed. Predictive estimates of the behavior of the porous system and swelling of the cladding material for a megawatt class nuclear power plant are made.

  20. Material selection for accident tolerant fuel cladding

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

    Pint, B. A.; Terrani, K. A.; Yamamoto, Y.; Snead, L. L.

    2015-09-14

    Alternative cladding materials are being investigated for accident tolerance, which can be defined as >100X improvement (compared to current Zr-based alloys) in oxidation resistance in steam environments at ≥1200°C for short (≤4 h) times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. Therefore, commercial Ti2AlC that is not single phase, formed a much thicker oxide at 1200°C in steammore » and significant TiO2, and therefore may be challenging to use as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation assisted Cr-rich α’ formation. The composition effects and critical limits to retaining protective scale formation at >1400°C are still being evaluated.« less

  1. Material selection for accident tolerant fuel cladding

    SciTech Connect (OSTI)

    Pint, B. A.; Terrani, K. A.; Yamamoto, Y.; Snead, L. L.

    2015-09-14

    Alternative cladding materials are being investigated for accident tolerance, which can be defined as >100X improvement (compared to current Zr-based alloys) in oxidation resistance in steam environments at ?1200C for short (?4 h) times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. Therefore, commercial Ti2AlC that is not single phase, formed a much thicker oxide at 1200C in steam and significant TiO2, and therefore may be challenging to use as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475C, while reducing its Cr content to minimize susceptibility to irradiation assisted Cr-rich ? formation. The composition effects and critical limits to retaining protective scale formation at >1400C are still being evaluated.

  2. Coupon Surveillance For Corrosion Monitoring In Nuclear Fuel Basin

    SciTech Connect (OSTI)

    Mickalonis, J. I.; Murphy, T. R.; Deible, R.

    2012-10-01

    Aluminum and stainless steel coupons were put into a nuclear fuel basin to monitor the effect of water chemistry on the corrosion of fuel cladding. These coupons have been monitored for over ten years. The corrosion and pitting data is being used to model the kinetics and estimate the damage that is occurring to the fuel cladding.

  3. Novel Accident-Tolerant Fuel Meat and Cladding

    SciTech Connect (OSTI)

    Robert D. Mariani; Pavel G Medvedev; Douglas L Porter; Steven L Hayes; James I. Cole; Xian-Ming Bai

    2013-09-01

    A novel accident-tolerant fuel meat and cladding are here proposed. The fuel meat design incorporates annular fuel with inserts and discs that are fabricated from a material having high thermal conductivity, for example niobium. The inserts are rods or tubes. Discs separate the fuel pellets. Using the BISON fuel performance code it was found that the peak fuel temperature can be lowered by more than 600 degrees C for one set of conditions with niobium metal as the thermal conductor. In addition to improved safety margin, several advantages are expected from the lower temperature such as decreased fission gas release and fuel cracking. Advantages and disadvantages are discussed. An enrichment of only 7.5% fully compensates the lost reactivity of the displaced UO2. Slightly higher enrichments, such as 9%, allow uprates and increased burnups to offset the initial costs for retooling. The design has applications for fast reactors and transuranic burning, which may accelerate its development. A zirconium silicide coating is also described for accident tolerant applications. A self-limiting degradation behavior for this coating is expected to produce a glassy, self-healing layer that becomes more protective at elevated temperature, with some similarities to MoSi2 and other silicides. Both the fuel and coating may benefit from the existing technology infrastructure and the associated wide expertise for a more rapid development in comparison to other, more novel fuels and cladding.

  4. Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO? fuel and APMT-Steel Clad/UN-U?Si? fuel concepts

    SciTech Connect (OSTI)

    Unal, Cetin; Galloway, Jack D.

    2014-09-12

    In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermal swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO? ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U?Si? fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.

  5. Milestone report - M4FT-14OR0302102b - Evaluation of Tritium Content and Release from Surry-2 Fuel Cladding

    SciTech Connect (OSTI)

    Robinson, Sharon M.; Chattin, Marc Rhea; Giaquinto, Joseph M.; Jubin, Robert Thomas

    2014-09-01

    To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified.To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. A sample of Surry-2 pressurized water reactor (PWR) cladding was heated to 11001200C to oxidize the zirconium and release all of the tritium in the cladding sample. The tritium content was measured to be ~240 Ci/g. Cladding samples were heated to 500C, which is within the temperature range (480 - 600C) expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700C) to determine the impact of tritium pretreatment on tritium release from the cladding. Heating at 500C for 24 hr removes ~0.2% of the tritium from the cladding, and heating at 700C for 24 hr removes ~9%. Thus, a significant fraction of the tritium remains bound in the cladding and must be considered in operations involving cladding recycle.

  6. High Temperature Steam Corrosion of Cladding for Nuclear Applications: Experimental

    SciTech Connect (OSTI)

    McHugh, Kevin M; Garnier, John E; Sergey Rashkeev; Michael V. Glazoff; George W. Griffith; Shannong M. Bragg-Sitton

    2013-01-01

    Stability of cladding materials under off-normal conditions is an important issue for the safe operation of light water nuclear reactors. Metals, ceramics, and metal/ceramic composites are being investigated as substitutes for traditional zirconium-based cladding. To support down-selection of these advanced materials and designs, a test apparatus was constructed to study the onset and evolution of cladding oxidation, and deformation behavior of cladding materials, under loss-of-coolant accident scenarios. Preliminary oxidation tests were conducted in dry oxygen and in saturated steam/air environments at 1000OC. Tube samples of Zr-702, Zr-702 reinforced with 1 ply of a -SiC CMC overbraid, and sintered a-SiC were tested. Samples were induction heated by coupling to a molybdenum susceptor inside the tubes. The deformation behavior of He-pressurized tubes of Zr-702 and SiC CMC-reinforced Zr-702, heated to rupture, was also examined.

  7. Nuclear Fuel Reprocessing

    SciTech Connect (OSTI)

    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 35% 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 fuels have been irradiated for different purposes, but the vast majority of commercial fuel is uranium oxide clad in zirconium alloy tubing. As a result, commercial reprocessing plants have relatively narrow technical requirements for used nuclear that is accepted for processing.

  8. Embrittlement and DBTT of High-Burnup PWR Fuel Cladding Alloys | Department

    Office of Environmental Management (EM)

    of Energy Embrittlement and DBTT of High-Burnup PWR Fuel Cladding Alloys Embrittlement and DBTT of High-Burnup PWR Fuel Cladding Alloys Structural analyses of high-burnup (HBU) fuel require cladding mechanical properties and failure limits to assess fuel behavior during long-term dry-cask storage and transportation. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and pressure-induced tensile hoop stresses relative to in-reactor operation

  9. FY2015 Status Report: CIRFT Testing of High-Burnup Used Nuclear Fuel Rods from Pressurized Water Reactor and BWR Environments

    Broader source: Energy.gov [DOE]

    High-burnup spent nuclear fuel cladding has a significant amount of microcracks and hydrides which will reduce the stress intensity required for crack growth.  Characteristics of cladding, fuel...

  10. Advanced nuclear fuel

    SciTech Connect (OSTI)

    Terrani, Kurt

    2014-07-14

    Kurt Terrani uses his expertise in materials science to develop safer fuel for nuclear power plants.

  11. Advanced nuclear fuel

    ScienceCinema (OSTI)

    Terrani, Kurt

    2014-07-15

    Kurt Terrani uses his expertise in materials science to develop safer fuel for nuclear power plants.

  12. CONTAINMENT ANALYSIS METHODOLOGY FOR TRANSPORT OF BREACHED CLAD ALUMINUM SPENT FUEL

    SciTech Connect (OSTI)

    Vinson, D.

    2010-07-11

    Aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site and placed in interim storage in a water basin. To enter the United States, a cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Many Al-SNF assemblies have suffered corrosion degradation in storage in poor quality water, and many of the fuel assemblies are 'failed' or have through-clad damage. A methodology was developed to evaluate containment of Al-SNF even with severe cladding breaches for transport in standard casks. The containment analysis methodology for Al-SNF is in accordance with the methodology provided in ANSI N14.5 and adopted by the U. S. Nuclear Regulatory Commission in NUREG/CR-6487 to meet the requirements of 10CFR71. The technical bases for the inputs and assumptions are specific to the attributes and characteristics of Al-SNF received from basin and dry storage systems and its subsequent performance under normal and postulated accident shipping conditions. The results of the calculations for a specific case of a cask loaded with breached fuel show that the fuel can be transported in standard shipping casks and maintained within the allowable release rates under normal and accident conditions. A sensitivity analysis has been conducted to evaluate the effects of modifying assumptions and to assess options for fuel at conditions that are not bounded by the present analysis. These options would include one or more of the following: reduce the fuel loading; increase fuel cooling time; reduce the degree of conservatism in the bounding assumptions; or measure the actual leak rate of the cask system. That is, containment analysis for alternative inputs at fuel-specific conditions and at cask-loading-specific conditions could be performed to demonstrate that release is within the allowable leak rates of the cask.

  13. Spent Nuclear Fuel

    Gasoline and Diesel Fuel Update (EIA)

    Nuclear & Uranium Glossary › FAQS › Overview Data Status of U.S. Nuclear Outages (interactive) Summary Uranium & nuclear fuel Nuclear power plants Spent nuclear fuel International All nuclear data reports Analysis & Projections Major Topics Most popular Nuclear plants and reactors Projections Recurring Uranium All reports Browse by Tag Alphabetical Frequency Tag Cloud Previous releases 2002 1998 Spent Nuclear Fuel Release date: December 7, 2015 Next release date: Late 2018 Spent

  14. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    SciTech Connect (OSTI)

    Field, Kevin G.; Howard, Richard H.

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. Additionally, they exhibit other behaviors conducive to nuclear reactor core internal applications including having the same excellent radiation-induced cavity swelling resistance as other ferritic alloy systems [1], a high resistance to environmental stress corrosion cracking (SCC) [2], and they maintain good mechanical properties at and above normal operating conditions of LWRs [3]. The culmination of these properties has enticed the research and development (R&D) community to strongly consider using FeCrAl alloys as a new cladding option with enhanced accident tolerance over the industry standard Zr-based cladding options. In particular, the high temperature steam oxidation resistance is the main attractive feature for using FeCrAl alloys as an accident tolerant cladding option in LWR applications. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI).

  15. Examination of Zircaloy-clad spent fuel after extended pool storage

    SciTech Connect (OSTI)

    Bradley, E.R.; Bailey, W.J.; Johnson, A.B. Jr.; Lowry, L.M.

    1981-09-01

    This report presents the results from metallurgical examinations of Zircaloy-clad fuel rods from two bundles (0551 and 0074) of Shippingport PWR Core 1 blanket fuel after extended water storage. Both bundles were exposed to water in the reactor from late 1957 until discharge. The estimated average burnups were 346 GJ/kgU (4000 MWd/MTU) for bundle 0551 and 1550 GJ/kgU (18,000 MWd/MTU) for bundle 0074. Fuel rods from bundle 0551 were stored in deionized water for nearly 21 yr prior to examination in 1980, representing the world's oldest pool-stored Zircaloy-clad fuel. Bundle 0074 has been stored in deionized water since reactor discharge in 1964. Data from the current metallurgical examinations enable a direct assessment of extended pool storage effects because the metallurgical condition of similar fuel rods was investigated and documented soon after reactor discharge. Data from current and past examinations were compared, and no significant degradation of the Zircaloy cladding was indicated after almost 21 yr in water storage. The cladding dimensions and mechanical properties, fission gas release, hydrogen contents of the cladding, and external oxide film thicknesses that were measured during the current examinations were all within the range of measurements made on fuel bundles soon after reactor discharge. The appearance of the external surfaces and the microstructures of the fuel and cladding were also similar to those reported previously. In addition, no evidence of accelerated corrosion or hydride redistribution in the cladding was observed.

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

    SciTech Connect (OSTI)

    Rebak, Raul B.

    2014-12-30

    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 provide hermetic seal. The replacement of a zirconium alloy using a ferritic material containing chromium and aluminum appears to be the most near term implementation for accident tolerant nuclear fuels.

  17. Structural analysis of fuel assembly clads for the Upgraded Transient Reactor Test Facility (TREAT Upgrade)

    SciTech Connect (OSTI)

    Ewing, T.F.; Wu, T.S.

    1986-01-01

    The Upgraded Transient Reactor Test Facility (TREAT Upgrade) is designed to test full-length, pre-irradiated fuel pins of the type used in large LMFBRs under accident conditions, such as severe transient overpower and loss-of-coolant accidents. In TREAT Upgrade, the central core region is to contain new fuel assemblies of higher fissile loadings to maximize the energy deposition to the test fuel. These fuel assemblies must withstand normal peak clad temperatures of 850/sup 0/C for hundreds of test transients. Due to high temperatures and gradients predicted in the clad, creep and plastic strain effects are significant, and the clad structural behavior cannot be analyzed by conventional linear techniques. Instead, the detailed elastic-plastic-creep behavior must be followed along the time-dependent load history. This paper presents details of the structural evaluations of the conceptual TREAT Upgrade fuel assembly clads.

  18. Laser-Based Characterization of Nuclear Fuel Plates

    SciTech Connect (OSTI)

    James A. Smith; David L. Cottle; Barry H. Rabin

    2013-07-01

    Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

  19. Models for the Configuration and Integrity of Partially Oxidized Fuel Rod Cladding at High Temperatures

    SciTech Connect (OSTI)

    Siefken, L.J.

    1999-01-01

    Models were designed to resolve deficiencies in the SCDAP/RELAP5/MOD3.2 calculations of the configuration and integrity of hot, partially oxidized cladding. These models are expected to improve the calculations of several important aspects of fuel rod behavior. First, an improved mapping was established from a compilation of PIE results from severe fuel damage tests of the configuration of melted metallic cladding that is retained by an oxide layer. The improved mapping accounts for the relocation of melted cladding in the circumferential direction. Then, rules based on PIE results were established for calculating the effect of cladding that has relocated from above on the oxidation and integrity of the lower intact cladding upon which it solidifies. Next, three different methods were identified for calculating the extent of dissolution of the oxidic part of the cladding due to its contact with the metallic part. The extent of dissolution effects the stress and thus the integrity of the oxidic part of the cladding. Then, an empirical equation was presented for calculating the stress in the oxidic part of the cladding and evaluating its integrity based on this calculated stress. This empirical equation replaces the current criterion for loss of integrity which is based on temperature and extent of oxidation. Finally, a new rule based on theoretical and experimental results was established for identifying the regions of a fuel rod with oxidation of both the inside and outside surfaces of the cladding. The implementation of these models is expected to eliminate the tendency of the SCDAP/RELAP5 code to overpredict the extent of oxidation of the upper part of fuel rods and to underpredict the extent of oxidation of the lower part of fuel rods and the part with a high concentration of relocated material. This report is a revision and reissue of the report entitled, Improvements in Modeling of Cladding Oxidation and Meltdown.

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

    DOE Patents [OSTI]

    Johnson, Gerald D. (Kennewick, WA); Lobsinger, Ralph J. (Kennewick, WA); Hamilton, Margaret L. (Richland, WA); Gelles, David S. (West Richland, WA)

    1990-01-01

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

  1. Cladding and Structural Materials for Advanced Nuclear Energy Systems

    SciTech Connect (OSTI)

    Was, G S; Allen, T R; Ila, D; C,; Levi,; Morgan, D; Motta, A; Wang, L; Wirth, B

    2011-06-30

    The goal of this consortium is to address key materials issues in the most promising advanced reactor concepts that have yet to be resolved or that are beyond the existing experience base of dose or burnup. The research program consists of three major thrusts: 1) high-dose radiation stability of advanced fast reactor fuel cladding alloys, 2) irradiation creep at high temperature, and 3) innovative cladding concepts embodying functionally-graded barrier materials. This NERI-Consortium final report represents the collective efforts of a large number of individuals over a period of three and a half years and included 9 PIs, 4 scientists, 3 post-docs and 12 students from the seven participating institutions and 8 partners from 5 national laboratories and 3 industrial institutions (see table). University participants met semi-annually and participants and partners met annually for meetings lasting 2-3 days and designed to disseminate and discuss results, update partners, address outstanding issues and maintain focus and direction toward achieving the objectives of the program. The participants felt that this was a highly successful program to address broader issues that can only be done by the assembly of a range of talent and capabilities at a more substantial funding level than the traditional NERI or NEUP grant. As evidence of the success, this group, collectively, has published 20 articles in archival journals and made 57 presentations at international conferences on the results of this consortium.

  2. Development of Diffusion barrier coatings and Deposition Technologies for Mitigating Fuel Cladding Chemical Interactions (FCCI)

    SciTech Connect (OSTI)

    Sridharan, Kumar; Allen, Todd; Cole, James

    2013-02-27

    The goal of this project is to develop diffusion barrier coatings on the inner cladding surface to mitigate fuel-cladding chemical interaction (FCCI). FCCI occurs due to thermal and radiation enhanced inter-diffusion between the cladding and fuel materials, and can have the detrimental effects of reducing the effective cladding wall thickness and lowering the melting points of the fuel and cladding. The research is aimed at the Advanced Burner Reactor (ABR), a sodium-cooled fast reactor, in which higher burn-ups will exacerbate the FCCI problem. This project will study both diffusion barrier coating materials and deposition technologies. Researchers will investigate pure vanadium, zirconium, and titanium metals, along with their respective oxides, on substrates of HT-9, T91, and oxide dispersion-strengthened (ODS) steels; these materials are leading candidates for ABR fuel cladding. To test the efficacy of the coating materials, the research team will perform high-temperature diffusion couple studies using both a prototypic metallic uranium fuel and a surrogate?¢????the rare-earth element lanthanum. Ion irradiation experiments will test the stability of the coating and the coating-cladding interface. A critical technological challenge is the ability to deposit uniform coatings on the inner surface of cladding. The team will develop a promising non-line-of-sight approach that uses nanofluids . Recent research has shown the feasibility of this simple yet novel approach to deposit coatings on test flats and inside small sections of claddings. Two approaches will be investigated: 1) modified electrophoretic deposition (MEPD) and 2) boiling nanofluids. The coatings will be evaluated in the as-deposited condition and after sintering.

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

    SciTech Connect (OSTI)

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

    2014-01-01

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

  4. Vented nuclear fuel element

    DOE Patents [OSTI]

    Grossman, Leonard N.; Kaznoff, Alexis I.

    1979-01-01

    A nuclear fuel cell for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel.

  5. Early implementation of SiC cladding fuel performance models in BISON

    SciTech Connect (OSTI)

    Powers, Jeffrey J.

    2015-09-18

    SiC-based ceramic matrix composites (CMCs) [5–8] are being developed and evaluated internationally as potential LWR cladding options. These development activities include interests within both the DOE-NE LWR Sustainability (LWRS) Program and the DOE-NE Advanced Fuels Campaign. The LWRS Program considers SiC ceramic matrix composites (CMCs) as offering potentially revolutionary gains as a cladding material, with possible benefits including more efficient normal operating conditions and higher safety margins under accident conditions [9]. Within the Advanced Fuels Campaign, SiC-based composites are a candidate ATF cladding material that could achieve several goals, such as reducing the rates of heat and hydrogen generation due to lower cladding oxidation rates in HT steam [10]. This work focuses on the application of SiC cladding as an ATF cladding material in PWRs, but these work efforts also support the general development and assessment of SiC as an LWR cladding material in a much broader sense.

  6. FY 2014 Status Report: of Vibration Testing of Clad Fuel (M4FT-14OR0805033)

    SciTech Connect (OSTI)

    Bevard, Bruce Balkcom

    2014-03-28

    The DOE Used Fuel Disposition Campaign (UFDC) tasked Oak Ridge National Laboratory (ORNL) to investigate the behavior of light-water-reactor (LWR) fuel cladding material performance related to extended storage and transportation of UNF. ORNL has been tasked to perform a systematic study on UNF integrity under simulated normal conditions of transportation (NCT) by using the recently developed hot-cell testing equipment, Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT). To support the testing on actual high-burnup UNF, fast-neutron irradiation of pre-hydrided zirconium-alloy cladding in the High Flux Isotope Reactor (HFIR) at elevated temperatures will be used to simulate the effects of high-burnup on fuel cladding for help in understanding the cladding materials properties relevant to extended storage and subsequent transportation. The irradiated pre-hydrided metallic materials testing will generate baseline data to benchmark hot-cell testing of the actual high-burnup UNF cladding. More importantly, the HFIR-irradiated samples will be free of alpha contamination and can be provided to researchers who do not have hot cell facilities to handle highly contaminated high-burnup UNF cladding to support their research projects for the UFDC.

  7. Methodology for Mechanical Property Testing of Fuel Cladding Using a Expanded Plug Wedge Test

    SciTech Connect (OSTI)

    Jiang, Hao; Wang, Jy-An John

    2014-01-01

    An expanded plug method was developed earlier for determining the tensile properties of irradiated fuel cladding. This method tests fuel rod cladding ductility by utilizing an expandable plug to radially stretch a small ring of irradiated cladding material. The circumferential or hoop strain is determined from the measured diametrical expansion of the ring. A developed procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves, from which material properties of the cladding can be extracted. However, several deficiencies existed in this expanded-plug test that can impact the accuracy of test results, such as that the large axial compressive stress resulted from the expansion plug test can potentially induce the shear failure mode of the tested specimen. Moreover, highly nonuniform stress and strain distribution in the deformed clad gage section and significant compressive stresses, induced by bending deformation due to clad bulging effect, will further result in highly nonconservative estimates of the mechanical properties for both strength and ductility of the tested clad. To overcome the aforementioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. By optimizing the specific geometry designs, selecting the appropriate material for the expansion plug, and adding new components into the testing system, a modified expansion plug testing protocol has been developed. A general procedure was also developed to determine the hoop stress in the tested ring specimen. A scaling factor, -factor, was used to convert the ring load Fring into hoop stress , and is written as _ = F_ring/tl , where t is the clad thickness and l is the clad length. The generated stress-strain curve agrees well with the associated tensile test data in both elastic and plastic deformation regions.

  8. Fuel Performance Experience, Analysis and Modeling: Deformations, Fission Gas Release and Pellet-Clad Interaction

    SciTech Connect (OSTI)

    Zhou, G.; Hallstadius, L.; Helmersson, S.; Massih, A.R.; Schrire, D.; Kaellstroem, R.; Wikmark, G.; Hellwig, C.

    2007-07-01

    Some basic attributes of light water reactor fuel performance, determined by measurements, are evaluated. In particular, data on fuel volume swelling, cladding creep/growth, fission product gas release and cladding deformation due to pellet-clad mechanical interaction of rods irradiated in power reactors to rod burnups up to about 70 MWd/kgU are presented and appraised. A thermal fuel matrix swelling caused by fission products shows a linear increase in the fuel volume fraction with burnup up to 70 MWd/kgU with a mean rate of 0.76% per 10 MWd/kgU at a best-estimate level. Cladding hoop strain data due to in-reactor creep as a function of burnup from 15 to 70 MWd/kgU for pressurized water reactor (PWR) rods and from 5 to 50 MWd/kgU for boiling water reactor (BWR) rods are presented. The maximum measured cladding creep-down hoop strain in the considered BWR rods is {epsilon}{sub {theta}} {approx_equal} -0.5% and in the PWR rods {epsilon}{sub {theta}} {approx_equal} -1.25%. Rod growth data on BWR and PWR rods as a function of burnup are presented and discussed. Rod internal free volume data, measured and calculated as a function of burnup, are presented. Recent high burnup (52-70 MWd/kgU) fission product gas release data obtained by destructive methods are evaluated with the STAV7 computer code. Finally, slow power ramp experiments conducted at the Studsvik R2 reactor are simulated with the STAV7 code and it is observed that by accounting the contribution of fuel thermal gaseous swelling, the code describes the clad diameter increase due to pellet-clad mechanical interaction under the power bump satisfactorily. (authors)

  9. Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

    SciTech Connect (OSTI)

    Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom; Howard, Rob L

    2014-01-01

    Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

  10. CEPAN method of analyzing creep collapse of oval cladding. Volume 5. Evaluation of interpellet gap formation and clad collapse in modern PWR fuel rods

    SciTech Connect (OSTI)

    Adams, W.M.; Fisher, H.D.; Litke, H.J.; Mordarski, W.J.

    1985-04-01

    This report presents the results from a review of interpellet-gap formation, ovality, creepdown and clad collapse data in modern PWR fuel rods. Conclusions are reached regarding the propensity of modern PWR fuel to form such gaps and to undergo clad collapse. CEPAN, a creep-collapse predictor approved by the NRC in 1976, has been reformulated to include the creep analysis of cladding with finite interpellet gaps. The basis for this reformulation is discussed in detail. The model previously used in the calculation of the augmentation factor, a peak linear heat rate penalty due to the presence of interpellet gaps within the fuel rod, has been modified to incorporate gap-formation statistics from modern fuel. Finnally, the benefits of the limited gap formation and the CEPAN reformulation for the licensing of modern PWR fuel rods are evaluated.

  11. Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

    DOE Patents [OSTI]

    Brehm, Jr., William F.; Colburn, Richard P.

    1982-01-01

    An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.

  12. Nuclear Fuel Cycle

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

    - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs Advanced Nuclear Energy Nuclear

  13. Nuclear Fuel Reprocessing

    SciTech Connect (OSTI)

    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.

  14. TEPP- Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    This scenario provides the planning instructions, guidance, and evaluation forms necessary to conduct an exercise involving a highway shipment of spent nuclear fuel.  This exercise manual is one in...

  15. Nuclear fuel element

    DOE Patents [OSTI]

    Zocher, Roy W. (Los Alamos, NM)

    1991-01-01

    A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

  16. High-temperature Chemical Compatibility of As-fabricated TRIGA Fuel and Type 304 Stainless Steel Cladding

    SciTech Connect (OSTI)

    Dennis D. Keiser, Jr.; Jan-Fong Jue; Eric Woolstenhulme; Kurt Terrani; Glenn A. Moore

    2012-09-01

    Chemical interaction between TRIGA fuel and Type-304 stainless steel cladding at relatively high temperatures is of interest from the point of view of understanding fuel behavior during different TRIGA reactor transient scenarios. Since TRIGA fuel comes into close contact with the cladding during irradiation, there is an opportunity for interdiffusion between the U in the fuel and the Fe in the cladding to form an interaction zone that contains U-Fe phases. Based on the equilibrium U-Fe phase diagram, a eutectic can develop at a composition between the U6Fe and UFe2 phases. This eutectic composition can become a liquid at around 725C. From the standpoint of safe operation of TRIGA fuel, it is of interest to develop better understanding of how a phase with this composition may develop in irradiated TRIGA fuel at relatively high temperatures. One technique for investigating the development of a eutectic phase at the fuel/cladding interface is to perform out-of-pile diffusion-couple experiments at relatively high temperatures. This information is most relevant for lightly irradiated fuel that just starts to touch the cladding due to fuel swelling. Similar testing using fuel irradiated to different fission densities should be tested in a similar fashion to generate data more relevant to more heavily irradiated fuel. This report describes the results for TRIGA fuel/Type-304 stainless steel diffusion couples that were annealed for one hour at 730 and 800C. Scanning electron microscopy with energy- and wavelength-dispersive spectroscopy was employed to characterize the fuel/cladding interface for each diffusion couple to look for evidence of any chemical interaction. Overall, negligible fuel/cladding interaction was observed for each diffusion couple.

  17. Nuclear fuel pin scanner

    DOE Patents [OSTI]

    Bramblett, Richard L. (Friendswood, TX); Preskitt, Charles A. (La Jolla, CA)

    1987-03-03

    Systems and methods for inspection of nuclear fuel pins to determine fiss loading and uniformity. The system includes infeed mechanisms which stockpile, identify and install nuclear fuel pins into an irradiator. The irradiator provides extended activation times using an approximately cylindrical arrangement of numerous fuel pins. The fuel pins can be arranged in a magazine which is rotated about a longitudinal axis of rotation. A source of activating radiation is positioned equidistant from the fuel pins along the longitudinal axis of rotation. The source of activating radiation is preferably oscillated along the axis to uniformly activate the fuel pins. A detector is provided downstream of the irradiator. The detector uses a plurality of detector elements arranged in an axial array. Each detector element inspects a segment of the fuel pin. The activated fuel pin being inspected in the detector is oscillated repeatedly over a distance equal to the spacing between adjacent detector elements, thereby multiplying the effective time available for detecting radiation emissions from the activated fuel pin.

  18. An evaluation of the nuclear fuel performance code BISON

    SciTech Connect (OSTI)

    Perez, D. M.; Williamson, R. L.; Novascone, S. R.; Larson, T. K.; Hales, J. D.; Spencer, B. W.; Pastore, G.

    2013-07-01

    BISON is a modern finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009. The code is applicable to both steady and transient fuel behavior and is used to analyze either 2D axisymmetric or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods and other well known fuel performance codes. Results from several assessment cases are reported, with emphasis on fuel centerline temperatures at various stages of fuel life, fission gas release, and clad deformation during pellet clad mechanical interaction (PCMI). BISON comparisons to fuel centerline temperature measurements are very good at beginning of life and reasonable at high burnup. Although limited to date, fission gas release comparisons are very good. Comparisons of rod diameter following significant power ramping are also good and demonstrate BISON's unique ability to model discrete pellet behavior and accurately predict clad ridging from PCMI. (authors)

  19. Nuclear Fuels | Department of Energy

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

    Nuclear Fuels Nuclear Fuels A reactor's ability to produce power efficiently is significantly affected by the composition and configuration of its fuel system. A nuclear fuel assembly consists of hundreds of thousands of uranium pellets, stacked and encapsulated within tubes called fuel rods or fuel pins which are then bundled together in various geometric arrangements. There are many design considerations for the material composition and geometric configuration of the various components

  20. Model for Simulation of Hydride Precipitation in Zr-Based Used Fuel Claddings: A Status Report on Current Model Capabilities

    Broader source: Energy.gov [DOE]

    The report demonstrates a meso-scale, microstructural evolution model for simulation of zirconium hydride precipitation in the cladding of used fuels during long-term dry-storage.

  1. Nuclear Fuel Cycle

    SciTech Connect (OSTI)

    Dale, Deborah J.

    2014-10-28

    These slides will be presented at the training course “International Training Course on Implementing State Systems of Accounting for and Control (SSAC) of Nuclear Material for States with Small Quantity Protocols (SQP),” on November 3-7, 2014 in Santa Fe, New Mexico. The slides provide a basic overview of the Nuclear Fuel Cycle. This is a joint training course provided by NNSA and IAEA.

  2. Nuclear Fuel Cycle Options Catalog

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

    Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management ...

  3. Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

    SciTech Connect (OSTI)

    Jason Hales; Various

    2014-06-01

    The US DOEs Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.

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

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    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.

  5. Nuclear fuel electrorefiner

    DOE Patents [OSTI]

    Ahluwalia, Rajesh K.; Hua, Thanh Q.

    2004-02-10

    The present invention relates to a nuclear fuel electrorefiner having a vessel containing a molten electrolyte pool floating on top of a cadmium pool. An anodic fuel dissolution basket and a high-efficiency cathode are suspended in the molten electrolyte pool. A shroud surrounds the fuel dissolution basket and the shroud is positioned so as to separate the electrolyte pool into an isolated electrolyte pool within the shroud and a bulk electrolyte pool outside the shroud. In operation, unwanted noble-metal fission products migrate downward into the cadmium pool and form precipitates where they are removed by a filter and separator assembly. Uranium values are transported by the cadmium pool from the isolated electrolyte pool to the bulk electrolyte pool, and then pass to the high-efficiency cathode where they are electrolytically deposited thereto.

  6. Analytical model for transient gas flow in nuclear fuel rods. [PWR; BWR

    SciTech Connect (OSTI)

    Rowe, D.S.; Oehlberg, R.N.

    1981-08-01

    An analytical model for calculating gas flow and pressure inside a nuclear fuel rod is presented. Such a model is required to calculate the pressure loading of cladding during ballooning that could occur for postulated reactor accidents. The mathematical model uses a porous media (permeability) concept to define the resistance to gas flow along the fuel rod. 7 refs.

  7. Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

    SciTech Connect (OSTI)

    Jan-Fong Jue; Dennis D. Keiser, Jr.; Cynthia R. Breckenridge; Glenn A. Moore; Mitchell K. Meyer

    2014-05-01

    Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative (GTRI) is developing an advanced monolithic fuel to convert US high performance research reactors to low-enriched uranium. Hot-isostatic-press bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the UMo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between fuel meat, cladding, and diffusion barrier, as well as U10Mo fuel meat and Al6061 cladding were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are A typical Zr diffusion barrier of thickness 25 m Transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 m Chemical banding, in some areas more than 100 m in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 713 wt% Decomposed areas containing plate-shaped low-Mo phase A typical Zr/cladding interaction layer of thickness 1-2 m A visible UZr2 bearing layer of thickness 1-2 m Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the UMo matrix No excessive interaction between cladding and the uncoated fuel edge Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. Some of these attributes might be critical to the irradiation performance of monolithic U-10Mo nuclear fuel. There are several issues or concerns that warrant more detailed study, such as precipitation along cladding-to-cladding bond line, chemical banding, uncovered fuel-zone edge, and interaction layer between UMo fuel meat and zirconium. Future post-irradiation examination results will focus, among other things, on identifying in-reactor failure mechanisms and, eventually, directing further fresh fuel characterization efforts.

  8. Swelling-resistant nuclear fuel

    DOE Patents [OSTI]

    Arsenlis, Athanasios; Satcher, Jr., Joe; Kucheyev, Sergei O.

    2011-12-27

    A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

  9. Fuel pin

    DOE Patents [OSTI]

    Christiansen, David W. (Kennewick, WA); Karnesky, Richard A. (Richland, WA); Leggett, Robert D. (Richland, WA); Baker, Ronald B. (Richland, WA)

    1989-01-01

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  10. Fuel pin

    DOE Patents [OSTI]

    Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

    1987-11-24

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  11. Experimental study of the deformation of Zircaloy PWR fuel rod cladding under mainly convective cooling

    SciTech Connect (OSTI)

    Hindle, E.D.; Mann, C.A.

    1982-01-01

    Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 915 degree C in flowing steam at atmospheric pressure. Internal test pressures were in the range 0.69 to 11.0 MPa. The length of cladding strained 33 percent or more was greatest (about 20 times the original diameter) when the initial pressure was 1.38/plus or minus/0.17MPa. This results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilizing the deformation or partial superplastic deformation, or both. For adjacent rods in a fuel assembly not to touch at any temperature, the pressure would have to be less than about 1 MPa. These results are compared with those form multirod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behavior of fuel elements in a loss-of-coolant accident are outlined. 37 refs.

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

    SciTech Connect (OSTI)

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

    2012-10-01

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

  13. Disposition of ORNL's Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Turner, D. W.; DeMonia, B. C.; Horton, L. L.

    2002-02-26

    This paper describes the process of retrieving, repackaging, and preparing Oak Ridge spent nuclear fuel (SNF) for off-site disposition. The objective of the Oak Ridge SNF Project is to safely, reliably, and efficiently manage SNF that is stored on the Oak Ridge Reservation until it can be shipped off-site. The project required development of several unique processes and the design and fabrication of special equipment to enable the successful retrieval, transfer, and repackaging of Oak Ridge SNF. SNF was retrieved and transferred to a hot cell for repackaging. After retrieval of SNF packages, the storage positions were decontaminated and stainless steel liners were installed to resolve the vulnerability of water infiltration. Each repackaged SNF canister has been transferred from the hot cell back to dry storage until off-site shipments can be made. Three shipments of aluminum-clad SNF were made to the Savannah River Site (SRS), and five shipments of non-aluminum-clad SNF are planned to the Idaho National Engineering and Environmental Laboratory (INEEL). Through the integrated cooperation of several organizations including the U.S. Department of Energy (DOE), Bechtel Jacobs Company LLC (BJC), Oak Ridge National Laboratory (ORNL), and various subcontractors, preparations for the disposition of SNF in Oak Ridge have been performed in a safe and successful manner.

  14. BISON Enhanced with Improved Models for Cladding and Coolant Channels |

    Energy Savers [EERE]

    Department of Energy Enhanced with Improved Models for Cladding and Coolant Channels BISON Enhanced with Improved Models for Cladding and Coolant Channels January 29, 2013 - 10:54am Addthis Pin-scale Code Development Development of BISON for the engineering-scale simulation of nuclear fuel performance continued. Enhancements during this quarter include implementation of a nonlinear material model for Zircaloy cladding that simultaneously combines the phenomena of plasticity, thermal creep,

  15. Fuel subassembly leak test chamber for a nuclear reactor

    DOE Patents [OSTI]

    Divona, Charles J.

    1978-04-04

    A container with a valve at one end is inserted into a nuclear reactor coolant pool. Once in the pool, the valve is opened by a mechanical linkage. An individual fuel subassembly is lifted into the container by a gripper; the valve is then closed providing an isolated chamber for the subassembly. A vacuum is drawn on the chamber to encourage gaseous fission product leakage through any defects in the cladding of the fuel rods comprising the subassembly; this leakage may be detected by instrumentation, and the need for replacement of the assembly ascertained.

  16. Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

    SciTech Connect (OSTI)

    Powers, Jeffrey J.; George, Nathan; Maldonado, G. Ivan; Worrall, Andrew

    2015-08-28

    This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

  17. Advanced Nuclear Fuel Cycle Options

    SciTech Connect (OSTI)

    Roald Wigeland; Temitope Taiwo; Michael Todosow; William Halsey; Jess Gehin

    2010-06-01

    A systematic evaluation has been conducted of the potential for advanced nuclear fuel cycle strategies and options to address the issues ascribed to the use of nuclear power. Issues included nuclear waste management, proliferation risk, safety, security, economics and affordability, and sustainability. The two basic strategies, once-through and recycle, and the range of possibilities within each strategy, are considered for all aspects of the fuel cycle including options for nuclear material irradiation, separations if needed, and disposal. Options range from incremental changes to todays implementation to revolutionary concepts that would require the development of advanced nuclear technologies.

  18. Modeling and Simulation for Used Nuclear Fuel Loading and Effect of Hydriding during Transportation

    SciTech Connect (OSTI)

    Chakraborty, Pritam; Sabharwall, Piyush; Spears, Robert Edward; Coleman, Justin Leigh

    2015-09-30

    The United States nuclear contribution to the net energy production mix is approximately 20% per year. Meeting the demand every year adds to the spent nuclear fuel (SNF) inventory. The commercial nuclear fleet consists mainly of light water reactors (pressurized water reactors [PWRs] and boiling water reactors), with estimation of approximately 140,000 maximum transmission units (MTUs) of spent nuclear discharged from the operating reactor fleet [1]. The current average discharge burnup for PWRs is approximately 48 gigawatt-days (GWd)/MTU and for boiling water reactors is approximately 43 gigawatt-days GWd/MTU [2, 3], with a potential increase in the burnup looking into the future, which will lead to higher decay heat and potential changes in cladding thickness, higher hydrogen content in the cladding with increase in production of the fission gas, and hydride effect. Current fuels research and development examines various cladding materials such as ZIRLOTM and M5, and the accident-tolerant fuel program mainly focuses on other novel cladding materials such as SiC and FeCrAl as potential cladding materials for future.

  19. Determination of the bias in LOFT fuel peak cladding temperature data from the blowdown phase of large-break LOCA experiments

    SciTech Connect (OSTI)

    Berta, V.T.; Hanson, R.G.; Johnsen, G.W.; Schultz, R.R.

    1993-05-01

    Data from the Loss-of-Fluid Test (LOFT) Program help quantify the margin of safety inherent in pressurized water reactors during postulated loss-of-coolant accidents (LOCAs). As early as 1979, questions arose concerning the accuracy of LOFT fuel rod cladding temperature data during several large-break LOCA experiments. This report analyzes how well externally-mounted fuel rod cladding thermocouples in LOFT accurately reflected actual cladding surface temperature during large-break LOCA experiments. In particular, the validity of the apparent core-wide fuel rod cladding quench exhibited during blowdown in LOFT Experiments L2-2 and L2-3 is studied. Also addressed is the question of whether the externally-mounted thermocouples might have influenced cladding temperature. The analysis makes use of data and information from several sources, including later, similar LOFT Experiments in which fuel centerline temperature measurements were made, experiments in other facilities, and results from a detailed FRAP-T6 model of the LOFT fuel rod. The analysis shows that there can be a significant difference (referred to as bias) between the surface-mounted thermocouple reading and the actual cladding temperature, and that the magnitude of this bias depends on the rate of heat transfer between the fuel rod cladding and coolant. The results of the analysis demonstrate clearly that a core-wide cladding quench did occur in Experiments L2-2 and L2-3. Further, it is shown that, in terms of peak cladding temperature recording during LOFT large-break LOCA experiments, the mean bias is 11.4 {plus_minus} 16.2K (20.5 {plus_minus} 29.2{degrees} F). The best-estimate value of peak cladding temperature for LOFT LP-02-6 is 1,104.8 K. The best-estimate peak cladding temperature for LOFT LP-LB-1 is 1284.0 K.

  20. ATR LEU Monolithic Foil-Type Fuel with Integral Cladding Burnable Absorber Neutronics Performance Evaluation

    SciTech Connect (OSTI)

    Gray Chang

    2012-03-01

    The Advanced Test Reactor (ATR), currently operating in the United States, is used for material testing at very high neutron fluxes. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting HEU driven reactor cores to low-enriched uranium (LEU) cores. The burnable absorber - 10B, was added in the inner and outer plates to reduce the initial excess reactivity, and to improve the peak ratio of the inner/outer heat flux. The present work investigates the LEU Monolithic foil-type fuel with 10B Integral Cladding Burnable Absorber (ICBA) design and evaluates the subsequent neutronics operating effects of this proposed fuel designs. The proposed LEU fuel specification in this work is directly related to both the RERTR LEU Development Program and the Advanced Test Reactor (ATR) LEU Conversion Project at Idaho National Laboratory (INL).

  1. Nuclear Fuel Cycle & Vulnerabilities

    SciTech Connect (OSTI)

    Boyer, Brian D.

    2012-06-18

    The objective of safeguards is the timely detection of diversion of significant quantities of nuclear material from peaceful nuclear activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection. The safeguards system should be designed to provide credible assurances that there has been no diversion of declared nuclear material and no undeclared nuclear material and activities.

  2. Modeling the Nuclear Fuel Cycle

    SciTech Connect (OSTI)

    Jacob J. Jacobson; A. M. Yacout; G. E. Matthern; S. J. Piet; A. Moisseytsev

    2005-07-01

    The Advanced Fuel Cycle Initiative is developing a system dynamics model as part of their broad systems analysis of future nuclear energy in the United States. The model will be used to analyze and compare various proposed technology deployment scenarios. The model will also give a better understanding of the linkages between the various components of the nuclear fuel cycle that includes uranium resources, reactor number and mix, nuclear fuel type and waste management. Each of these components is tightly connected to the nuclear fuel cycle but usually analyzed in isolation of the other parts. This model will attempt to bridge these components into a single model for analysis. This work is part of a multi-national laboratory effort between Argonne National Laboratory, Idaho National Laboratory and United States Department of Energy. This paper summarizes the basics of the system dynamics model and looks at some results from the model.

  3. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

    Bassett, C.H.

    1961-05-16

    A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

  4. Recycling of nuclear spent fuel with AIROX processing

    SciTech Connect (OSTI)

    Majumdar, D.; Jahshan, S.N.; Allison, C.M.; Kuan, P.; Thomas, T.R.

    1992-12-01

    This report examines the concept of recycling light water reactor (LWR) fuel through use of a dry-processing technique known as the AIROX (Atomics International Reduction Oxidation) process. In this concept, the volatiles and the cladding from spent LWR fuel are separated from the fuel by the AIROX process. The fuel is then reenriched and made into new fuel pins with new cladding. The feasibility of the concept is studied from a technical and high level waste minimization perspective.

  5. NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 05 NUCLEAR FUELS...

    Office of Scientific and Technical Information (OSTI)

    Title list of documents made publicly available, January 1-31, 1998 NONE 21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 05 NUCLEAR FUELS; BIBLIOGRAPHIES; NUCLEAR POWER PLANTS;...

  6. NUCLEAR REACTOR FUEL SYSTEMS

    DOE Patents [OSTI]

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  7. Nuclear fuel recycling in 4 minutes | Argonne National Laboratory

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

    Nuclear fuel recycling in 4 minutes Share Topic Energy Energy sources Nuclear energy Nuclear fuel cycle Reactors

  8. Spent Nuclear Fuel Project Safety Management Plan

    SciTech Connect (OSTI)

    Garvin, L.J.

    1996-02-01

    The Spent Nuclear Fuel Project Safety Management Plan describes the new nuclear facility regulatory requirements basis for the Spemt Nuclear Fuel (SNF) Project and establishes the plan to achieve compliance with this basis at the new SNF Project facilities.

  9. Laser shockwave technique for characterization of nuclear fuel plate interfaces

    SciTech Connect (OSTI)

    Perton, M.; Levesque, D.; Monchalin, J.-P.; Lord, M. [National Research Council Canada, 75 de Mortagne Blvd, Boucherville, Quebec, J4B 6Y4 (Canada); Smith, J. A.; Rabin, B. H. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2013-01-25

    The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process. Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.

  10. Laser Shockwave Technique For Characterization Of Nuclear Fuel Plate Interfaces

    SciTech Connect (OSTI)

    James A. Smith; Barry H. Rabin; Mathieu Perton; Daniel Lvesque; Jean-Pierre Monchalin; Martin Lord

    2012-07-01

    The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process. Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.

  11. Nuclear Fuel Cycle Options Catalog

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

    Options Catalog - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs Advanced Nuclear

  12. Nuclear Fuels: Promise and Limitations

    SciTech Connect (OSTI)

    Harold F. McFarlane

    2012-03-01

    From 1950 through 1980, scientists, engineers and national leaders confidently predicted an early twenty-first century where fast breeder reactors and commercial nuclear fuel reprocessing were commonplace. Such a scenario seemed necessary for a world with the more than 1000 GWe of nuclear energy needed to meet such an ever-increasing thirst for energy. Thirty years later uranium reserves are increasing on pace with consumption, the growth of nuclear power has been slowed, commercial breeder reactors have yet to enter the marketplace, and less than a handful of commercial reprocessing plants operate. As Nobel Laureate Niels Bohr famously said, Prediction is very difficult, especially if its about the future. The programme for IChemEs 2012 conference on the nuclear fuel cycle features a graphic of an idealized nuclear fuel cycle that symbolizes the quest for a closed nuclear fuel cycle featuring careful husbanding of precious resources while minimizing the waste footprint. Progress toward achieving this ideal has been disrupted by technology innovations in the mining and petrochemical industries, as well as within the nuclear industry.

  13. Clad Degradation - FEPs Screening Arguments

    SciTech Connect (OSTI)

    E. Siegmann

    2004-03-17

    The purpose of this report is to document the screening of the cladding degradation features, events, and processes (FEPs) for commercial spent nuclear fuel (CSNF). This report also addresses the effect of some FEPs on both the cladding and the CSNF, DSNF, and HLW waste forms where it was considered appropriate to address the effects on both materials together. This report summarizes the work of others to screen clad degradation FEPs in a manner consistent with, and used in, the Total System Performance Assessment-License Application (TSPA-LA). This document was prepared according to ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA'' (BSC 2004a [DIRS 167796]).

  14. Proliferation Resistant Nuclear Reactor Fuel

    SciTech Connect (OSTI)

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and we posit that the exploration, development, and implementation of intrinsic mechanisms such as discussed here are part of a balanced approach aimed at preventing the misuse of nuclear material for nuclear-energy applications.

  15. Nuclear Fuel Cycle | Department of Energy

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

    Cycle Nuclear Fuel Cycle This is an illustration of a nuclear fuel cycle that shows the required steps to process natural uranium from ore for preparation for fuel to be loaded in nuclear reactors. This is an illustration of a nuclear fuel cycle that shows the required steps to process natural uranium from ore for preparation for fuel to be loaded in nuclear reactors. The mission of NE-54 is primarily focused on activities related to the front end of the nuclear fuel cycle which includes mining,

  16. Nuclear reactor composite fuel assembly

    DOE Patents [OSTI]

    Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  17. Categorization of Used Nuclear Fuel Inventory in Support of a...

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

    Categorization of Used Nuclear Fuel Inventory in Support of a Comprehensive National Nuclear Fuel Cycle Strategy Categorization of Used Nuclear Fuel Inventory in Support of a ...

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

    DOE Patents [OSTI]

    Forsberg, C.W.

    1992-03-24

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

  19. A new paradigm: near-complete recycling of spent fuel - A path to sustainable nuclear energy

    SciTech Connect (OSTI)

    Del Cul, Guillermo D.; Spencer, Barry B.; Collins, Emory D.

    2007-07-01

    Recent studies indicate that maximized recycling, where more than 95% of the components of spent nuclear fuel are reused, can be economically justified and can reduce the mass of waste products by a substantial amount. The potentially removable and reusable components include the uranium, zirconium from the cladding, structural hardware, certain noble metal fission products, and the transuranic radionuclides. The approach to maximizing recycle and minimizing emissions and wastes should improve public acceptance of nuclear energy. (authors)

  20. World nuclear fuel cycle requirements 1991

    SciTech Connect (OSTI)

    Not Available

    1991-10-10

    The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs.

  1. Fully ceramic nuclear fuel and related methods

    DOE Patents [OSTI]

    Venneri, Francesco; Katoh, Yutai; Snead, Lance Lewis

    2016-03-29

    Various embodiments of a nuclear fuel for use in various types of nuclear reactors and/or waste disposal systems are disclosed. One exemplary embodiment of a nuclear fuel may include a fuel element having a plurality of tristructural-isotropic fuel particles embedded in a silicon carbide matrix. An exemplary method of manufacturing a nuclear fuel is also disclosed. The method may include providing a plurality of tristructural-isotropic fuel particles, mixing the plurality of tristructural-isotropic fuel particles with silicon carbide powder to form a precursor mixture, and compacting the precursor mixture at a predetermined pressure and temperature.

  2. Compositions and methods for treating nuclear fuel

    DOE Patents [OSTI]

    Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

    2014-01-28

    Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

  3. Compositions and methods for treating nuclear fuel

    DOE Patents [OSTI]

    Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

    2013-08-13

    Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

  4. Multidimensional Multiphysics Simulation of Nuclear Fuel Behavior

    SciTech Connect (OSTI)

    R. L. Williamson; J. D. Hales; S. R. Novascone; M. R. Tonks; D. R. Gaston; C. J. Permann; D. Andrs; R. C. Martineau

    2012-04-01

    Important aspects of fuel rod behavior, for example pellet-clad mechanical interaction (PCMI), fuel fracture, oxide formation, non- axisymmetric cooling, and response to fuel manufacturing defects, are inherently multidimensional in addition to being complicated multiphysics problems. Many current modeling tools are strictly 2D axisymmetric or even 1.5D. This paper outlines the capabilities of a new fuel modeling tool able to analyze either 2D axisymmetric or fully 3D models. These capabilities include temperature-dependent thermal conductivity of fuel; swelling and densification; fuel creep; pellet fracture; fission gas release; cladding creep; irradiation growth; and gap mechanics (contact and gap heat transfer). The need for multiphysics, multidimensional modeling is then demonstrated through a discussion of results for a set of example problems. The first, a 10-pellet rodlet, demonstrates the viability of the solution method employed. This example highlights the effect of our smeared cracking model and also shows the multidimensional nature of discrete fuel pellet modeling. The second example relies on our multidimensional, multiphysics approach to analyze a missing pellet surface problem. The next example is the analysis of cesium diffusion in a TRISO fuel particle with defects. As a final example, we show a lower-length-scale simulation coupled to a continuum-scale simulation.

  5. Lanthanides in Metallic Nuclear Fuels: Their Behavior and Methods for Their Control

    SciTech Connect (OSTI)

    Robert D. Mariani; Douglas L. Porter; Thomas P. O'Holleran; Steven L. Hayes; J. Rory Kennedy

    2011-12-01

    The thermodynamic and experimental basis is given for using dopant additives to bind lanthanides as intermetallic compounds in metallic nuclear fuels. Lanthanide fission products are a major factor in limiting the lifetime of the fuel, because they migrate to the fuel slug peripheral surface where they participate in fuel-cladding chemical interactions (FCCI) with the steel cladding. Lanthanide carryover in recycled metal fuels can accelerate FCCI, as recycled lanthanides would likely segregate from the fuel phase, putting the lanthanides in prompt contact with the cladding. In out-of-pile tests we examined the use of Pd for binding the lanthanides, with Pd selected because of its known metallurgical properties in fuel related systems and because of its known behavior in irradiated EBR-II fuels. Initial results confirmed that palladium may be expected to mitigate FCCI arising from lanthanides, and it has been recommended for in-pile tests. We also evaluated transport phenomena responsible for lanthanide migration, and identified liquid-like behaviors as being dominant. Liquid-like behaviors include transport with liquid metals, liquid metal solutions, and rapid surface transport of alloys/metals near their melting temperatures. The analysis led to establishing general criteria for selecting alternate dopant additives, and identifying Sn, Sb, and Te as alternates for further testing.

  6. Fast, Quantitative, and Nondestructive Evaluation on Hydrided LWR Fuel Cladding by Small Angle Incoherent Neutron Scattering of Hydrogen

    SciTech Connect (OSTI)

    Yan, Yong; Qian, Shuo; Littrell, Ken; Parish, Chad M; Plummer, Lee K

    2015-01-01

    A non-destructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor (LWR) fuel cladding was developed. Zircaloy-4 cladding used in commercial LWRs was used to produce hydrided specimens. The hydriding apparatus consists of a closed stainless steel vessel that contains Zr alloy specimens and hydrogen gas. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentration were selected for the neutron study. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall. Small angle neutron incoherent scattering was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Our study demonstrates that the hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes by this nondestructive method over a wide range of hydrogen concentrations from a very small amount ( 20 ppm) to over 1000 ppm. The hydrogen distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor determined by a calibration process using standard, destructive direct chemical analysis methods on the specimens. This scale factor can be used in future tests with unknown hydrogen concentrations, thus providing a nondestructive method for absolute hydrogen concentration determination.

  7. Literature search for the non-aqueous separation of zinc from fuel rod cladding. [After dissolution in liquid metal

    SciTech Connect (OSTI)

    Sandvig, R. L.; Dyer, S. J.; Lambert, G. A.; Baldwin, C. E.

    1980-06-21

    This report reviews the literature of processes for the nonaqueous separation of zinc from dissolved fuel assembly cladding. The processes considered were distillation, pyrochemical processing, and electrorefining. The last two techniques were only qualitatively surveyed while the first, distillation, was surveyed in detail. A survey of available literature from 1908 through 1978 on the distillation of zinc was performed. The literature search indicated that a zinc recovery rate in excess of 95% is possible; however, technical problems exist because of the high temperatures required and the corrosive nature of liquid zinc. The report includes a bibliography of the surveyed literature and a computer simulation of vapor pressures in binary systems. 129 references.

  8. Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

    SciTech Connect (OSTI)

    N /A

    2000-04-14

    The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign and domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.

  9. Nuclear Fuel Cycle | Department of Energy

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

    Fuel Cycle Nuclear Fuel Cycle GC-52 provides legal advice to DOE regarding research and development of nuclear fuel and waste management technologies that meet the nation's energy supply, environmental, and energy security needs. GC-52 also advises DOE on issues involving support for international fuel cycle initiatives aimed at advancing a common vision of the necessity of the expansion of nuclear energy for peaceful purposes worldwide in a safe and secure manner. In addition, GC-52 provides

  10. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding. [PWR

    SciTech Connect (OSTI)

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed.

  11. Annealing induced interfacial layers in niobium-clad stainless steel developed as a bipolar plate material for polymer electrolyte membrane fuel cell stacks

    SciTech Connect (OSTI)

    Hong, Sung Tae; Weil, K. Scott; Choi, Jung-Pyung; Bae, In-Tae; Pan, Jwo

    2010-05-01

    Niobium (Nb)-clad 304L stainless steel (SS) manufactured by cold rolling is currently under consideration for use as a bipolar plate material in polymer electrolyte membrane fuel cell (PEMFC) stacks. To make the fabrication of bipolar plates using the Nb-clad SS feasible, annealing may be necessary for the Nb-clad SS to reduce the springback induced by cold rolling. However, the annealing can develop an interfacial layer between the Nb cladding and the SS core and the interfacial layer plays a key role in the failure of the Nb-clad SS as reported earlier [JPS our work]. In this investigation, the Nb-clad SS specimens in as-rolled condition were annealed at different combinations of temperature and time. Based on the results of scanning electron microscope (SEM) analysis, an annealing process map for the Nb-clad SS was obtained. The results of SEM analysis and Transmission Electron Microscope (TEM) analysis also suggest that different interfacial layers occurred based on the given annealing conditions.

  12. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    SciTech Connect (OSTI)

    Johnson, G.L.

    1988-09-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97{degree}C and whether the cladding of the stored spent fuel ever exceeds 350{degree}C. Limiting the borehole to temperatures of 97{degree}C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350{degree}C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97{degree}C for the full 1000-yr analysis period.

  13. Corrosion and hydriding performance evaluation of three Zircaloy-2 clad fuel assemblies after continuous exposure in PWR cores 1 and 2 at Shippingport, PA. Addendum. LWBR Development Program

    SciTech Connect (OSTI)

    Hillner, E.

    1983-12-01

    The cladding from one additional Zircaloy-2 clad fuel rod from the pressurized water reactor at Shippingport, Pa. was destructively examined for corrosion film thickness and hydrogen accumulation. These additional examinations were conducted primarily to determine whether or not the hydrogen pickup ratio (..delta..H/..delta..O) increased with increasing neutron exposure, as had been suggested by the results from earlier studies on these fuel rods. The current results indicate that the hydrogen pickup ratio for Zircaloy-2 does not change with increasing neutron exposure and suggest that some of the earlier reported data may be anomolous.

  14. Nuclear core and fuel assemblies

    DOE Patents [OSTI]

    Downs, Robert E. (Monroeville, PA)

    1981-01-01

    A fast flux nuclear core of a plurality of rodded, open-lattice assemblies having a rod pattern rotated relative to a rod support structure pattern. Elongated fuel rods are oriented on a triangular array and laterally supported by grid structures positioned along the length of the assembly. Initial inter-assembly contact is through strongbacks at the corners of the support pattern and peripheral fuel rods between adjacent assemblies are nested so as to maintain a triangular pitch across a clearance gap between the other portions of adjacent assemblies. The rod pattern is rotated relative to the strongback support pattern by an angle .alpha. equal to sin .sup.-1 (p/2c), where p is the intra-assembly rod pitch and c is the center-to-center spacing among adjacent assemblies.

  15. Anticipating Potential Waste Acceptance Criteria for Defense Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Rechard, R.P.; Lord, M.E.; Stockman, C.T.; McCurley, R.D.

    1997-12-31

    The Office of Environmental Management of the U.S. Department of Energy is responsible for the safe management and disposal of DOE owned defense spent nuclear fuel and high level waste (DSNF/DHLW). A desirable option, direct disposal of the waste in the potential repository at Yucca Mountain, depends on the final waste acceptance criteria, which will be set by DOE`s Office of Civilian Radioactive Waste Management (OCRWM). However, evolving regulations make it difficult to determine what the final acceptance criteria will be. A method of anticipating waste acceptance criteria is to gain an understanding of the DOE owned waste types and their behavior in a disposal system through a performance assessment and contrast such behavior with characteristics of commercial spent fuel. Preliminary results from such an analysis indicate that releases of 99Tc and 237Np from commercial spent fuel exceed those of the DSNF/DHLW; thus, if commercial spent fuel can meet the waste acceptance criteria, then DSNF can also meet the criteria. In large part, these results are caused by the small percentage of total activity of the DSNF in the repository (1.5%) and regulatory mass (4%), and also because commercial fuel cladding was assumed to provide no protection.

  16. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    SciTech Connect (OSTI)

    R. Schreiner

    2004-10-21

    The purpose of this report is to evaluate and document the screening of the clad degradation features, events, and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment-License Application (TSPA-LA). This report also addresses the effect of certain FEPs on both the cladding and the commercial spent nuclear fuel (CSNF), DOE-owned spent nuclear fuel (DSNF), and defense high-level waste (DHLW) waste forms, as appropriate to address the effects on multiple materials and both components (FEPs 2.1.09.09.0A, 2.1.09.11.0A, 2.1.11.05.0A, 2.1.12.02.0A, and 2.1.12.03.0A). These FEPs are expected to affect the repository performance during the postclosure regulatory period of 10,000 years after permanent closure. Table 1-1 provides the list of cladding FEPs, including their screening decisions (include or exclude). The primary purpose of this report is to identify and document the analysis, screening decision, and TSPA-LA disposition (for included FEPs) or screening argument (for excluded FEPs) for these FEPs related to clad degradation. In some cases, where a FEP covers multiple technical areas and is shared with other FEP reports, this report may provide only a partial technical basis for the screening of the FEP. The full technical basis for shared FEPs is addressed collectively by the sharing FEP reports. The screening decisions and associated TSPA-LA dispositions or screening arguments from all of the FEP reports are cataloged in a project-specific FEPs database.

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

    SciTech Connect (OSTI)

    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 discharge reuse. The EM2 waste disposal profile is effectively only fission products, which reduces the mass (about 3% vs LWR), average half life, heat and long term radio-toxicity of the disposal. Widespread implementation of EM2 fuel cycle is highly significant as it would increase world energy reserves; the remaining energy in U.S. LWR SNF alone exceeds that in the U.S. natural gas reserves. Unlike many LWR SNF disposition concepts, the EM2 fuel cycle conversion of SNF produces energy and associated revenue such that the overall project is cost effective. By providing conversion of SNF to fission products the fuel cycle is closed and a non-repository LWR SNF disposition path is created and overall repository requirements are significantly reduced. (authors)

  18. Supplemental information for a notice of construction for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    SciTech Connect (OSTI)

    Not Available

    1989-08-01

    This ''Notice of Construction'' has been submitted by the US Department of Energy-Richland Operations Office (P.O. Box 550, Richland, Washington 99352), pursuant to WAC 402-80-070, for three new sources of radionuclide emissions at the Hanford Site in Washington State (Figure 1). The three new sources, the Fueled Clad Fabrication System (FCFS) the Radioisotope Power Systems Facility (RPSF) and the Fuel Assembly Area (FAA) will be located in one facility, the Fuels and materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post- irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies to be used in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, to the extent possible, these systems will be dealt with separately. The FAA is a comparatively independent operation though it will share the FMEF complex.

  19. Prevention of significant deterioration permit application for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    SciTech Connect (OSTI)

    Not Available

    1989-08-01

    This New Source Review'' has been submitted by the US Department of Energy-Richland Operations Office (PO Box 550, Richland, Washington 99352), pursuant to WAC 173-403-050 and in compliance with the Department of Ecology Guide to Processing A Prevention Of Significant Deterioration (PSD) Permit'' for three new sources of radionuclide emissions at the Hanford Site in Washington State. The three new sources, the Fueled Clad Fabrication System (FCFS), the Radioisotope Power Systems Facility (RPSF), and the Fuel Assembly Area (FAA), will be located in one facility, the Fuels and Materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post-irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies for use in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, these systems will be dealt with separately to the extent possible. The FAA is a comparatively independent operation though it will share the FMEF complex.

  20. PACKAGE IMPACT MODELS AS A PRECURSOR TO CLADDING ANALYSIS

    SciTech Connect (OSTI)

    Klymyshyn, Nicholas A.; Adkins, Harold E.; Bajwa, Christopher S.; Piotter, Jason

    2013-02-01

    The evaluation of spent nuclear fuel storage casks and transportation packages under impact loading is an important safety topic that is reviewed as part of cask and package certification by the United States Nuclear Regulatory Commission. Explicit dynamic finite element models of full systems are increasingly common in industry for determining structural integrity during hypothetical drop accidents. Full cask and package model results are also used as the loading basis for single fuel pin impact models, which evaluate the response of fuel cladding under drop conditions. In this paper, a simplified package system is evaluated to illustrate several important structural dynamic phenomena, including the effect of gaps between components, the difference in local response at various points on a package during impact, and the effect of modeling various simplified representations of the basket and fuel assemblies. This paper focuses on the package impact analysis, and how loading conditions for a subsequent fuel assembly or fuel cladding analysis can be extracted.

  1. Reservoir Claddings

    SciTech Connect (OSTI)

    2009-05-14

    This information sheet explains how to properly decouple reservoir claddings from water sensitive materials of the wall assembly.

  2. Monitoring arrangement for vented nuclear fuel elements

    DOE Patents [OSTI]

    Campana, Robert J. (Solana Beach, CA)

    1981-01-01

    In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180.degree. rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements.

  3. Corrosion and hydriding performance evaluation of three zircaloy-2 clad fuel assemblies after continuous exposure in PWR cores 1 and 2, at Shippingport, PA

    SciTech Connect (OSTI)

    Hillner, E.

    1980-01-01

    Three original Zircaloy-2 clad blanket fuel bundles from the pressurized water reactor (PWR) at the Shippingport Atomic Power Station were discharged after continuous exposure during Cores 1 and 2. Detailed visual examination of these components after approx. 6300 calendar days of operation (51,140 EFPH) revealed only the anticipated uniform light gray (post-transition) corrosion products with no evidence of unexpected corrosion deterioration, fuel rod warpage, or other damage. All corrosion films were found to be tightly adherent to the underlying cladding. An extensive destructive examination of a selected fuel rod from each of three fuel bundles produced appreciably greater end-of-life rod average oxide film thickness when compared with corresponding values produced from a set of empirical equations generated from the out-of-pile (autoclave) testing of Zircaloy coupons.

  4. Spent Nuclear Fuel (SNF) Project Execution Plan

    SciTech Connect (OSTI)

    LEROY, P.G.

    2000-11-03

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

  5. Automated closure system for nuclear reactor fuel assemblies

    DOE Patents [OSTI]

    Christiansen, David W. (Kennewick, WA); Brown, William F. (West Richland, WA)

    1985-01-01

    A welder for automated closure of fuel pins by a pulsed magnetic process in which the open end of a length of cladding is positioned within a complementary tube surrounded by a pulsed magnetic welder. Seals are provided at each end of the tube, which can be evacuated or can receive tag gas for direct introduction to the cladding interior. Loading of magnetic rings and end caps is accomplished automatically in conjunction with the welding steps carried out within the tube.

  6. Simulated nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Berta, Victor T. (Idaho Falls, ID)

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  7. Simulated nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Berta, V.T.

    1993-04-06

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  8. fuel | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Jobs Working at NNSA Blog Home fuel fuel Cheaper catalyst may lower fuel costs for hydrogen-powered cars Sandia National Laboratories post-doctoral fellow Stan Chou demonstrates...

  9. World nuclear fuel cycle requirements 1990

    SciTech Connect (OSTI)

    Not Available

    1990-10-26

    This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under three nuclear supply scenarios. Two of these scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries with free market economies (FME countries). A No New Orders scenario is presented only for the United States. These nuclear supply scenarios are described in Commercial Nuclear Power 1990: Prospects for the United States and the World (DOE/EIA-0438(90)). This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the FME projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2030 for the Lower and Upper Reference cases and through 2040, the last year in which spent fuel is discharged, for the No New Orders case. These disaggregated projections are provided at the request of the Department of Energy's Office of Civilian Radioactive Waste Management.

  10. Energy Department Announces New Investment in Nuclear Fuel Storage...

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

    Investment in Nuclear Fuel Storage Research Energy Department Announces New Investment in Nuclear Fuel Storage Research April 16, 2013 - 12:19pm Addthis NEWS MEDIA CONTACT (202)...

  11. Department of Energy Awards $15 Million for Nuclear Fuel Cycle...

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

    Million for Nuclear Fuel Cycle Technology Research and Development Department of Energy Awards 15 Million for Nuclear Fuel Cycle Technology Research and Development August 1,...

  12. A Critical Step Toward Sustainable Nuclear Fuel Disposal | Department...

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

    A Critical Step Toward Sustainable Nuclear Fuel Disposal A Critical Step Toward Sustainable Nuclear Fuel Disposal January 26, 2012 - 2:30pm Addthis Secretary Chu Secretary Chu...

  13. Review of Used Nuclear Fuel Storage and Transportation Technical...

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

    Review of Used Nuclear Fuel Storage and Transportation Technical Gap Analysis Review of Used Nuclear Fuel Storage and Transportation Technical Gap Analysis While both wet and dry...

  14. Current Status of the Spent Nuclear Fuel Management Program in...

    Office of Scientific and Technical Information (OSTI)

    Current Status of the Spent Nuclear Fuel Management Program in the United States. Citation Details In-Document Search Title: Current Status of the Spent Nuclear Fuel Management...

  15. Nuclear Energy Systems Laboratory (NESL) / Transient Nuclear Fuels Testing

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

    Transient Nuclear Fuels Testing - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs

  16. Comparison of GAPCON-THERMAL-3 and FRAPCON-2 fuel-performance codes to in-reactor measurement of elastic cladding deformation. [PWR; BWR

    SciTech Connect (OSTI)

    Lanning, D.D.; Rausch, W.N.; Williford, R.E.

    1981-01-01

    A revision of the GAPCON-3 computer code became part of the NRC-sponsored FRAPCON-2 code. This paper presents a comparison of both codes to in-reactor data from IFA-508, a 3-rod test rig in the Halden Reactor, Norway, which features simultaneous measurements of fuel temperature, power, axial elongation, and diametral strain. The modeling revisions included putting all regions of the fuel in contact with cladding at all time, but assigning non-linear, spatially dependent, anisotropic elastic moduli to the fuel on an incremental load step basis. The moduli are functions of the local available void within the cladding. These concepts bring demonstrable improvement to the code predictions.

  17. Used Fuel Disposition Used Nuclear Fuel Storage and Transportation

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

    Storage and Transportation Overview Steve Marschman Field Demonstration Lead Idaho National Laboratory NEET ASI Review Meeting September 17, 2014 Used Fuel Disposition Today's Discussion n Our R&D Objectives n What Guides Our Work n FY14 and FY15 Work - Full-Scale High Burn-Up Demo - Experiments - Transportation - Analysis Used Fuel Disposition 3 Overall Objectives * Develop the technical bases to demonstrate the continued safe and secure storage of used nuclear fuel for extended

  18. W.T.; Rainey, R.H. 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS;...

    Office of Scientific and Technical Information (OSTI)

    thorium fuel reprocessing experience Brooksbank, R.E.; McDuffee, W.T.; Rainey, R.H. 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; NUCLEAR MATERIALS DIVERSION; SAFEGUARDS; SPENT FUELS;...

  19. Pyrochemical Treatment of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    K. M. Goff; K. L. Howden; G. M. Teske; T. A. Johnson

    2005-10-01

    Over the last 10 years, pyrochemical treatment of spent nuclear fuel has progressed from demonstration activities to engineering-scale production operations. As part of the Advanced Fuel Cycle Initiative within the U.S. Department of Energys Office of Nuclear Energy, Science and Technology, pyrochemical treatment operations are being performed as part of the treatment of fuel from the Experimental Breeder Reactor II at the Idaho National Laboratory. Integral to these treatment operations are research and development activities that are focused on scaling further the technology, developing and implementing process improvements, qualifying the resulting high-level waste forms, and demonstrating the overall pyrochemical fuel cycle.

  20. Method for shearing spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Weil, Bradley S.; Watson, Clyde D.

    1977-01-01

    A method is disclosed for shearing spent nuclear fuel assemblies of the type wherein a plurality of long metal tubes packed with ceramic fuel are supported in a spaced apart relationship within an outer metal shell or shroud which provides structural support to the assembly. Spent nuclear fuel assemblies are first compacted in a stepwise manner between specially designed gag-compactors and then sheared into short segments amenable to chemical processing by shear blades contoured to mate with the compacted surface of the fuel assembly.

  1. Annotated Bibliography for Drying Nuclear Fuel

    SciTech Connect (OSTI)

    Rebecca E. Smith

    2011-09-01

    Internationally, the nuclear industry is represented by both commercial utilities and research institutions. Over the past two decades many of these entities have had to relocate inventories of spent nuclear fuel from underwater storage to dry storage. These efforts were primarily prompted by two factors: insufficient storage capacity (potentially precipitated by an open-ended nuclear fuel cycle) or deteriorating quality of existing underwater facilities. The intent of developing this bibliography is to assess what issues associated with fuel drying have been identified, to consider where concerns have been satisfactorily addressed, and to recommend where additional research would offer the most value to the commercial industry and the U. S. Department of Energy.

  2. International Nuclear Fuel Cycle Fact Book

    SciTech Connect (OSTI)

    Leigh, I.W.; Patridge, M.D.

    1991-05-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECN/NEA activities reports; not reflect any one single source but frequently represent a consolidation/combination of information.

  3. Nuclear Fuels Storage & Transportation Planning Project Documents |

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

    Department of Energy Fuel Cycle Technologies » Nuclear Fuels Storage & Transportation Planning Project » Nuclear Fuels Storage & Transportation Planning Project Documents Nuclear Fuels Storage & Transportation Planning Project Documents October 1, 2014 Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites In January 2013, the Department of Energy issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. Among

  4. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    SciTech Connect (OSTI)

    Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy; Scaglione, John M.

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.

  5. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

    Bassett, C.H.

    1961-11-21

    A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

  6. Apparatus for shearing spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Weil, Bradley S.; Metz, III, Curtis F.

    1980-01-01

    A method and apparatus are described for shearing spent nuclear fuel assemblies of the type comprising an array of fuel pins disposed within an outer metal shell or shroud. A spent fuel assembly is first compacted in a known manner and then incrementally sheared using fixed and movable shear blades having matched laterally projecting teeth which slidably intermesh to provide the desired shearing action. Incremental advancement of the fuel assembly after each shear cycle is limited to a distance corresponding to the lateral projection of the teeth to ensure fuel assembly breakup into small uniform segments which are amenable to remote chemical processing.

  7. Electrochemical Processing of Used Nuclear Fuel

    SciTech Connect (OSTI)

    K. M. Goff; J. C. Wass; G. M. Teske

    2011-08-01

    As part of the Department of Energys Fuel Cycle Research and Development Program an electrochemical technology employing molten salts is being developed for recycle of metallic fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. This technology has been deployed for treatment of used fuel from the Experimental Breeder Reactor II (EBR-II) in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory. This process is based on dry (non-aqueous) technologies that have been developed and demonstrated since the 1960s. These technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including preparation of associated high-level waste forms.

  8. Composite construction for nuclear fuel containers

    DOE Patents [OSTI]

    Cheng, B. C.; Rosenbaum, H. S.; Armijo, J. S.

    1987-04-21

    Disclosed is an improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof. 1 fig.

  9. Composite construction for nuclear fuel containers

    DOE Patents [OSTI]

    Cheng, Bo-Ching; Rosenbaum, Herman S.; Armijo, Joseph S.

    1987-01-01

    An improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof.

  10. Nuclear Fuel Facts: Uranium | Department of Energy

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

    Facts: Uranium Nuclear Fuel Facts: Uranium Nuclear Fuel Facts: Uranium Uranium is a silvery-white metallic chemical element in the periodic table, with atomic number 92. It is assigned the chemical symbol U. A uranium atom has 92 protons and 92 electrons, of which 6 are valence electrons. Uranium has the highest atomic weight (19 kg m) of all naturally occurring elements. Uranium occurs naturally in low concentrations in soil, rock and water, and is commercially extracted from uranium-bearing

  11. High Temperature and Pressure Steam-H2 Interaction with Candidate Advanced LWR Fuel Claddings

    SciTech Connect (OSTI)

    Pint, Bruce A

    2012-08-01

    This report summarizes the work completed to evaluate cladding materials that could serve as improvements to Zircaloy in terms of accident tolerance. This testing involved oxidation resistance to steam or H{sub 2}-50% steam environments at 800-1350 C at 1-20 bar for short times. A selection of conventional alloys, SiC-based ceramics and model alloys were used to explore a wide range of materials options and provide guidance for future materials development work. Typically, the SiC-based ceramic materials, alumina-forming alloys and Fe-Cr alloys with {ge}25% Cr showed the best potential for oxidation resistance at {ge}1200 C. At 1350 C, FeCrAl alloys and SiC remained oxidation resistant in steam. Conventional austenitic steels do not have sufficient oxidation resistance with only {approx}18Cr-10Ni. Higher alloyed type 310 stainless steel is protective but Ni is not a desirable alloy addition for this application and high Cr contents raise concern about {alpha}{prime} formation. Higher pressures (up to 20.7 bar) and H{sub 2} additions appeared to have a limited effect on the oxidation behavior of the most oxidation resistant alloys but higher pressures accelerated the maximum metal loss for less oxidation resistant steels and less metal loss was observed in a H{sub 2}-50%H{sub 2}O environment at 10.3 bar. As some of the results regarding low-alloyed FeCrAl and Fe-Cr alloys were unexpected, further work is needed to fundamentally understand the minimum Cr and Al alloy contents needed for protective behavior in these environments in order to assist in alloy selection and guide alloy development.

  12. Fresh nuclear fuel measurements at Ukrainian nuclear power plants

    SciTech Connect (OSTI)

    Kuzminski, Jozef; Ewing, Tom; Dickman, Debbie; Gavrilyuk, Victor; Drapey, Sergey; Kirischuk, Vladimir; Strilchuk, Nikolay

    2009-01-01

    In 2005, the Provisions on Nuclear Material Measurement System was enacted in Ukraine as an important regulatory driver to support international obligations in nuclear safeguards and nonproliferation. It defines key provisions and requirements for material measurement and measurement control programs to ensure the quality and reliability of measurement data within the framework of the State MC&A System. Implementing the Provisions requires establishing a number of measurement techniques for both fresh and spent nuclear fuel for various types of Ukrainian reactors. Our first efforts focused on measurements of fresh nuclear fuel from a WWR-1000 power reactor.

  13. Dry Processing of Used Nuclear Fuel

    SciTech Connect (OSTI)

    K. M. Goff; M. F. Simpson

    2009-09-01

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

  14. DEMONSTRATION OF LONG-TERM STORAGE CAPABILITY FOR SPENT NUCLEAR FUEL IN L BASIN

    SciTech Connect (OSTI)

    Sindelar, R.; Deible, R.

    2011-04-27

    The U.S. Department of Energy decisions for the ultimate disposition of its inventory of used nuclear fuel presently in, and to be received and stored in, the L Basin at the Savannah River Site, and schedule for project execution have not been established. A logical decision timeframe for the DOE is following the review of the overall options for fuel management and disposition by the Blue Ribbon Commission on America's Nuclear Future (BRC). The focus of the BRC review is commercial fuel; however, the BRC has included the DOE fuel inventory in their review. Even though the final report by the BRC to the U.S. Department of Energy is expected in January 2012, no timetable has been established for decisions by the U.S. Department of Energy on alternatives selection. Furthermore, with the imminent lay-up and potential closure of H-canyon, no ready path for fuel disposition would be available, and new technologies and/or facilities would need to be established. The fuel inventory in wet storage in the 3.375 million gallon L Basin is primarily aluminum-clad, aluminum-based fuel of the Materials Test Reactor equivalent design. An inventory of non-aluminum-clad fuel of various designs is also stored in L Basin. Safe storage of fuel in wet storage mandates several high-level 'safety functions' that would be provided by the Structures, Systems, and Components (SSCs) of the storage system. A large inventory of aluminum-clad, aluminum-based spent nuclear fuel, and other nonaluminum fuel owned by the U.S. Department of Energy is in wet storage in L Basin at the Savannah River Site. An evaluation of the present condition of the fuel, and the Structures, Systems, or Components (SSCs) necessary for its wet storage, and the present programs and storage practices for fuel management have been performed. Activities necessary to validate the technical bases for, and verify the condition of the fuel and the SSCs under long-term wet storage have also been identified. The overall conclusion is that the fuel can be stored in L Basin, meeting general safety functions for fuel storage, for an additional 50 years and possibly beyond contingent upon continuation of existing fuel management activities and several augmented program activities. It is concluded that the technical bases and well-founded technologies have been established to store spent nuclear fuel in the L Basin. Methodologies to evaluate the fuel condition and characteristics, and systems to prepare fuel, isolate damaged fuel, and maintain water quality storage conditions have been established. Basin structural analyses have been performed against present NPH criteria. The aluminum fuel storage experience to date, supported by the understanding of the effects of environmental variables on materials performance, demonstrates that storage systems that minimize degradation and provide full retrievability of the fuel up to and greater than 50 additional years will require maintaining the present management programs, and with the recommended augmented/additional activities in this report.

  15. Reversible Bending Fatigue Test System for Investigating Vibration Integrity of Spent Nuclear Fuel during Transportation

    SciTech Connect (OSTI)

    Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom; Howard, Rob L; Flanagan, Michelle

    2013-01-01

    Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading during road or rail shipment. Oak Ridge National Laboratory (ORNL) has been developing testing capabilities that can be used to improve the understanding of the impacts on SNF integrity due to vibration loading, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety and security of spent nuclear fuel storage and transport operations. The ORNL developed test system can perform reversible-bending fatigue testing to evaluate both the static and dynamic mechanical response of SNF rods under simulated loads. The testing apparatus is also designed to meet the challenges of hot-cell operation, including remote installation and detachment of the SNF test specimen, in-situ test specimen deformation measurement, and implementation of a driving system suitable for use in a hot cell. The system contains a U-frame set-up equipped with uniquely designed grip rigs, to protect SNF rod and to ensure valid test results, and use of 3 specially designed LVDTs to obtain the in-situ curvature measurement. A variety of surrogate test rods have been used to develop and calibrate the test system as well as in performing a series of systematic cyclic fatigue tests. The surrogate rods include stainless steel (SS) cladding, SS cladding with cast epoxy, and SS cladding with alumina pellets inserts simulating fuel pellets. Testing to date has shown that the interface bonding between the SS cladding and the alumina pellets has a significant impact on the bending response of the test rods as well as their fatigue strength. The failure behaviors observed from tested surrogate rods provides a fundamental understanding of the underlying failure mechanisms of the SNF surrogate rod under vibration which has not been achieved previously. The newly developed device is scheduled to be installed in the hot-cell in summer 2013 to test high burnup SNF.

  16. Deformation Behavior of Laser Welds in High Temperature Oxidation Resistant Fe-Cr-Al Alloys for Fuel Cladding Applications

    SciTech Connect (OSTI)

    Field, Kevin G; Gussev, Maxim N; Yamamoto, Yukinori; Snead, Lance Lewis

    2014-11-01

    Ferritic-structured Fe-Cr-Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability of three model alloys in a range of Fe-(13-17.5)Cr-(3-4.4)Al in weight percent with a minor addition of yttrium using laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds has been carried out to determine the performance of welds as a function of alloy composition. Laser welding resulted in a defect free weld devoid of cracking or inclusions for all alloys studied. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. No significant correlation was found between the deformation behavior/mechanical performance of welds and the level of Cr or Al in the alloy ranges studied.

  17. Nuclear Fuels & Materials Spotlight Volume 4

    SciTech Connect (OSTI)

    I. J. van Rooyen,; T. M. Lillo; Y. Q. WU; P.A. Demkowicz; L. Scott; D.M. Scates; E. L. Reber; J. H. Jackson; J. A. Smith; D.L. Cottle; B.H. Rabin; M.R. Tonks; S.B. Biner; Y. Zhang; R.L. Williamson; S.R. Novascone; B.W. Spencer; J.D. Hales; D.R. Gaston; C.J. Permann; D. Anders; S.L. Hayes; P.C. Millett; D. Andersson; C. Stanek; R. Ali; S.L. Garrett; J.E. Daw; J.L. Rempe; J. Palmer; B. Tittmann; B. Reinhardt; G. Kohse; P. Ramuhali; H.T. Chien; T. Unruh; B.M. Chase; D.W. Nigg; G. Imel; J. T. Harris

    2014-04-01

    As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

  18. Operational Challenges of Extended Dry Storage of Spent Nuclear Fuel - 12550

    SciTech Connect (OSTI)

    Nichol, M. [Nuclear Energy Institute, Washington DC (United States)

    2012-07-01

    As a result of the termination of the Yucca Mountain used fuel repository program and a continuing climate of uncertainty in the national policy for nuclear fuel disposition, the likelihood has increased that extended storage, defined as more than 60 years, and subsequent transportation of used nuclear fuel after periods of extended storage may become necessary. Whether at the nation's 104 nuclear energy facilities, or at one or more consolidated interim storage facilities, the operational challenges of extended storage and transportation will depend upon the future US policy for Used Fuel Management and the future Regulatory Framework for EST, both of which should be developed with consideration of their operational impacts. Risk insights into the regulatory framework may conclude that dry storage and transportation operations should focus primarily on ensuring canister integrity. Assurance of cladding integrity may not be beneficial from an overall risk perspective. If assurance of canister integrity becomes more important, then mitigation techniques for potential canister degradation mechanisms will be the primary source of operational focus. If cladding integrity remains as an important focus, then operational challenges to assure it would require much more effort. Fundamental shifts in the approach to design a repository and optimize the back-end of the fuel cycle will need to occur in order to address the realities of the changes that have taken place over the last 30 years. Direct disposal of existing dual purpose storage and transportation casks will be essential to optimizing the back end of the fuel cycle. The federal used fuel management should focus on siting and designing a repository that meets this objective along with the development of CIS, and possibly recycling. An integrated approach to developing US policy and the regulatory framework must consider the potential operational challenges that they would create. Therefore, it should be integral to these efforts to redefine retrievability to apply to the dual purpose cask, and not to apply to individual assemblies. (authors)

  19. fuel cells | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    fuel cells | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets Newsletters Press Releases Photo Gallery Jobs Apply for Our Jobs Our Jobs Working at NNSA Blog

  20. Dry Transfer Systems for Used Nuclear Fuel

    SciTech Connect (OSTI)

    Brett W. Carlsen; Michaele BradyRaap

    2012-05-01

    The potential need for a dry transfer system (DTS) to enable retrieval of used nuclear fuel (UNF) for inspection or repackaging will increase as the duration and quantity of fuel in dry storage increases. This report explores the uses for a DTS, identifies associated general functional requirements, and reviews existing and proposed systems that currently perform dry fuel transfers. The focus of this paper is on the need for a DTS to enable transfer of bare fuel assemblies. Dry transfer systems for UNF canisters are currently available and in use for transferring loaded canisters between the drying station and storage and transportation casks.

  1. Nuclear fuel particles and method of making nuclear fuel compacts therefrom

    DOE Patents [OSTI]

    DeVelasco, Rubin I.; Adams, Charles C.

    1991-01-01

    Methods for making nuclear fuel compacts exhibiting low heavy metal contamination and fewer defective coatings following compact fabrication from a mixture of hardenable binder, such as petroleum pitch, and nuclear fuel particles having multiple layer fission-product-retentive coatings, with the dense outermost layer of the fission-product-retentive coating being surrounded by a protective overcoating, e.g., pyrocarbon having a density between about 1 and 1.3 g/cm.sup.3. Such particles can be pre-compacted in molds under relatively high pressures and then combined with a fluid binder which is ultimately carbonized to produce carbonaceous nuclear fuel compacts having relatively high fuel loadings.

  2. Waste Stream Analyses for Nuclear Fuel Cycles

    SciTech Connect (OSTI)

    N. R. Soelberg

    2010-08-01

    A high-level study was performed in Fiscal Year 2009 for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) Advanced Fuel Cycle Initiative (AFCI) to provide information for a range of nuclear fuel cycle options (Wigeland 2009). At that time, some fuel cycle options could not be adequately evaluated since they were not well defined and lacked sufficient information. As a result, five families of these fuel cycle options are being studied during Fiscal Year 2010 by the Systems Analysis Campaign for the DOE NE Fuel Cycle Research and Development (FCRD) program. The quality and completeness of data available to date for the fuel cycle options is insufficient to perform quantitative radioactive waste analyses using recommended metrics. This study has been limited thus far to qualitative analyses of waste streams from the candidate fuel cycle options, because quantitative data for wastes from the front end, fuel fabrication, reactor core structure, and used fuel for these options is generally not yet available.

  3. Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

    DOE Patents [OSTI]

    Bradley, John G.

    1982-01-01

    A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception.

  4. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    DOE Patents [OSTI]

    Mariani, Robert Dominick

    2014-09-09

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  5. Method of monitoring stored nuclear fuel elements

    SciTech Connect (OSTI)

    Borloo, E.; Buergers, W.; Crutzen, S.; Vinche, C.

    1983-05-24

    To monitor a nuclear fuel element or fuel elements located in a store, e.g. a pond in a swimming pool reactor, the store is illuminated ultrasonically using one or more transducers transmitting ultrasonic signals in one or more predetermined directions to obtain an output which, because it depends on the number and relative location of the fuel elements in the store, and the structure of the store itself is distinctive to the fuel elements or elements stored therein. From this distinctive output is derived an identity unique to the stored fuel element or elements and a reference signal indicative of the whole structure when intact, the reference signal and identity being recorded. Subsequent ultrasonic testing of the store and its contents under identical operating conditions produces a signal which is compared to the recorded reference signal and if different therefrom reveals the occurrence of tampering with the store and/or the fuel element or elements.

  6. Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

    SciTech Connect (OSTI)

    Carmack, W.J.; Husser, D.L.; Mohr, T.C.; Richardson, W.C.

    2004-02-04

    New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developed to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.

  7. International nuclear fuel cycle fact book

    SciTech Connect (OSTI)

    Leigh, I.W.

    1988-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source or information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

  8. International Nuclear Fuel Cycle Fact Book

    SciTech Connect (OSTI)

    Leigh, I.W.

    1992-05-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need exists costs for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book has been compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NMEA activities reports; and proceedings of conferences and workshops. The data listed typically do not reflect any single source but frequently represent a consolidation/combination of information.

  9. International Nuclear Fuel Cycle Fact Book

    SciTech Connect (OSTI)

    Leigh, I W; Mitchell, S J

    1990-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops, etc. The data listed do not reflect any one single source but frequently represent a consolidation/combination of information.

  10. Nuclear fuel elements made from nanophase materials

    DOE Patents [OSTI]

    Heubeck, N.B.

    1998-09-08

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

  11. Nuclear fuel elements made from nanophase materials

    DOE Patents [OSTI]

    Heubeck, Norman B.

    1998-01-01

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

  12. Computational Design of Advanced Nuclear Fuels

    SciTech Connect (OSTI)

    Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

    2014-06-03

    The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

  13. Advanced LWR Nuclear Fuel Development

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

    LWRS Overview and Select Research Highlights J. Busby and K. Leonard Oak Ridge National Laboratory with a host of contributors Overview Presentation for the Nuclear Energy Enabling Technologies September 15, 2015 Video Conference FY16 - Consolidated Innovative Nuclear Research * Light water reactor sustainability (LWRS) is one of the four technical areas in the Reactor Concepts Research, Development and Demonstration Program sections in the FOA. * Activities in LWRS also overlap with NEET

  14. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  15. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States (CIS)). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  16. Pilot Application to Nuclear Fuel Cycle Options | Department of Energy

    Energy Savers [EERE]

    Pilot Application to Nuclear Fuel Cycle Options Pilot Application to Nuclear Fuel Cycle Options A Screening Method for Guiding R&D Decisions: Pilot Application to Screen Nuclear Fuel Cycle Options The Department of Energy's Office of Nuclear Energy (DOE-NE) invests in research and development (R&D) to ensure that the United States will maintain its domestic nuclear energy capability and scientific and technical leadership in the international community of nuclear power nations in the

  17. Nuclear Fuel Cycle Option Catalog SAND2015-2174 W

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

    of Energy which provided information about the potential benefits and challenges of nuclear fuel cycle options (i.e., the complete nuclear energy system from mining to...

  18. Activities Related to Storage of Spent Nuclear Fuel | Department...

    Office of Environmental Management (EM)

    Related to Storage of Spent Nuclear Fuel More Documents & Publications Nuclear Regulatory Commission Fifth National Report for the Joint Convention on the Safety of Spent...

  19. Future nuclear fuel cycles: prospects and challenges

    SciTech Connect (OSTI)

    Boullis, Bernard

    2008-07-01

    Solvent extraction has played, from the early steps, a major role in the development of nuclear fuel cycle technologies, both in the front end and back end. Today's stakes in the field of energy enhance further than before the need for a sustainable management of nuclear materials. Recycling actinides appears as a main guideline, as much for saving resources as for minimizing the final waste impact, and many options can be considered. Strengthened by the important and outstanding performance of recent PUREX processing plants, solvent-extraction processes seem a privileged route to meet the new and challenging requirements of sustainable future nuclear systems. (author)

  20. Spent Nuclear Fuel Alternative Technology Decision Analysis

    SciTech Connect (OSTI)

    Shedrow, C.B.

    1999-11-29

    The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

  1. Strategy for the Management and Disposal of Used Nuclear Fuel...

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

    Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level...

  2. Summary of nuclear fuel reprocessing activities around the world

    SciTech Connect (OSTI)

    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.

  3. Advanced LWR Nuclear Fuel Development

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

    LOCALIZED DEFORMATION IN IRRADIATED AUSTENITIC STEELS ORNL: M.N. Gussev*, K.G. Field, J.T. Busby, K.J. Leonard, PNNL: T.S. Byun, University of Michigan: K.J. Stephenson, G.S. Was. 1 *Oak Ridge National Laboratory Fuel Cycle & Isotopes Division P.O. Box 2008, MS-6151 Oak Ridge, TN 37831, USA Tel. (865) 574-44-56 Fax. (865) 241-3650 gussevmn@ornl.gov Credits This presentation is based on several papers and conference talks presented/published recently. Some of these are listed below: SEM-EBSD

  4. Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites |

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

    Department of Energy Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites In January 2013, the Department of Energy issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of

  5. Supply Security in Future Nuclear Fuel Markets

    SciTech Connect (OSTI)

    Seward, Amy M.; Wood, Thomas W.; Gitau, Ernest T.; Ford, Benjamin E.

    2013-11-18

    Previous PNNL work has shown the existing nuclear fuel markets to provide a high degree of supply security, including the ability to respond to supply disruptions that occur for technical and non-technical reasons. It is in the context of new reactor designs that is, reactors likely to be licensed and market ready over the next several decades that fuel supply security is most relevant. Whereas the fuel design and fabrication technology for existing reactors are well known, the construction of a new set of reactors could stress the ability of the existing market to provide adequate supply redundancy. This study shows this is unlikely to occur for at least thirty years, as most reactors likely to be built in the next three decades will be evolutions of current designs, with similar fuel designs to existing reactors.

  6. Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

    SciTech Connect (OSTI)

    Xu Wu; Piyush Sabharwall; Jason Hales

    2014-07-01

    This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INLs fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO2/zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

  7. Nuclear Fuel Storage and Transportation Planning Project Overview |

    Energy Savers [EERE]

    Department of Energy Fuel Storage and Transportation Planning Project Overview Nuclear Fuel Storage and Transportation Planning Project Overview PDF icon Nuclear Fuel Storage and Transportation Planning Project Overview More Documents & Publications Section 180(c) Ad Hoc Working Group DOE Office of Nuclear Energy Transportation Plan Ad Hoc Working Group

  8. Alabama Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","5,043",15.6,"37,941",24.9 "Coal","11,441",35.3,"63,050",41.4 "Hydro and Pumped

  9. Arizona Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (nw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",3937,14.9,"31,200",27.9 "Coal","6,233",23.6,"43,644",39.1 "Hydro and Pumped Storage","2,937",11.1,"6,831",6.1

  10. Arkansas Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,835",11.5,"15,023",24.6 "Coal","4,535",28.4,"28,152",46.2 "Hydro and Pumped

  11. California Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,390",6.5,"32,201",15.8 "Coal",374,0.6,"2,100",1.0 "Hydro and Pumped Storage","13,954",20.7,"33,260",16.3

  12. Connecticut Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,103",25.4,"16,750",50.2 "Coal",564,6.8,"2,604",7.8 "Hydro and Pumped Storage",151,1.8,400,1.2 "Natural

  13. Florida Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (nw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,924",6.6,"23,936",10.4 "Coal","9,975",16.9,"59,897",26.1 "Hydro and Pumped Storage",55,0.1,177,0.1 "Natural

  14. Georgia Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,061",11.1,"33,512",24.4 "Coal","13,230",36.1,"73,298",53.3 "Hydro and Pumped

  15. Illinois Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","11,441",25.9,"96,190",47.8 "Coal","15,551",35.2,"93,611",46.5 "Hydro and Pumped Storage",34,0.1,119,0.1 "Natural

  16. Iowa Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",601,4.1,"4,451",7.7 "Coal","6,956",47.7,"41,283",71.8 "Hydro and Pumped Storage",144,1.0,948,1.6 "Natural

  17. Kansas Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,160",9.2,"9,556",19.9 "Coal","5,179",41.3,"32,505",67.8 "Hydro and Pumped Storage",3,"*",13,"*"

  18. Louisiana Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,142",8.0,"18,639",18.1 "Coal","3,417",12.8,"23,924",23.3 "Hydro and Pumped Storage",192,0.7,"1,109",1.1

  19. Maryland Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,705",13.6,"13,994",32.1 "Coal","4,886",39.0,"23,668",54.3 "Hydro and Pumped Storage",590,4.7,"1,667",3.8

  20. Massachusetts Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, smmer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",685,5.0,"5,918",13.8 "Coal","1,669",12.2,"8,306",19.4 "Hydro and Pumped Storage","1,942",14.2,659,1.5 "Natural

  1. Michigan Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy Source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,947",13.2,"29,625",26.6 "Coal","11,531",38.7,"65,604",58.8 "Hydro and Pumped Storage","2,109",7.1,228,0.2

  2. Minnesota Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,594",10.8,"13,478",25.1 "Coal","4,789",32.5,"28,083",52.3 "Hydro and Pumped Storage",193,1.3,840,1.6 "Natural

  3. Mississippi Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,251",8.0,"9,643",17.7 "Coal","2,526",16.1,"13,629",25.0 "Natural Gas","11,640",74.2,"29,619",54.4

  4. Nebraska Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,245",15.8,"11,054",30.2 "Coal","3,932",50.0,"23,363",63.8 "Hydro and Pumped Storage",278,3.5,"1,314",3.6

  5. Tennessee Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,401",15.9,"27,739",33.7 "Coal","8,805",41.1,"43,670",53.0 "Hydro and Pumped

  6. Texas Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,966",4.6,"41,335",10.0 "Coal","22,335",20.6,"150,173",36.5 "Hydro and Pumped Storage",689,0.6,"1,262",0.3

  7. Ohio Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,134",6.5,"15,805",11.0 "Coal","21,360",64.6,"117,828",82.1 "Hydro and Pumped Storage",101,0.3,429,0.3 "Natural

  8. Pennsylvania Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","9,540",20.9,"77,828",33.9 "Coal","18,481",40.6,"110,369",48.0 "Hydro and Pumped

  9. Vermont Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",620,55.0,"4,782",72.2 "Hydro and Pumped Storage",324,28.7,"1,347",20.3 "Natural Gas","-","-",4,0.1 "Other

  10. Virginia Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,501",14.5,"26,572",36.4 "Coal","5,868",24.3,"25,459",34.9 "Hydro and Pumped Storage","4,107",17.0,10,"*"

  11. Wisconsin Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (nw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,584",8.9,"13,281",20.7 "Coal","8,063",45.2,"40,169",62.5 "Hydro and Pumped Storage",492,2.8,"2,112",3.3

  12. Nuclear power generation and fuel cycle report 1996

    SciTech Connect (OSTI)

    1996-10-01

    This report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the worldwide nuclear fuel market. Long term projections of U.S. nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed. A discussion on decommissioning of nuclear power plants is included.

  13. LWRS Fuels Pathway: Engineering Design and Fuels Pathway Initial Testing of the Hot Water Corrosion System

    SciTech Connect (OSTI)

    Dr. John Garnier; Dr. Kevin McHugh

    2012-09-01

    The Advanced LWR Nuclear Fuel Development R&D pathway performs strategic research focused on cladding designs leading to improved reactor core economics and safety margins. The research performed is to demonstrate the nuclear fuel technology advancements while satisfying safety and regulatory limits. These goals are met through rigorous testing and analysis. The nuclear fuel technology developed will assist in moving existing nuclear fuel technology to an improved level that would not be practical by industry acting independently. Strategic mission goals are to improve the scientific knowledge basis for understanding and predicting fundamental nuclear fuel and cladding performance in nuclear power plants, and to apply this information in the development of high-performance, high burn-up fuels. These will result in improved safety, cladding, integrity, and nuclear fuel cycle economics. To achieve these goals various methods for non-irradiated characterization testing of advanced cladding systems are needed. One such new test system is the Hot Water Corrosion System (HWCS) designed to develop new data for cladding performance assessment and material behavior under simulated off-normal reactor conditions. The HWCS is capable of exposing prototype rodlets to heated, high velocity water at elevated pressure for long periods of time (days, weeks, months). Water chemistry (dissolved oxygen, conductivity and pH) is continuously monitored. In addition, internal rodlet heaters inserted into cladding tubes are used to evaluate repeated thermal stressing and heat transfer characteristics of the prototype rodlets. In summary, the HWCS provides rapid ex-reactor evaluation of cladding designs in normal (flowing hot water) and off-normal (induced cladding stress), enabling engineering and manufacturing improvements to cladding designs before initiation of the more expensive and time consuming in-reactor irradiation testing.

  14. Missouri Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,190",5.5,"8,996",9.7 "Coal","12,070",55.5,"75,047",81.3 "Hydro and Pumped Storage","1,221",5.6,"2,427",2.6

  15. Washington Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary Energy Source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,097",3.6,"9,241",8.9 "Coal","1,340",4.4,"8,527",8.2 "Hydro and Pumped Storage","21,495",70.5,"68,342",66.0

  16. Nuclear reactor fuel rod attachment system

    DOE Patents [OSTI]

    Not Available

    1980-09-17

    A reusable system is described for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member. The locking cap has two opposing fingers shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed.

  17. A MULTIDIMENSIONAL AND MULTIPHYSICS APPROACH TO NUCLEAR FUEL BEHAVIOR SIMULATION

    SciTech Connect (OSTI)

    R. L. Williamson; J. D. Hales; S. R. Novascone; M. R. Tonks; D. R. Gaston; C. J. Permann; D. Andrs; R. C. Martineau

    2012-04-01

    Important aspects of fuel rod behavior, for example pellet-clad mechanical interaction (PCMI), fuel fracture, oxide formation, non-axisymmetric cooling, and response to fuel manufacturing defects, are inherently multidimensional in addition to being complicated multiphysics problems. Many current modeling tools are strictly 2D axisymmetric or even 1.5D. This paper outlines the capabilities of a new fuel modeling tool able to analyze either 2D axisymmetric or fully 3D models. These capabilities include temperature-dependent thermal conductivity of fuel; swelling and densification; fuel creep; pellet fracture; fission gas release; cladding creep; irradiation growth; and gap mechanics (contact and gap heat transfer). The need for multiphysics, multidimensional modeling is then demonstrated through a discussion of results for a set of example problems. The first, a 10-pellet rodlet, demonstrates the viability of the solution method employed. This example highlights the effect of our smeared cracking model and also shows the multidimensional nature of discrete fuel pellet modeling. The second example relies on our the multidimensional, multiphysics approach to analyze a missing pellet surface problem. As a final example, we show a lower-length-scale simulation coupled to a continuum-scale simulation.

  18. Method and apparatus for close packing of nuclear fuel assemblies

    DOE Patents [OSTI]

    Newman, D.F.

    1993-03-30

    The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

  19. Method and apparatus for close packing of nuclear fuel assemblies

    DOE Patents [OSTI]

    Newman, Darrell F.

    1993-01-01

    The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

  20. Pyrochemical processing of DOE spent nuclear fuel

    SciTech Connect (OSTI)

    Laidler, J.J.

    1995-02-01

    A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or {open_quotes}pyroprocessing,{close_quotes} provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (>99.9%) separation of transuranics. The resultant waste forms from the pyroprocess, are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory.

  1. Ion irradiation testing of Improved Accident Tolerant Cladding Materials

    SciTech Connect (OSTI)

    Anderoglu, Osman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tesmer, Joseph R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maloy, Stuart A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-01-14

    This report summarizes the results of ion irradiations conducted on two FeCrAl alloys (named as ORNL A&B) for improving the accident tolerance of LWR nuclear fuel cladding. After irradiation with 1.5 MeV protons to ~0.5 to ~1 dpa and 300C nanoindentations were performed on the cross-sections along the ion range. An increase in hardness was observed in both alloys. Microstructural analysis shows radiation induced defects.

  2. Nuclear Fuels Storage and Transportation Planning Project (NFST) Program

    Office of Environmental Management (EM)

    Status | Department of Energy Fuels Storage and Transportation Planning Project (NFST) Program Status Nuclear Fuels Storage and Transportation Planning Project (NFST) Program Status Presentation made by Jeff Williams for the NTSF annual meeting held from May 14-16, 2013 in Buffalo, NY. PDF icon Nuclear Fuels Storage and Transportation Planning Project (NFST) Program Status More Documents & Publications Nuclear Fuel Storage and Transportation Planning Project Overview DOE Office of

  3. Report on interim storage of spent nuclear fuel

    SciTech Connect (OSTI)

    Not Available

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  4. Nuclear Fuels Storage & Transportation Planning Project | Department of

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

    Energy Nuclear Fuels Storage & Transportation Planning Project Nuclear Fuels Storage & Transportation Planning Project Independent Spent Fuel Storage Installation (ISFSI) at the shutdown Connecticut Yankee site. The ISFSI includes 40 multi-purpose canisters, within vertical concrete storage casks, containing 1019 used nuclear fuel assemblies [412.3 metric ton heavy metal (MTHM)] and 3 canisters of greater-than-class-C (GTCC) low-level radioactive waste. Photo courtesy of Connecticut

  5. Optimally moderated nuclear fission reactor and fuel source therefor

    DOE Patents [OSTI]

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  6. Instrumentation. Nondestructive Examination for Verification of Canister and Cladding Integrity - FY2013 Status Update

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Jones, Anthony M.; Pardini, Allan F.; Denslow, Kayte M.; Crawford, Susan L.; Larche, Michael R.

    2013-09-30

    This report documents FY13 efforts for two instrumentation subtasks under storage and transportation. These instrumentation tasks relate to developing effective nondestructive evaluation (NDE) methods and techniques to (1) verify the integrity of metal canisters for the storage of used nuclear fuel (UNF) and to (2) characterize hydrogen effects in UNF cladding to facilitate safe storage and retrieval.

  7. Dynamic Systems Analysis Report for Nuclear Fuel Recycle

    SciTech Connect (OSTI)

    Brent Dixon; Sonny Kim; David Shropshire; Steven Piet; Gretchen Matthern; Bill Halsey

    2008-12-01

    This report examines the time-dependent dynamics of transitioning from the current United States (U.S.) nuclear fuel cycle where used nuclear fuel is disposed in a repository to a closed fuel cycle where the used fuel is recycled and only fission products and waste are disposed. The report is intended to help inform policy developers, decision makers, and program managers of system-level options and constraints as they guide the formulation and implementation of advanced fuel cycle development and demonstration efforts and move toward deployment of nuclear fuel recycling infrastructure.

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

    SciTech Connect (OSTI)

    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 products, fission products, and actinides. Waste forms now used or studied for used LWR fuels can be used for some of these waste streams but some waste forms may need to be developed for unique waste streams.

  9. OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

    SciTech Connect (OSTI)

    Sindelar, R.; Vinson, D.; Iyer, N.; Fisher, D.

    2010-11-03

    Following discharge from research reactors, spent nuclear fuel may be stored 'wet' in water pools or basins, or it may be stored 'dry' in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel. For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

  10. High Burn-Up Spent Nuclear Fuel Vibration Integrity Study 15134

    SciTech Connect (OSTI)

    Wang, Jy-An John; Wang, Hong; Jiang, Hao; Bevard, Bruce Balkcom; Howard, Rob L; Scaglione, John M

    2015-01-01

    The Oak Ridge National Laboratory (ORNL) has developed the cyclic integrated reversible-bending fatigue tester (CIRFT) approach to successfully demonstrate the controllable fatigue fracture on high burnup (HBU) spent nuclear fuel (SNF) in a normal vibration mode. CIRFT enables examination of the underlying mechanisms of SNF system dynamic performance. Due to the inhomogeneous composite structure of the SNF system, the detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained from a CIRFT system measurement. Therefore, finite element analyses (FEAs) are used to translate the global moment-curvature measurement into local stress-strain profiles for further investigation. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance. Fuel structure contributes to SNF system stiffness. There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interactions. SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.

  11. Pyroprocess for processing spent nuclear fuel

    DOE Patents [OSTI]

    Miller, William E.; Tomczuk, Zygmunt

    2002-01-01

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

  12. Nuclear fuel rods along the sand dunes

    SciTech Connect (OSTI)

    Regan, M.B.

    1993-05-10

    Just north of the small town of Covert, Michigan, Consumers Power Co. officials and environmental activists are locked in a battle that marks a new phase in the nation's long-running struggle over nuclear power. The company's Palisades power plant reactor needs refueling. But the utility has no more room for the spent fuel rods it must place in its water-filled storage pool. So Consumers is taking advantage of a 1990 Nuclear Regulatory Commission rule that lets utilities store waste above ground without agency review. Palisades officials plan to transfer older radioactive fuel rods from its storage pool into concrete and steel silo-like casks on a site overlooking Lake Michigan. Over the next decade, nearly half of the nation's 109 operating nuclear plants will run out of space in water-filled storage pools and be forced to consider aboveground storage. The Palisades plant is causing a stir because it is the first to exploit the 1990 NRC rule, which doesn't require utilities to seek approval for waste-storage sites as long as the waste is stored in an approved container. Before 1990, five other utilities had received the agency O.K. for above-ground storage - but only after a lengthy and exhaustive analysis of each site.

  13. Proceedings of GLOBAL 2013: International Nuclear Fuel Cycle Conference - Nuclear Energy at a Crossroads

    SciTech Connect (OSTI)

    2013-07-01

    The Global conference is a forum for the discussion of the scientific, technical, social and regulatory aspects of the nuclear fuel cycle. Relevant topics include global utilization of nuclear energy, current fuel cycle technologies, advanced reactors, advanced fuel cycles, nuclear nonproliferation and public acceptance.

  14. Fuel handling system for a nuclear reactor

    DOE Patents [OSTI]

    Saiveau, James G.; Kann, William J.; Burelbach, James P.

    1986-01-01

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  15. Nuclear reactor fuel rod attachment system

    DOE Patents [OSTI]

    Christiansen, David W. (Kennewick, WA)

    1982-01-01

    A reusable system for removably attaching a nuclear reactor fuel rod (12) to a support member (14). A locking cap (22) is secured to the fuel rod (12) and a locking strip (24) is fastened to the support member (14). The locking cap (22) has two opposing fingers (24a and 24b) shaped to form a socket having a body portion (26). The locking strip has an extension (36) shaped to rigidly attach to the socket's body portion (26). The locking cap's fingers are resiliently deflectable. For attachment, the locking cap (22) is longitudinally pushed onto the locking strip (24) causing the extension (36) to temporarily deflect open the fingers (24a and 24b) to engage the socket's body portion (26). For removal, the process is reversed.

  16. Spent nuclear fuel project path forward preliminary safety evaluation

    SciTech Connect (OSTI)

    Brehm, J.R.; Crowe, R.D.; Siemer, J.M.; Wojdac, L.F.; Hosler, A.G.

    1995-03-01

    This preliminary safety evaluation (PSE) provides validation of the initial project design criteria for the Spent Nuclear Fuel Project (SNFP) Path Forward for removal of fuel from K Basins.

  17. Fuel assembly transfer basket for pool type nuclear reactor vessels

    DOE Patents [OSTI]

    Fanning, Alan W. (San Jose, CA); Ramsour, Nicholas L. (San Jose, CA)

    1991-01-01

    A fuel assembly transfer basket for a pool type, liquid metal cooled nuclear reactor having a side access loading and unloading port for receiving and relinquishing fuel assemblies during transfer.

  18. EA-1977: Acceptance and Disposition of Used Nuclear Fuel Containing...

    Energy Savers [EERE]

    Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany EA-1977: Acceptance and Disposition of Used Nuclear Fuel Containing U.S.-Origin...

  19. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    SciTech Connect (OSTI)

    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 then discussed with respect to what is known in the literature about their behavior in a reprocessing facility. The context for the evaluation in this document is a UO2-based fuel processed through an aqueous-based reprocessing system with a TBP-based solvent extraction chemistry. None of these elements form sufficiently volatile compounds in the context of the reprocessing facility to be of regulatory concern.

  20. DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER

    SciTech Connect (OSTI)

    F. Habashi

    1998-06-26

    The DOE Spent Nuclear Fuel Disposal Container (SNF DC) supports the confinement and isolation of waste within the Engineered Barrier System of the Mined Geologic Disposal System (MGDS). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the access mains, and emplaced in emplacement drifts. The DOE Spent Nuclear Fuel Disposal Container provides long term confinement of DOE SNF waste, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The DOE SNF Disposal Containers provide containment of waste for a designated period of time, and limit radionuclide release thereafter. The disposal containers maintain the waste in a designated configuration, withstand maximum handling and rockfall loads, limit the individual waste canister temperatures after emplacement. The disposal containers also limit the introduction of moderator into the disposal container during the criticality control period, resist corrosion in the expected repository environment, and provide complete or limited containment of waste in the event of an accident. Multiple disposal container designs may be needed to accommodate the expected range of DOE Spent Nuclear Fuel. The disposal container will include outer and inner barrier walls and outer and inner barrier lids. Exterior labels will identify the disposal container and contents. Differing metal barriers will support the design philosophy of defense in depth. The use of materials with different failure mechanisms prevents a single mode failure from breaching the waste package. The corrosion-resistant inner barrier and inner barrier lid will be constructed of a high-nickel alloy and the corrosion-allowance outer barrier and outer barrier lid will be made of carbon steel. The DOE Spent Nuclear Fuel Disposal Containers interface with the emplacement drift environment by transferring heat from the waste to the external environment and by protecting the DOE waste canisters and their contents from damage/degradation by the external environment. The disposal containers also interface with the SNF by limiting access of moderator and oxidizing agents to the waste. The disposal containers interface with the Ex-Container System's emplacement drift disposal container supports. The disposal containers interface with the Canister Transfer System, Waste Emplacement System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement and remediation of the disposal container.

  1. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    SciTech Connect (OSTI)

    Wang, Jy-An John; Jiang, Hao

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as used nuclear fuel [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 Nm was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 Nm. Failure was observed in nine of the tested rod specimens. The cycles-to-failure ranged from 1.22 105 to 4.70 106, and the amplitudes varied from 15.2 to 7.6 Nm. The measurements at the interrupts indicated a range of flexural rigidity from 30 to 50 Nm2. Online monitoring revealed that the flexural rigidity was slightly lower due to the higher loading, from 25 to 42 Nm2. Generally, no substantial change of rigidity was observed based on the online monitoring during the cyclic fatigue testing process. Overall, the decreasing trend of sample lifetime with increasing amplitude is well defined.

  2. International nuclear fuel cycle fact book. Revision 6

    SciTech Connect (OSTI)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.; Jeffs, A.G.

    1986-01-01

    The International Fuel Cycle Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2.

  3. EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing

    Energy Savers [EERE]

    U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany | Department of Energy 77: Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany SUMMARY This EA will evaluate the potential environmental impacts of a DOE proposal to accept spent nuclear fuel from the

  4. Used Nuclear Fuel Loading and Structural Performance Under Normal

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

    Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan | Department of Energy Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan Used nuclear fuel (UNF) must maintain its integrity during the storage

  5. EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina | Department

    Office of Environmental Management (EM)

    of Energy 79: Spent Nuclear Fuel Management, Aiken, South Carolina EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina SUMMARY The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. PUBLIC COMMENT

  6. Safer nuclear power

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

    Safer nuclear power 1663 Los Alamos science and technology magazine Latest Issue:October 2015 past issues All Issues » submit Safer nuclear power Experiments at Los Alamos reveal that alternative fuel rod cladding materials can make nuclear power plants dramatically less likely to suffer a Fukushima-type explosion in the event of a nuclear accident March 25, 2013 Safer nuclear power Nuclear generating station Los Alamos scientists, in collaboration with scientists from the Idaho and Oak Ridge

  7. FY-13 FCRD Milestone M3FT-13OR0202311 Weldability of ORNL Accident Tolerant Fuel Cladding Model Alloys For Thin Walled Tubes

    SciTech Connect (OSTI)

    Field, Kevin G [ORNL; Gussev, Maxim N [ORNL; Yamamoto, Yukinori [ORNL

    2013-07-01

    Ferritic FeCrAl-based alloys show increased oxidation resistance for accident tolerant applications as fuel cladding. This study focuses on investigating the weldability of three model FeCrAl alloys with varying alloy compositions using laser-welding techniques. A detailed study of the mechanical properties of bead-on-plate welds was used to determine the quality of welds as a function of alloy composition. Laser welding resulted in defect free welds devoid of cracking or inclusions. Initial results indicate a reduction in the yield strength of weldments compared to the base material due to distinct changes in the microstructure within the fusion zone. Although a loss of yield strength was observed, there was no significant difference in the magnitude of the tensile property changes with varying Cr or Al content. Also, there was no evidence of embrittlement; the material in the fusion zones demonstrated ductile behavior with high local ductility.

  8. Spent Nuclear Fuel project integrated safety management plan

    SciTech Connect (OSTI)

    Daschke, K.D.

    1996-09-17

    This document is being revised in its entirety and the document title is being revised to ``Spent Nuclear Fuel Project Integrated Safety Management Plan.

  9. Measurement of plutonium in spent nuclear fuel by self-induced x-ray fluorescence

    SciTech Connect (OSTI)

    Hoover, Andrew S; Rudy, Cliff R; Tobin, Steve J; Charlton, William S; Stafford, A; Strohmeyer, D; Saavadra, S

    2009-01-01

    Direct measurement of the plutonium content in spent nuclear fuel is a challenging problem in non-destructive assay. The very high gamma-ray flux from fission product isotopes overwhelms the weaker gamma-ray emissions from plutonium and uranium, making passive gamma-ray measurements impossible. However, the intense fission product radiation is effective at exciting plutonium and uranium atoms, resulting in subsequent fluorescence X-ray emission. K-shell X-rays in the 100 keV energy range can escape the fuel and cladding, providing a direct signal from uranium and plutonium that can be measured with a standard germanium detector. The measured plutonium to uranium elemental ratio can be used to compute the plutonium content of the fuel. The technique can potentially provide a passive, non-destructive assay tool for determining plutonium content in spent fuel. In this paper, we discuss recent non-destructive measurements of plutonium X-ray fluorescence (XRF) signatures from pressurized water reactor spent fuel rods. We also discuss how emerging new technologies, like very high energy resolution microcalorimeter detectors, might be applied to XRF measurements.

  10. A Multi-Stage Wear Model for Grid-to-Rod Fretting of Nuclear Fuel Rods

    SciTech Connect (OSTI)

    Blau, Peter Julian

    2014-01-01

    The wear of fuel rod cladding against the supporting structures in the cores of pressurized water nuclear reactors (PWRs) is an important and potentially costly tribological issue. Grid-to-rod fretting (GTRF), as it is known, involves not only time-varying contact conditions, but also elevated temperatures, flowing hot water, aqueous tribo-corrosion, and the embrittling effects of neutron fluences. The multi-stage, closed-form analytical model described in this paper relies on published out-of-reactor wear and corrosion data and a set of simplifying assumptions to portray the conversion of frictional work into wear depth. The cladding material of interest is a zirconium-based alloy called Zircaloy-4, and the grid support is made of a harder and more wear-resistant material. Focus is on the wear of the cladding. The model involves an incubation stage, a surface oxide wear stage, and a base alloy wear stage. The wear coefficient, which is a measure of the efficiency of conversion of frictional work into wear damage, can change to reflect the evolving metallurgical condition of the alloy. Wear coefficients for Zircaloy-4 and for a polyphase zirconia layer were back-calculated for a range of times required to wear to a critical depth. Inputs for the model, like the friction coefficient, are taken from the tribology literature in lieu of in-reactor tribological data. Concepts of classical fretting were used as a basis, but are modified to enable the model to accommodate the complexities of the PWR environment. Factors like grid spring relaxation, pre-oxidation of the cladding, multiple oxide phases, gap formation, impact, and hydrogen embrittlement are part of the problem definition but uncertainties in their relative roles limits the ability to validate the model. Sample calculations of wear depth versus time in the cladding illustrate how GTRF wear might occur in a discontinuous fashion during months-long reactor operating cycles. A means to account for grid/rod gaps and repetitive impact effects on GTRF wear is proposed

  11. EM Safely and Efficiently Manages Spent Nuclear Fuel | Department...

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

    (MVDS) is a dry storage system used at Idaho Engineering Laboratory and at Fort St. Vrain to store spent nuclear fuel. Each storage location holds one spent nuclear container. ...

  12. Method and means of packaging nuclear fuel rods for handling

    DOE Patents [OSTI]

    Adam, Milton F.

    1979-01-01

    Nuclear fuel rods, especially spent nuclear fuel rods that may show physical distortion, are encased within a metallic enclosing structure by forming a tube about the fuel rod. The tube has previously been rolled to form an overlapping tubular structure and then unrolled and coiled about an axis perpendicular to the tube. The fuel rod is inserted into the tube as the rolled tube is removed from a coiled strip and allowed to reassume its tubular shape about the fuel rod. Rollers support the coiled strip in an open position as the coiled strip is uncoiled and allowed to roll about the fuel rod.

  13. A Critical Step Toward Sustainable Nuclear Fuel Disposal | Department of

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

    Energy A Critical Step Toward Sustainable Nuclear Fuel Disposal A Critical Step Toward Sustainable Nuclear Fuel Disposal January 26, 2012 - 2:30pm Addthis Secretary Chu Secretary Chu Former Secretary of Energy The Blue Ribbon Commission on America's Nuclear Future was formed at the direction of the President to conduct a comprehensive review of polices for managing the back end of the nuclear fuel cycle. If we are going to ensure that the United States remains at the forefront of nuclear

  14. Used Fuel Disposition Used Nuclear Fuel Storage and Transportation

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

    ASI Review Meeting September 17, 2014 Used Fuel Disposition Today's Discussion n Our ... - Transportation - Analysis Used Fuel Disposition 3 Overall Objectives * Develop ...

  15. Categorization of Used Nuclear Fuel Inventory in Support of a Comprehensive

    Office of Environmental Management (EM)

    National Nuclear Fuel Cycle Strategy | Department of Energy Categorization of Used Nuclear Fuel Inventory in Support of a Comprehensive National Nuclear Fuel Cycle Strategy Categorization of Used Nuclear Fuel Inventory in Support of a Comprehensive National Nuclear Fuel Cycle Strategy The Office of Nuclear Energy has conducted a technical review and assessment of the total current inventory [~70,150 MTHM as of 2011] of domestic discharged used nuclear fuel (UNF) and estimated an amount to be

  16. Separator assembly for use in spent nuclear fuel shipping cask

    DOE Patents [OSTI]

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  17. EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering

    Office of Environmental Management (EM)

    Laboratory Environmental Restoration and Waste Management Programs | Department of Energy 3: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs SUMMARY This EIS considers programmatic (DOE-wide) alternative approaches to safely, efficiently, and responsibly manage existing and

  18. Technology Readiness Levels for Advanced Nuclear Fuels and Materials Development

    SciTech Connect (OSTI)

    Jon Carmack

    2014-01-01

    The Technology Readiness Level (TRL) process is used to quantitatively assess the maturity of a given technology. The TRL process has been developed and successfully used by the Department of Defense (DOD) for development and deployment of new technology and systems for defense applications. In addition, NASA has also successfully used the TRL process to develop and deploy new systems for space applications. Advanced nuclear fuels and materials development is a critical technology needed for closing the nuclear fuel cycle. Because the deployment of a new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the advanced fuel development program is very useful as a management and tracking tool. This report provides definition of the technology readiness level assessment process as defined for use in assessing nuclear fuel technology development for the Advanced Fuel Campaign (AFC).

  19. Characterization plan for Hanford spent nuclear fuel

    SciTech Connect (OSTI)

    Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

    1994-12-01

    Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF.

  20. Electric heater for nuclear fuel rod simulators

    DOE Patents [OSTI]

    McCulloch, Reginald W. (Knoxville, TN); Morgan, Jr., Chester S. (Oak Ridge, TN); Dial, Ralph E. (Concord, TN)

    1982-01-01

    The present invention is directed to an electric cartridge-type heater for use as a simulator for a nuclear fuel pin in reactor studies. The heater comprises an elongated cylindrical housing containing a longitudinally extending helically wound heating element with the heating element radially inwardly separated from the housing. Crushed cold-pressed preforms of boron nitride electrically insulate the heating element from the housing while providing good thermal conductivity. Crushed cold-pressed preforms of magnesia or a magnesia-15 percent boron nitride mixture are disposed in the cavity of the helical heating element. The coefficient of thermal expansion of the magnesia or the magnesia-boron nitride mixture is higher than that of the boron nitride disposed about the heating element for urging the boron nitride radially outwardly against the housing during elevated temperatures to assure adequate thermal contact between the housing and the boron nitride.

  1. Discrete Modeling of Early-Life Thermal Fracture in Ceramic Nuclear Fuel

    SciTech Connect (OSTI)

    Spencer, B. W.; Huang, H.; Dolbow, J. E.; Hales, J. D.

    2015-03-01

    Fracturing of ceramic fuel pellets heavily influences performance of light water reactor (LWR) fuel. Early in the life of fuel, starting with the initial power ramp, large thermal gradients cause high tensile hoop and axial stresses in the outer region of the fuel pellets, resulting in the formation of radial and axial cracks. Circumferential cracks form due to thermal gradients that occur when the power is ramped down. These thermal cracks cause the fuel to expand radially, closing the pellet/cladding gap and enhancing the thermal conductance across that gap, while decreasing the effective conductivity of the fuel in directions normal to the cracking. At lower length scales, formation of microcracks is an important contributor to the decrease in bulk thermal conductivity that occurs over the life of the fuel as the burnup increases. Because of the important effects that fracture has on fuel performance, a realistic, physically based fracture modeling capability is essential to predict fuel behavior in a wide variety of normal and abnormal conditions. Modeling fracture within the context of the finite element method, which is based on continuous interpolations of solution variables, has always been challenging because fracture is an inherently discontinuous phenomenon. Work is underway at Idaho National Laboratory to apply two modeling techniques model fracture as a discrete displacement discontinuity to nuclear fuel: The extended finite element method (XFEM), and discrete element method (DEM). XFEM is based on the standard finite element method, but with enhancements to represent discontinuous behavior. DEM represents a solid as a network of particles connected by bonds, which can arbitrarily fail if a fracture criterion is reached. This paper presents initial results applying the aforementioned techniques to model fuel fracturing. This work has initially focused on early life behavior of ceramic LWR fuel. A coupled thermal-mechanical XFEM method that includes discontinuities in both temperature and displacement fields at crack locations has been developed and is being applied to thermal fracture of LWR fuel. A DEM model of coupled heat conduction and solid mechanics has been developed and used to simulate random initiation and propagation of thermally driven cracks during initial power cycles. This DEM model predicts the formation of realistic radial cracking patterns during power rise and circumferential cracks as power is ramped down. These initial results are very encouraging, and these techniques are expected to provide improved understanding of fuel behavior in a wide variety of conditions.

  2. Apparatus for leaching core material from clad nuclear fuel pin segments

    DOE Patents [OSTI]

    Yarbro, Orlan O. (Knoxville, TN)

    1980-01-01

    This invention relates to improved apparatus for counter-currently contacting liquids and solids to dissolve, or leach, a selected component of the solids while minimizing back-mixing of the liquid phase. The apparatus includes an elongated drum which is rotatable about its longitudinal axis in either direction and is partitioned radially into a solids-inlet/liquid-outlet compartment at one end, a solids-outlet/liquid-inlet compartment at its other end, and leaching compartments therebetween. The drum is designed to operate with its acid-inlet end elevated and with the longitudinal axis of the drum at an angle in the range of from about 3.degree. to 14.degree. to the horizontal. Each leaching compartment contains a chute assembly for advancing solids into the next compartment in the direction of solids flow when the drum is rotated in a selected direction. The chute assembly includes a solids-transfer baffle and a chute in the form of a slotted, skewed, conical frustum portion. When the drum is rotated in the direction opposite to that effecting solids transfer, the solids-transfer baffles continually separate and re-mix the solids and liquids in their respective compartments. The partitions defining the leaching compartments are formed with corresponding outer, annular, imperforate regions, each region extending inwardly from the partition rim to an annular array of perforations concentric with the rim. In each leaching compartment, the spacing between the rim and the perforations determines the depth of liquid at the liquid-outlet end of the compartment. The liquid input to the drum assembly flows continuously through the compartments, preventing back-mixing due to density differences, whereas backflow due to waves generated by the solids-transfer baffles is virtually eliminated because of the tilted orientation of the drum assembly.

  3. Impact of Nuclear Energy Futures on Advanced Fuel Cycle Options

    SciTech Connect (OSTI)

    Dixon, B.W.; Piet, S.J.

    2004-10-03

    The Nuclear Waste Policy Act requires the Secretary of Energy to inform Congress before 2010 on the need for a second geologic repository for spent nuclear fuel. By that time, the spent fuel discharged from current commercial reactors will exceed the statutory limit of the first repository. There are several approaches to eliminate the need for another repository in this century. This paper presents a high-level analysis of these spent fuel management options in the context of a full range of possible nuclear energy futures. The analysis indicates the best option to implement varies depending on the nuclear energy future selected.

  4. Spent nuclear fuel discharges from U.S. reactors 1994

    SciTech Connect (OSTI)

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  5. Benefits and concerns of a closed nuclear fuel cycle

    SciTech Connect (OSTI)

    Widder, Sarah H.

    2010-11-17

    Nuclear power can play an important role in our energy future, contributing to increasing electricity demand while at the same time decreasing carbon dioxide emissions. However, the nuclear fuel cycle in the United States today is unsustainable. As stated in the 1982 Nuclear Waste Policy Act, the U.S. Department of Energy is responsible for disposing of spent nuclear fuel generated by commercial nuclear power plants operating in a once-through fuel cycle in the deep geologic repository located at Yucca Mountain. However, unyielding political opposition to the site has hindered the commissioning process to the extant that the current administration has recently declared the unsuitability of the Yucca Mountain site. In light of this the DOE is exploring other options, including closing the fuel cycle through recycling and reprocessing of spent nuclear fuel. The possibility of closing the fuel cycle is receiving special attention because of its ability to minimize the final high level waste (HLW) package as well as recover additional energy value from the original fuel. The technology is, however, still very controversial because of the increased cost and proliferation risk it can present. To lend perspective on the closed fuel cycle alternative, this presents the arguments for and against closing the fuel cycle with respect to sustainability, proliferation risk, commercial viability, waste management, and energy security.

  6. Nuclear fuel cycle facility accident analysis handbook

    SciTech Connect (OSTI)

    1998-03-01

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

  7. Evaluation of prompt release of fission gas from a breached cladding. [PWR; BWR

    SciTech Connect (OSTI)

    El-Genk, M.S.; Kumar, R.M.

    1981-01-01

    It is a concern in the current safety analysis of nuclear reactors to understand the different release mechanisms of fission products to accurately determine the radiological source term for a wide variety of accidents. The mechanism which is least understood and which produces an uncertainty in determining the radiological source term during a reactor accident is the early release of fission gas present in the fuel-cladding gap through a cladding breach. In a loss-of-coolant type accident the fuel rods would be surrounded mainly by steam, therefore, the release of the gap gas can simply be treated as a discharge problem through an orifice. However, during reactor normal operation or in those accidents where the failed fuel rods are surrounded by liquid coolant, the release process of the gap gas would be strongly influenced by the coolant conditions (pressure, temperature and flow rate). The purpose of this work is to describe analytically the prompt escape of volatiles and gaseous fission products, present in the fuel-cladding gap through a cladding breach, where the fuel rod is surrounded by liquid coolant.

  8. Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

    SciTech Connect (OSTI)

    Katoh, Yutai; Terrani, Kurt A.

    2015-08-01

    Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

  9. Fuel rod retention device for a nuclear reactor

    DOE Patents [OSTI]

    Hylton, Charles L. (Madison Heights, VA)

    1984-01-01

    A device is described for supporting a nuclear fuel rod in a fuel rod assembly which allows the rod to be removed without disturbing other rods in the assembly. A fuel rod cap connects the rod to a bolt which is supported in the assembly end fitting by means of a locking assembly. The device is designed so that the bolt is held securely during normal reactor operation yet may be easily disengaged and the fuel rod removed when desired.

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

    National Nuclear Security Administration (NNSA)

    Administration (MOX) Fuel Fabrication Facility | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets Newsletters Press Releases Photo Gallery Jobs Apply for

  11. Mixed Oxide Fuel Fabrication Facility | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration Fuel Fabrication Facility | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets Newsletters Press Releases Photo Gallery Jobs Apply for Our

  12. Methods for making a porous nuclear fuel element

    DOE Patents [OSTI]

    Youchison, Dennis L; Williams, Brian E; Benander, Robert E

    2014-12-30

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  13. Spent nuclear fuel discharges from US reactors 1993

    SciTech Connect (OSTI)

    Not Available

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  14. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    SciTech Connect (OSTI)

    Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.

    2013-09-30

    This report fulfills the M2 milestone M2FT-13PN0912022, “Stranded Sites De-Inventorying Report.” In January 2013, the U.S. Department of Energy (DOE) issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013). Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of the Blue Ribbon Commission on America’s Nuclear Future, which identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses (BRC 2012). Shutdown sites are defined as those commercial nuclear power reactor sites where the nuclear power reactors have been shut down and the site has been decommissioned or is undergoing decommissioning. In this report, a preliminary evaluation of removing used nuclear fuel from 12 shutdown sites was conducted. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. These sites have no other operating nuclear power reactors at their sites and have also notified the U.S. Nuclear Regulatory Commission that their reactors have permanently ceased power operations and that nuclear fuel has been permanently removed from their reactor vessels. Shutdown reactors at sites having other operating reactors are not included in this evaluation.

  15. Thermoacoustic sensor for nuclear fuel temperaturemonitoring and heat transfer enhancement

    SciTech Connect (OSTI)

    James A. Smith; Dale K. Kotter; Randall A. Alli; Steven L. Garrett

    2013-05-01

    A new acoustical sensing system for the nuclear power industry has been developed at The Pennsylvania State University in collaboration with Idaho National Laboratories. This sensor uses the high temperatures of nuclear fuel to convert a nuclear fuel rod into a standing-wave thermoacoustic engine. When a standing wave is generated, the sound wave within the fuel rod will be propagated, by acoustic radiation, through the cooling fluid within the reactor or spent fuel pool and can be monitored a remote location external to the reactor. The frequency of the sound can be correlated to an effective temperature of either the fuel or the surrounding coolant. We will present results for a thermoacoustic resonator built into a Nitonic-60 (stainless steel) fuel rod that requires only one passive component and no heat exchangers.

  16. What to Expect When Readying to Move Spent Nuclear Fuel from...

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

    What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power Plants What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power ...

  17. AN ANALYTICAL FRAMEWORK FOR ASSESSING RELIABLE NUCLEAR FUEL SERVICE APPROACHES: ECONOMIC AND NON-PROLIFERATION MERITS OF NUCLEAR FUEL LEASING

    SciTech Connect (OSTI)

    Kreyling, Sean J.; Brothers, Alan J.; Short, Steven M.; Phillips, Jon R.; Weimar, Mark R.

    2010-08-11

    The goal of international nuclear policy since the dawn of nuclear power has been the peaceful expansion of nuclear energy while controlling the spread of enrichment and reprocessing technology. Numerous initiatives undertaken in the intervening decades to develop international agreements on providing nuclear fuel supply assurances, or reliable nuclear fuel services (RNFS) attempted to control the spread of sensitive nuclear materials and technology. In order to inform the international debate and the development of government policy, PNNL has been developing an analytical framework to holistically evaluate the economics and non-proliferation merits of alternative approaches to managing the nuclear fuel cycle (i.e., cradle-to-grave). This paper provides an overview of the analytical framework and discusses preliminary results of an economic assessment of one RNFS approach: full-service nuclear fuel leasing. The specific focus of this paper is the metrics under development to systematically evaluate the non-proliferation merits of fuel-cycle management alternatives. Also discussed is the utility of an integrated assessment of the economics and non-proliferation merits of nuclear fuel leasing.

  18. Pyroprocessing of fast flux test facility nuclear fuel

    SciTech Connect (OSTI)

    Westphal, B.R.; Wurth, L.A.; Fredrickson, G.L.; Galbreth, G.G.; Vaden, D.; Elliott, M.D.; Price, J.C.; Honeyfield, E.M.; Patterson, M.N. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID, 83415 (United States)

    2013-07-01

    Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primary fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electro-refined uranium products exceeded 99%. (authors)

  19. Pyroprocessing of Fast Flux Test Facility Nuclear Fuel

    SciTech Connect (OSTI)

    B.R. Westphal; G.L. Fredrickson; G.G. Galbreth; D. Vaden; M.D. Elliott; J.C. Price; E.M. Honeyfield; M.N. Patterson; L. A. Wurth

    2013-10-01

    Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primary fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electrorefined uranium products exceeded 99%.

  20. Method of increasing the deterrent to proliferation of nuclear fuels

    DOE Patents [OSTI]

    Rampolla, Donald S.

    1982-01-01

    A process of recycling protactinium-231 to enhance the utilization of radioactively hot uranium-232 in nuclear fuel for the purpose of making both fresh and spent fuel more resistant to proliferation. The uranium-232 may be obtained by the irradiation of protactinium-231 which is normally found in the spent fuel rods of a thorium base nuclear reactor. The production of protactinium-231 and uranium-232 would be made possible by the use of the thorium uranium-233 fuel cycle in power reactors.

  1. Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    B.R. Westphal; D. Vaden; S.X. Li; G.L. Fredrickson; R.D. Mariani

    2009-09-01

    During the pyroprocessing of spent nuclear fuel by electrochemical techniques, fission products are separated as the fuel is oxidized at the anode and refined uranium is deposited at the cathode. Those fission products that are oxidized into the molten salt electrolyte are considered active metals while those that do not react are considered noble metals. The primary noble metals encountered during pyroprocessing are molybdenum, zirconium, ruthenium, rhodium, palladium, and technetium. Pyroprocessing of spent fuel to date has involved two distinctly different electrorefiner designs, in particular the anode to cathode configuration. For one electrorefiner, the anode and cathode collector are horizontally displaced such that uranium is transported across the electrolyte medium. As expected, the noble metal removal from the uranium during refining is very high, typically in excess of 99%. For the other electrorefiner, the anode and cathode collector are vertically collocated to maximize uranium throughput. This arrangement results in significantly less noble metals removal from the uranium during refining, typically no better than 20%. In addition to electrorefiner design, operating parameters can also influence the retention of noble metals, albeit at the cost of uranium recovery. Experiments performed to date have shown that as much as 100% of the noble metals can be retained by the cladding hulls while affecting the uranium recovery by only 6%. However, it is likely that commercial pyroprocessing of spent fuel will require the uranium recovery to be much closer to 100%. The above mentioned design and operational issues will likely be driven by the effects of noble metal contamination on fuel fabrication and performance. These effects will be presented in terms of thermal properties (expansion, conductivity, and fusion) and radioactivity considerations. Ultimately, the incorporation of minor amounts of noble metals from pyroprocessing into fast reactor metallic fuel will be shown to be of no consequence to reactor performance.

  2. A Comparison Study of Various Nuclear Fuel Cycle Alternatives

    SciTech Connect (OSTI)

    Kwon, Eun-ha; Ko, Won-il

    2007-07-01

    As a nation develops its nuclear strategies, it must consider various aspects of nuclear energy such as sustainability, environmental-friendliness, proliferation-resistance, economics, technologies, and so on. Like all the policy decision, however, a nuclear fuel cycle option can not be superior in all aspects; the nation must identify its top priority and accordingly evaluate all the possible nuclear fuel cycle options. For such a purpose, this paper takes four different fuel cycle options that are likely adopted by the Korean government, considering the current status of nuclear power generation and the 3. Comprehensive Nuclear Energy Promotion Plan (CNEPP) - Once-through Cycle, DUPIC Recycle, Thermal Recycle and GEN-IV Recycle. The paper then evaluates each option in terms of resource utilization and waste generation. The analysis shows that the GEN-IV Recycle appears to be most competitive from these aspects. (authors)

  3. Fabrication of high exposure nuclear fuel pellets

    DOE Patents [OSTI]

    Frederickson, James R.

    1987-01-01

    A method is disclosed for making a fuel pellet for a nuclear reactor. A mixture is prepared of PuO.sub.2 and UO.sub.2 powders, where the mixture contains at least about 30% PuO.sub.2, and where at least about 12% of the Pu is the Pu.sup.240 isotope. To this mixture is added about 0.3 to about 5% of a binder having a melting point of at least about 250.degree. F. The mixture is pressed to form a slug and the slug is granulated. Up to about 4.7% of a lubricant having a melting point of at least about 330.degree. F. is added to the granulated slug. Both the binder and the lubricant are selected from a group consisting of polyvinyl carboxylate, polyvinyl alcohol, naturally occurring high molecular weight cellulosic polymers, chemically modified high molecular weight cellulosic polymers, and mixtures thereof. The mixture is pressed to form a pellet and the pellet is sintered.

  4. 105 K East and 105 K West fuel transfer bay crane use strategy for spent nuclear fuel path-forward

    SciTech Connect (OSTI)

    Ard, K.E.

    1996-04-02

    The purpose of this document is to outline the K Basins 30 ton crane qualification strategy for use in the Spent Nuclear Fuel Project fuel relocation campaign.

  5. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Guenther, R.J.; Johnson, A.B. Jr.; Lund, A.L.; Gilbert, E.R.

    1996-07-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  6. Energy Return on Investment from Recycling Nuclear Fuel

    SciTech Connect (OSTI)

    2011-08-17

    This report presents an evaluation of the Energy Return on Investment (EROI) from recycling an initial batch of 800 t/y of used nuclear fuel (UNF) through a Recycle Center under a number of different fuel cycle scenarios. The study assumed that apart from the original 800 t of UNF only depleted uranium was available as a feed. Therefore for each subsequent scenario only fuel that was derived from the previous fuel cycle scenario was considered. The scenarios represent a good cross section of the options available and the results contained in this paper and associated appendices will allow for other fuel cycle options to be considered.

  7. Neutron radiography of irradiated nuclear fuel at Idaho National Laboratory

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

    Craft, Aaron E.; Wachs, Daniel M.; Okuniewski, Maria A.; Chichester, David L.; Williams, Walter J.; Papaioannou, Glen C.; Smolinski, Andrew T.

    2015-09-10

    Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Idaho National Laboratory (INL) has multiple nuclear fuels research and development programs that routinely evaluate irradiated fuels using neutron radiography. The Neutron Radiography reactor (NRAD) sits beneath a shielded hot cell facility where neutron radiography and other evaluation techniques are performed on these highly radioactive objects. The NRAD currently uses the foil-film transfer technique for imaging fuel that is time consuming but provides high spatial resolution. This study describes the NRAD and hot cell facilities,more » the current neutron radiography capabilities available at INL, planned upgrades to the neutron imaging systems, and new facilities being brought online at INL related to neutron imaging.« less

  8. A review of nuclear fuel cycle options for developing nations

    SciTech Connect (OSTI)

    Harrison, R.K.; Scopatz, A.M.; Ernesti, M.

    2007-07-01

    A study of several nuclear reactor and fuel cycle options for developing nations was performed. All reactor choices were considered under a GNEP framework. Two advanced alternative reactor types, a nuclear battery-type reactor and a fuel reprocessing fast reactor were examined and compared with a conventional Generation III+ LWR reactor. The burn of nuclear fuel was simulated using ORIGEN 2.2 for each reactor type and the resulting information was used to compare the options in terms of waste produced, waste quality and repository impact. The ORIGEN data was also used to evaluate the economics of the fuel cycles using unit costs, discount rates and present value functions with the material balances. The comparison of the fuel cycles and reactors developed in this work provides a basis for the evaluation of subsidy programs and cost-benefit comparisons for various reactor parameters such as repository impact and proliferation risk versus economic considerations. (authors)

  9. Neutron radiography of irradiated nuclear fuel at Idaho National Laboratory

    SciTech Connect (OSTI)

    Craft, Aaron E.; Wachs, Daniel M.; Okuniewski, Maria A.; Chichester, David L.; Williams, Walter J.; Papaioannou, Glen C.; Smolinski, Andrew T.

    2015-09-10

    Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Idaho National Laboratory (INL) has multiple nuclear fuels research and development programs that routinely evaluate irradiated fuels using neutron radiography. The Neutron Radiography reactor (NRAD) sits beneath a shielded hot cell facility where neutron radiography and other evaluation techniques are performed on these highly radioactive objects. The NRAD currently uses the foil-film transfer technique for imaging fuel that is time consuming but provides high spatial resolution. This study describes the NRAD and hot cell facilities, the current neutron radiography capabilities available at INL, planned upgrades to the neutron imaging systems, and new facilities being brought online at INL related to neutron imaging.

  10. EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing...

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

    Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched ...

  11. Discovery sheds light on nuclear reactor fuel behavior during...

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

    Discovery sheds light on nuclear reactor fuel behavior during a severe event By Angela Hardin * November 20, 2014 Tweet EmailPrint A new discovery about the atomic structure of...

  12. Means for supporting fuel elements in a nuclear reactor

    DOE Patents [OSTI]

    Andrews, Harry N.; Keller, Herbert W.

    1980-01-01

    A grid structure for a nuclear reactor fuel assembly comprising a plurality of connecting members forming at least one longitudinally extending opening peripheral and inner fuel element openings through each of which openings at least one nuclear fuel element extends, said connecting members forming wall means surrounding said each peripheral and inner fuel element opening, a pair of rigid projections longitudinally spaced from one another extending from a portion of said wall means into said each peripheral and inner opening for rigidly engaging said each fuel element, respectively, yet permit individual longitudinal slippage thereof, and resilient means formed integrally on and from said wall means and positioned in said each peripheral and inner opening in opposed relationship with said projections and located to engage said fuel element to bias the latter into engagement with said rigid projections, respectively

  13. Coolant mass flow equalizer for nuclear fuel

    DOE Patents [OSTI]

    Betten, Paul R. (Windsor, CT)

    1978-01-01

    The coolant mass flow distribution in a liquid metal cooled reactor is enhanced by restricting flow in sub-channels defined in part by the peripheral fuel elements of a fuel assembly. This flow restriction, which results in more coolant flow in interior sub-channels, is achieved through the use of a corrugated liner positioned between the bundle of fuel elements and the inner wall of the fuel assembly coolant duct. The corrugated liner is expandable to accommodate irradiation induced growth of fuel assembly components.

  14. Advanced Fuels Campaign FY 2014 Accomplishments Report

    SciTech Connect (OSTI)

    Lori Braase; W. Edgar May

    2014-10-01

    The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance.

  15. Metallic Fast Reactor Fuel Fabrication for Global Nuclear Energy Partnership

    SciTech Connect (OSTI)

    Douglas E. Burkes; Randall S. Fielding; Douglas L. Porter

    2009-07-01

    Fast reactors are once again being considered for nuclear power generation, in addition to transmutation of long-lived fission products resident in spent nuclear fuels. This re-consideration follows with intense developmental programs for both fuel and reactor design. One of the two leading candidates for next generation fast reactor fuel is metal alloys, resulting primarily from the successes achieved in the 1960s to early 1990s with both the experimental breeding reactor-II and the fast flux test facility. The goal of the current program is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional, fast-spectrum nuclear fuel while destroying recycled actinides, thereby closing the nuclear fuel cycle. In order to meet this goal, the program must develop efficient and safe fuel fabrication processes designed for remote operation. This paper provides an overview of advanced casting processes investigated in the past, and the development of a gaseous diffusion calculation that demonstrates how straightforward process parameter modification can mitigate the loss of volatile minor actinides in the metal alloy melt.

  16. Advanced nuclear fuel cycles - Main challenges and strategic choices

    SciTech Connect (OSTI)

    Le Biez, V.; Machiels, A.; Sowder, A.

    2013-07-01

    A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness.

  17. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    SciTech Connect (OSTI)

    2000-10-12

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in the emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Multiple boiling water reactor (BWR) and pressurized water reactor (PWR) disposal container designs are needed to accommodate the expected range of spent fuel assemblies and provide long-term confinement of the commercial SNF. The disposal container will include outer and inner cylinder walls, outer cylinder lids (two on the top, one on the bottom), inner cylinder lids (one on the top, one on the bottom), and an internal metallic basket structure. Exterior labels will provide a means by which to identify the disposal container and its contents. The two metal cylinders, in combination with the cladding, Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lid will be made of high-nickel alloy. The basket will assist criticality control, provide structural support, and improve heat transfer. The Uncanistered SNF Disposal Container System interfaces with the emplacement drift environment and internal waste by transferring heat from the SNF to the external environment and by protecting the SFN assemblies and their contents from damage/degradation by the external environment. The system also interfaces with the SFN by limiting access of moderator and oxidizing agents of the SFN. The waste package interfaces with the Emplacement Drift System's emplacement drift pallets upon which the wasted packages are placed. The disposal container interfaces with the Assembly Transfer System, Waste Emplacement/Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement and retrieval of the disposal container/waste package.

  18. Mox fuel arrangement for nuclear core

    DOE Patents [OSTI]

    Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

    2001-05-15

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

  19. MOX fuel arrangement for nuclear core

    DOE Patents [OSTI]

    Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

    2001-07-17

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

  20. MOX fuel arrangement for nuclear core

    DOE Patents [OSTI]

    Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

    1998-01-01

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

  1. MOX fuel arrangement for nuclear core

    DOE Patents [OSTI]

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

    1998-10-13

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

  2. Technical bases for interim storage of spent nuclear fuel

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.

    1981-06-01

    The experience base for water storage of spent nuclear fuel has evolved since 1943. The technology base includes licensing documentation, standards, technology studies, pool operator experience, and documentation from public hearings. That base reflects a technology which is largely successful and mundane. It projects probable satisfactory water storage of spent water reactor fuel for several decades. Interim dry storage of spent water reactor fuel is not yet licensed in the US, but a data base and documentation have developed. There do not appear to be technological barriers to interim dry storage, based on demonstrations with irradiated fuel. Water storage will continue to be a part of spent fuel management at reactors. Whether dry storage becomes a prominent interim fuel management option depends on licensing and economic considerations. National policies will strongly influence how long the spent fuel remains in interim storage and what its final disposition will be.

  3. International nuclear fuel cycle fact book. Revision 4

    SciTech Connect (OSTI)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

    1984-03-01

    This Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids - international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

  4. International Nuclear Fuel Cycle Fact Book. Revision 5

    SciTech Connect (OSTI)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.; Jeffs, A.G.

    1985-01-01

    This Fact Book has been compiled in an effort to provide: (1) an overview of worldwide nuclear power and fuel cycle programs; and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

  5. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    SciTech Connect (OSTI)

    Clayton, J C

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated.

  6. Spent nuclear fuel discharges from US reactors 1992

    SciTech Connect (OSTI)

    Not Available

    1994-05-05

    This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

  7. Ukraine Fuel Removal: Fact Sheet | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Ukraine Fuel Removal: Fact Sheet | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets Newsletters Press Releases Photo Gallery Jobs Apply for Our Jobs Our Jobs

  8. The Prospective Role of JAEA Nuclear Fuel Cycle Engineering Laboratories

    SciTech Connect (OSTI)

    Ojima, Hisao; Dojiri, Shigeru; Tanaka, Kazuhiko; Takeda, Seiichiro; Nomura, Shigeo

    2007-07-01

    JAEA Nuclear Fuel Cycle Engineering Laboratories was established in 2005 to take over the activities of the JNC Tokai Works. Many kinds of development activities have been carried out since 1959. Among these, the results on the centrifuge for U enrichment, LWR spent fuel reprocessing and MOX fuel fabrication have already provided the foundation of the fuel cycle industry in Japan. R and D on the treatment and disposal of high-level waste and FBR fuel reprocessing has also been carried out. Through such activities, radioactive material release to the environment has been appropriately controlled and all nuclear materials have been placed under IAEA safeguards. The Laboratories has sufficient experience and ability to establish the next generation closed cycle and strives to become a world-class Center Of Excellence (COE). (authors)

  9. Damaged Spent Nuclear Fuel at U.S. DOE Facilities Experience and Lessons Learned

    SciTech Connect (OSTI)

    Brett W. Carlsen; Eric Woolstenhulme; Roger McCormack

    2005-11-01

    From a handling perspective, any spent nuclear fuel (SNF) that has lost its original technical and functional design capabilities with regard to handling and confinement can be considered as damaged. Some SNF was damaged as a result of experimental activities and destructive examinations; incidents during packaging, handling, and transportation; or degradation that has occurred during storage. Some SNF was mechanically destroyed to protect proprietary SNF designs. Examples of damage to the SNF include failed cladding, failed fuel meat, sectioned test specimens, partially reprocessed SNFs, over-heated elements, dismantled assemblies, and assemblies with lifting fixtures removed. In spite of the challenges involved with handling and storage of damaged SNF, the SNF has been safely handled and stored for many years at DOE storage facilities. This report summarizes a variety of challenges encountered at DOE facilities during interim storage and handling operations along with strategies and solutions that are planned or were implemented to ameliorate those challenges. A discussion of proposed paths forward for moving damaged and nondamaged SNF from interim storage to final disposition in the geologic repository is also presented.

  10. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, Charles W.

    1998-01-01

    A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

  11. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, C.W.

    1998-11-03

    A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

  12. Impact of Nuclear Energy Futures on Advanced Fuel Cycle Options

    SciTech Connect (OSTI)

    Brent W. Dixon; Steven J. Piet

    2004-10-01

    The Nuclear Waste Policy Act requires the Secretary of Energy to inform Congress before 2010 on the need for a second geologic repository for spent nuclear fuel. By that time, the spent fuel discharged from current commercial reactors will exceed the statutory limit of the first repository (63,000 MTiHM commercial, 7,000 MT non-commercial). There are several approaches to eliminate the need for another repository in this century. This paper presents a high-level analysis of these spent fuel management options in the context of a full range of possible nuclear energy futures. The analysis indicates the best option to implement varies depending on the nuclear energy future selected. The first step in understanding the need for different spent fuel management approaches is to understand the size of potential spent fuel inventories. A full range of potential futures for domestic commercial nuclear energy is considered. These energy futures are as follows: 1. Existing License Completion - Based on existing spent fuel inventories plus extrapolation of future plant-by-plant discharges until the end of each operating license, including known license extensions. 2. Extended License Completion - Based on existing spent fuel inventories plus a plant-by-plant extrapolation of future discharges assuming on all operating plants having one 20-year extension. 3. Continuing Level Energy Generation - Based on extension of the current ~100 GWe installed commercial base and average spent fuel discharge of 2100 MT/yr through the year 2100. 4. Continuing Market Share Generation Based on a 1.8% compounded growth of the electricity market through the year 2100, matched by growing nuclear capacity and associated spent fuel discharge. 5. Growing Market Share Generation - Extension of current nuclear capacity and associated spent fuel discharge through 2100 with 3.2% growth representing 1.5% market growth (all energy, not just electricity) and 1.7% share growth. Share growth results in tripling market share by 2100 from the current 8.4% to 25%, equivalent to continuing the average market growth of last 50 years for an additional 100 years. Five primary spent fuel management strategies are assessed against each of the energy futures to determine the number of geological repositories needed and how the first repository would be used. The geological repository site at Yucca Mountain, Nevada, has the physical potential to accommodate all the spent fuel that will be generated by the current fleet of domestic commercial nuclear reactors, even with license extensions. If new nuclear plants are built in the future as replacements or additions, the United States will need to adopt spent fuel treatment to extend the life of the repository. Should a significant number of new nuclear plants be built, advanced fuel recycling will be needed to fully manage the spent fuel within a single repository. The analysis also considers the timeframe for most efficient implementation of new spent fuel management strategies. The mix of unprocessed spent fuel and processed high level waste in Yucca Mountain varies with each future and strategy. Either recycling must start before there is too much unprocessed waste emplaced or unprocessed waste will have to be retrieved later with corresponding costs. For each case, the latest date to implement reprocessing without subsequent retrieval is determined.

  13. Fuel Cycle Technologies Near Term Planning for Storage and Transportation of Used Nuclear Fuel

    Office of Environmental Management (EM)

    Nuclear Fuels Storage and Transportation Planning Project (NFST) Program Status Jeff Williams Project Director National Transportation Stakeholders Forum Buffalo, New York May 2013 2  "With the appropriate authorizations from Congress, the Administration currently plans to implement a program over the next 10 years that:  Sites, designs and licenses, constructs and begins operations of a pilot interim storage facility by 2021 with an initial focus on accepting used nuclear fuel from

  14. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  15. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  16. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2010-02-23

    Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

  17. Measures of the environmental footprint of the front end of the nuclear fuel cycle

    SciTech Connect (OSTI)

    E. Schneider; B. Carlsen; E. Tavrides; C. van der Hoeven; U. Phathanapirom

    2013-11-01

    Previous estimates of environmental impacts associated with the front end of the nuclear fuel cycle (FEFC) have focused primarily on energy consumption and CO2 emissions. Results have varied widely. This work builds upon reports from operating facilities and other primary data sources to build a database of front end environmental impacts. This work also addresses land transformation and water withdrawals associated with the processes of the FEFC. These processes include uranium extraction, conversion, enrichment, fuel fabrication, depleted uranium disposition, and transportation. To allow summing the impacts across processes, all impacts were normalized per tonne of natural uranium mined as well as per MWh(e) of electricity produced, a more conventional unit for measuring environmental impacts that facilitates comparison with other studies. This conversion was based on mass balances and process efficiencies associated with the current once-through LWR fuel cycle. Total energy input is calculated at 8.7 x 10- 3 GJ(e)/MWh(e) of electricity and 5.9 x 10- 3 GJ(t)/MWh(e) of thermal energy. It is dominated by the energy required for uranium extraction, conversion to fluoride compound for subsequent enrichment, and enrichment. An estimate of the carbon footprint is made from the direct energy consumption at 1.7 kg CO2/MWh(e). Water use is likewise dominated by requirements of uranium extraction, totaling 154 L/MWh(e). Land use is calculated at 8 x 10- 3 m2/MWh(e), over 90% of which is due to uranium extraction. Quantified impacts are limited to those resulting from activities performed within the FEFC process facilities (i.e. within the plant gates). Energy embodied in material inputs such as process chemicals and fuel cladding is identified but not explicitly quantified in this study. Inclusion of indirect energy associated with embodied energy as well as construction and decommissioning of facilities could increase the FEFC energy intensity estimate by a factor of up to 2.

  18. Fuel cycle cost uncertainty from nuclear fuel cycle comparison

    SciTech Connect (OSTI)

    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.

  19. Simulated dry storage test of a spent PWR nuclear fuel assembly in air

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.; Gilbert, E.R.; Oden, D.R.; Stidham, D.L.; Garnier, J.E.; Weeks, D.L.; Dobbins, J.C.

    1985-02-01

    The purpose of the dry storage test was to investigate the behavior of Zircaloy-clad spent fuel in air between 200 and 275/sup 0/C. Atmospheric air was used for the cover gas because of the interest in establishing regimes where air inleakage into an initially inert system would not cause potential fuel degradation. Samples of the cover gas atmosphere were extracted monthly to determine fission gas concentrations as a function of time. The oxygen concentration was monitored to detect oxygen depletion, which would signal oxidation of the fuel. The gas analyses indicated very low but detectable levels of /sup 85/Kr during the first month of the test. A large increase (five orders of magnitude) in /sup 85/Kr and the appearance of helium in the cover gas indicated that a fuel rod had breached during the second month of the test. Stress rupture calculations showed that the stresses and temperatures were too low to expect breaches to form in defect-free cladding. It is theorized that the breach occurred in a fuel rod weakened by an existing cladding or end cap defect. Calculations based on the rate of /sup 85/Kr release suggest that the diameter of the initial breach was about 25 microns. A post-test fuel examination will be performed to locate and investigate the cause of the cladding breach and to determine if detectable fuel degradation progressed after the breach occurred. The post-test evaluation will define the consequences of a fuel rod breach occurring in an air cover gas at 270/sup 0/C, followed by subsequent exposure to air at a prototypic descending temperature.

  20. Gap Analysis to Support Extended Storage of Used Nuclear Fuel | Department

    Office of Environmental Management (EM)

    of Energy Gap Analysis to Support Extended Storage of Used Nuclear Fuel Gap Analysis to Support Extended Storage of Used Nuclear Fuel The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives

  1. Temperature measuring analysis of the nuclear reactor fuel assembly

    SciTech Connect (OSTI)

    Urban, F. E-mail: zdenko.zavodny@stuba.sk; Ku?k, L. E-mail: zdenko.zavodny@stuba.sk; Bereznai, J. E-mail: zdenko.zavodny@stuba.sk; Zvodn, Z. E-mail: zdenko.zavodny@stuba.sk; Mukt, P. E-mail: zdenko.zavodny@stuba.sk

    2014-08-06

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  2. International nuclear fuel cycle fact book. [Contains glossary

    SciTech Connect (OSTI)

    Leigh, I.W.; Lakey, L.T.; Schneider, K.J.; Silviera, D.J.

    1987-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is a consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

  3. International Nuclear Fuel Cycle Fact Book. Revision 12

    SciTech Connect (OSTI)

    Leigh, I.W.

    1992-05-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need exists costs for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book has been compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NMEA activities reports; and proceedings of conferences and workshops. The data listed typically do not reflect any single source but frequently represent a consolidation/combination of information.

  4. International nuclear fuel cycle fact book: Revision 9

    SciTech Connect (OSTI)

    Leigh, I.W.

    1989-01-01

    The International Nuclear Fuel Cycle Fact Book has been compiled in an effort to provide current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. The Fact Book contains: national summaries in which a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; and international agencies in which a section for each of the international agencies which has significant fuel cycle involvement, and a listing of nuclear societies. The national summaries, in addition to the data described above, feature a small map for each country as well as some general information. The latter is presented from the perspective of the Fact Book user in the United States.

  5. Stainless steel 304 cladding mechanical properties and limitations during

    Office of Scientific and Technical Information (OSTI)

    steady state operation of U-ZrH TRIGA type fuel. (Conference) | SciTech Connect Stainless steel 304 cladding mechanical properties and limitations during steady state operation of U-ZrH TRIGA type fuel. Citation Details In-Document Search Title: Stainless steel 304 cladding mechanical properties and limitations during steady state operation of U-ZrH TRIGA type fuel. No abstract prepared. Authors: Coats, Richard Lee ; Dohner, Jeffrey Lynn ; Dahl, James J. ; Walker, Sharon Ann ; Greutman,

  6. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOE Patents [OSTI]

    Flint, O.

    1961-01-10

    Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

  7. Apparatus for injection casting metallic nuclear energy fuel rods

    DOE Patents [OSTI]

    Seidel, Bobby R. (Idaho Falls, ID); Tracy, Donald B. (Firth, ID); Griffiths, Vernon (Butte, MT)

    1991-01-01

    Molds for making metallic nuclear fuel rods are provided which present reduced risks to the environment by reducing radioactive waste. In one embodiment, the mold is consumable with the fuel rod, and in another embodiment, part of the mold can be re-used. Several molds can be arranged together in a cascaded manner, if desired, or several long cavities can be integrated in a monolithic multiple cavity re-usable mold.

  8. Technology Insights and Perspectives for Nuclear Fuel Cycle Concepts

    SciTech Connect (OSTI)

    S. Bays; S. Piet; N. Soelberg; M. Lineberry; B. Dixon

    2010-09-01

    The following report provides a rich resource of information for exploring fuel cycle characteristics. The most noteworthy trends can be traced back to the utilization efficiency of natural uranium resources. By definition, complete uranium utilization occurs only when all of the natural uranium resource can be introduced into the nuclear reactor long enough for all of it to undergo fission. Achieving near complete uranium utilization requires technologies that can achieve full recycle or at least nearly full recycle of the initial natural uranium consumed from the Earth. Greater than 99% of all natural uranium is fertile, and thus is not conducive to fission. This fact requires the fuel cycle to convert large quantities of non-fissile material into fissile transuranics. Step increases in waste benefits are closely related to the step increase in uranium utilization going from non-breeding fuel cycles to breeding fuel cycles. The amount of mass requiring a disposal path is tightly coupled to the quantity of actinides in the waste stream. Complete uranium utilization by definition means that zero (practically, near zero) actinide mass is present in the waste stream. Therefore, fuel cycles with complete (uranium and transuranic) recycle discharge predominately fission products with some actinide process losses. Fuel cycles without complete recycle discharge a much more massive waste stream because only a fraction of the initial actinide mass is burned prior to disposal. In a nuclear growth scenario, the relevant acceptable frequency for core damage events in nuclear reactors is inversely proportional to the number of reactors deployed in a fuel cycle. For ten times the reactors in a fleet, it should be expected that the fleet-average core damage frequency be decreased by a factor of ten. The relevant proliferation resistance of a fuel cycle system is enhanced with: decreasing reliance on domestic fuel cycle services, decreasing adaptability for technology misuse, enablement of material accountability, and decreasing material attractiveness.

  9. Next-generation nuclear fuel withstands high-temperature accident

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

    conditions U.S. DEPARTMENT OF ENERGY IDAHO FALLS, IDAHO, 83403 For Immediate Release: Sept. 25, 2013 Media Contacts: Teri Ehresman, 208-526-7785 teri.ehresman@inl.gov Bill Cabage (ORNL), 865-574-4399, cabagewh@ornl.gov Next-generation nuclear fuel withstands high-temperature accident conditions IDAHO FALLS - A safer and more efficient nuclear fuel is on the horizon. A team of researchers at the U.S. Department of Energy's Idaho National Laboratory (INL) and Oak Ridge National Laboratory

  10. Preparation of nuclear fuel spheres by flotation-internal gelation

    DOE Patents [OSTI]

    Haas, Paul A.; Fowler, Victor L.; Lloyd, Milton H.

    1987-01-01

    A simplified internal gelation process for the preparation of gel spheres of nuclear fuels. The process utilizes perchloroethylene as a gelation medium. Gelation is accomplished by directing droplets of a nuclear fuel broth into a moving volume of hot perchloroethylene (about 85.degree. C.) in a trough. Gelation takes place as the droplets float on the surface of the perchloroethylene and the resultant gel spheres are carried directly into an ager column which is attached to the trough. The aged spheres are disengaged from the perchloroethylene on a moving screen and are deposited in an aqueous wash column.

  11. Preparation of nuclear fuel spheres by flotation-internal gelation

    DOE Patents [OSTI]

    Haas, P.A.; Fowler, V.L.; Lloyd, M.H.

    1984-12-21

    A simplified internal gelation process is claimed for the preparation of gel spheres of nuclear fuels. The process utilizes perchloroethylene as a gelation medium. Gelation is accomplished by directing droplets of a nuclear fuel broth into a moving volume of hot perchloroethylene (about 85/sup 0/C) in a trough. Gelation takes place as the droplets float on the surface of the perchloroethylene and the resultant gel spheres are carried directly into an ager column which is attached to the trough. The aged spheres are disengaged from the perchloroethylene on a moving screen and are deposited in an aqueous wash column. 3 figs.

  12. SUPPLEMENT ANALYSIS PROPOSED SHIPMENT OF COMMERCIAL SPENT NUCLEAR FUEL

    Energy Savers [EERE]

    SUPPLEMENT ANALYSIS PROPOSED SHIPMENT OF COMMERCIAL SPENT NUCLEAR FUEL TO DOE NATIONAL LABORATORIES FOR RESEARCH AND DEVELOPMENT PURPOSES Office of Nuclear Energy U.S. DEPARTMENT OF ENERGY DECEMBER 2015 DOE/EIS-0203-SA-07 DOE/EIS-0250F-S-1-SA-02 Commercial Fuel Shipment SA DOE/EIS-0203-SA-07 December 2015 CONVERSION FACTORS Metric to English English to Metric Multiply by To get Multiply by To get Area Square kilometers 247.1 Acres Square kilometers 0.3861 Square miles Square meters 10.764 Square

  13. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    Energy Savers [EERE]

    Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites Prepared for U.S. Department of Energy Nuclear Fuels Storage and Transportation Planning Project Steven J. Maheras (PNNL) Ralph E. Best (PNNL) Steven B. Ross (PNNL) Kenneth A. Buxton (PNNL) Jeffery L. England (SRNL) Paul E. McConnell (SNL) Lawrence M. Massaro (FRA) Philip J. Jensen (PNNL) October 1, 2014 FCRD- NFST-2014-000091 Rev. 1 PNNL-22676 Rev. 4 DISCLAIMER This information was prepared as an account of work sponsored

  14. Railroad transportation of spent nuclear fuel

    SciTech Connect (OSTI)

    Wooden, D.G.

    1986-03-01

    This report documents a detailed analysis of rail operations that are important for assessing the risk of transporting high-level nuclear waste. The major emphasis of the discussion is towards ''general freight'' shipments of radioactive material. The purpose of this document is to provide a basis for selecting models and parameters that are appropriate for assessing the risk of rail transportation of nuclear waste.

  15. Quantitative assessment of proposals on assurance of nuclear fuel supply

    SciTech Connect (OSTI)

    Tanaka, T.; Kuno, Y.; Tanaka, S.

    2013-07-01

    The assurance of nuclear fuel supply has the potential to contribute to balancing peaceful use of nuclear power and nuclear nonproliferation. 5 proposals which provide the backup supply of the enrichment service in case of supply disruption, are investigated in this study. We investigated the 20 NPT countries which are non-nuclear-weapon states and possess operable commercial LWRs in October 2012 as potential participants for each proposal. As a result of literature researching, we have extracted factors that can be considered as important for a country to participate or not participate in the assurance of nuclear fuel supply. Then we have computed incentive and disincentive parameters for each country. The results show that the participation expectancy decreases in the order of IAEA Fuel Bank proposal, Russian LEU Reserve proposal, AFS proposal, WNA proposal and 6-Country proposal. The 'IAEA fuel bank proposal' would be triggered in case of the supply disruption which cannot be solved by the market mechanism and bilateral agreements.

  16. Nuclear waste actinides as fissile fuel in hybrid blankets

    SciTech Connect (OSTI)

    Sahin, S.; Al-Kusayer, T.A.

    1983-12-01

    The widespread use of the present LWRs produces substantial quantities of nuclear waste materials. Among those, actinide nuclear waste poses a serious problem of stockage because the associated half life times for actinides is measured in terms of geological time periods (several millions of years) so that no waste disposal guarantee over such time intervals can be given, except for space disposal. On the other hand, these nuclear waste actinides are very good fissionable materials for high energetic (D,T) fusion neutrons. It is therefore worthwhile to investigate their quality as potential nuclear fuel in hybrid blankets. The present study investigates the neutronic performance of hybrid blankets containing Np/sup 237/ and Cm/sup 244/ as fissile materials. The isotopic composition of Americium has been adjusted to the spent fuel isotope composition of a LWR. The geometrical design has been made, according to the AYMAN fussion-fission (hybrid) experimental facility, now in the very early phase of planning.

  17. Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

    SciTech Connect (OSTI)

    Muhammad Abir; Fahima Islam; Hyoung Koo Lee; Daniel Wachs

    2014-11-01

    The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the High Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.

  18. Nuclear fuel elements and method of making same

    DOE Patents [OSTI]

    Schweitzer, Donald G. (Bayport, NY)

    1992-01-01

    A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

  19. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    SciTech Connect (OSTI)

    Wang, Jy-An John; Wang, Hong; Jiang, Hao

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10? cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 Nm was obtained at maximum curvature 4.0 m?. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 Nm. Failure was observed in nine of the tested rod specimens. The cycles to failure were from 1.22 10? to 4.70 10?, when the amplitude varied from 15.2 to 7.6 Nm. The measurements at the interrupts indicated a range of flexural rigidity from 30 to 50 Nm. The on-line monitoring revealed that the flexural rigidity was a little lower due to the high level of loading, from 25 to 42 Nm. Generally, no substantial change of rigidity was observed based on on-line monitoring during the cyclic fatigue testing process. Overall, the decreasing trend of lifetime with the increasing amplitude is well defined.

  20. Nuclear Fuel Cycle Reasoner: PNNL FY13 Report

    SciTech Connect (OSTI)

    Hohimer, Ryan E.; Strasburg, Jana D.

    2013-09-30

    In Fiscal Year 2012 (FY12) PNNL implemented a formal reasoning framework and applied it to a specific challenge in nuclear nonproliferation. The Semantic Nonproliferation Analysis Platform (SNAP) was developed as a preliminary graphical user interface to demonstrate the potential power of the underlying semantic technologies to analyze and explore facts and relationships relating to the nuclear fuel cycle (NFC). In Fiscal Year 2013 (FY13) the SNAP demonstration was enhanced with respect to query and navigation usability issues.

  1. Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel Management

    Office of Environmental Management (EM)

    NRC's Integrated Strategy for NRC s Integrated Strategy for Spent Fuel Management Earl Easton 1 U.S. Nuclear Regulatory Commission May 25, 2010 Road to Yucca Mountain * 20+ years of preparation for the licensing i review * DOE application received in June 2008 and accepted for review in September 2008 * President Obama pursues alternatives to Yucca Mountain * DOE motion to withdraw in March 2010 2 * DOE motion to withdraw in March 2010 * Blue Ribbon Commission on America's Nuclear Future 2

  2. Air Shipment of Spent Nuclear Fuel from Romania to Russia

    SciTech Connect (OSTI)

    Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

    2010-10-01

    Romania successfully completed the worlds first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities for shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.

  3. Systems for the Intermodal Routing of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Peterson, Steven K; Liu, Cheng

    2015-01-01

    The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable system for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the selection of prioritization constraints and modifiers to determine route selection. The limitations of the current model and future directions for development are discussed, including the current state of information on possible intermodal transfer locations for spent fuel.

  4. Modeling Deep Burn TRISO Particle Nuclear Fuel

    SciTech Connect (OSTI)

    Besmann, Theodore M [ORNL; Stoller, Roger E [ORNL; Samolyuk, German D [ORNL; Schuck, Paul C [ORNL; Rudin, Sven [Los Alamos National Laboratory (LANL); Wills, John [Los Alamos National Laboratory (LANL); Wirth, Brian D. [University of California, Berkeley; Kim, Sungtae [University of Wisconsin, Madison; Morgan, Dane [University of Wisconsin, Madison; Szlufarska, Izabela [University of Wisconsin, Madison

    2012-01-01

    Under the DOE Deep Burn program TRISO fuel is being investigated as a fuel form for consuming plutonium and minor actinides, and for greater efficiency in uranium utilization. The result will thus be to drive TRISO particulate fuel to very high burn-ups. In the current effort the various phenomena in the TRISO particle are being modeled using a variety of techniques. The chemical behavior is being treated utilizing thermochemical analysis to identify phase formation/transformation and chemical activities in the particle, including kernel migration. First principles calculations are being used to investigate the critical issue of fission product palladium attack on the SiC coating layer. Density functional theory is being used to understand fission product diffusion within the plutonia oxide kernel. Kinetic Monte Carlo techniques are shedding light on transport of fission products, most notably silver, through the carbon and SiC coating layers. The diffusion of fission products through an alternative coating layer, ZrC, is being assessed via DFT methods. Finally, a multiscale approach is being used to understand thermal transport, including the effect of radiation damage induced defects, in a model SiC material.

  5. Interim report spent nuclear fuel retrieval system fuel handling development testing

    SciTech Connect (OSTI)

    Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

    1997-06-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

  6. U.S. Spent Nuclear Fuel Data as of December 31,2002 Table 3

    U.S. Energy Information Administration (EIA) Indexed Site

    equal sum of components because of independent rounding. Source: Energy Information Administration, Form RW-859, "Nuclear Fuel Data" (2002). Table 1 | Table 2 NuclearUranium Data...

  7. U.S. Spent Nuclear Fuel Data as of December 31, 2002

    U.S. Energy Information Administration (EIA) Indexed Site

    Latest spent nuclear data Release Date: October 1, 2004 Spent nuclear fuel data is collected by the Energy Information Administration (EIA) for the Office of Civilian Radioactive...

  8. Method of controlling crystallite size in nuclear-reactor fuels

    DOE Patents [OSTI]

    Lloyd, Milton H. (Oak Ridge, TN); Collins, Jack L. (Knoxville, TN); Shell, Sam E. (Oak Ridge, TN)

    1985-01-01

    Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules.

  9. Methods and apparatuses for the development of microstructured nuclear fuels

    DOE Patents [OSTI]

    Jarvinen, Gordon D.; Carroll, David W.; Devlin, David J.

    2009-04-21

    Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission products mixture.

  10. Nuclear Fuel Cycle Reasoner: PNNL FY12 Report

    SciTech Connect (OSTI)

    Hohimer, Ryan E.; Pomiak, Yekaterina G.; Neorr, Peter A.; Gastelum, Zoe N.; Strasburg, Jana D.

    2013-05-03

    Building on previous internal investments and leveraging ongoing advancements in semantic technologies, PNNL implemented a formal reasoning framework and applied it to a specific challenge in nuclear nonproliferation. The Semantic Nonproliferation Analysis Platform (SNAP) was developed as a preliminary graphical user interface to demonstrate the potential power of the underlying semantic technologies to analyze and explore facts and relationships relating to the nuclear fuel cycle (NFC). In developing this proof of concept prototype, the utility and relevancy of semantic technologies to the Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D) has been better understood.

  11. DEVELOPMENT OF METHODOLOGY AND FIELD DEPLOYABLE SAMPLING TOOLS FOR SPENT NUCLEAR FUEL INTERROGATION IN LIQUID STORAGE

    SciTech Connect (OSTI)

    Berry, T.; Milliken, C.; Martinez-Rodriguez, M.; Hathcock, D.; Heitkamp, M.

    2012-06-04

    This project developed methodology and field deployable tools (test kits) to analyze the chemical and microbiological condition of the fuel storage medium and determine the oxide thickness on the spent fuel basin materials. The overall objective of this project was to determine the amount of time fuel has spent in a storage basin to determine if the operation of the reactor and storage basin is consistent with safeguard declarations or expectations. This project developed and validated forensic tools that can be used to predict the age and condition of spent nuclear fuels stored in liquid basins based on key physical, chemical and microbiological basin characteristics. Key parameters were identified based on a literature review, the parameters were used to design test cells for corrosion analyses, tools were purchased to analyze the key parameters, and these were used to characterize an active spent fuel basin, the Savannah River Site (SRS) L-Area basin. The key parameters identified in the literature review included chloride concentration, conductivity, and total organic carbon level. Focus was also placed on aluminum based cladding because of their application to weapons production. The literature review was helpful in identifying important parameters, but relationships between these parameters and corrosion rates were not available. Bench scale test systems were designed, operated, harvested, and analyzed to determine corrosion relationships between water parameters and water conditions, chemistry and microbiological conditions. The data from the bench scale system indicated that corrosion rates were dependent on total organic carbon levels and chloride concentrations. The highest corrosion rates were observed in test cells amended with sediment, a large microbial inoculum and an organic carbon source. A complete characterization test kit was field tested to characterize the SRS L-Area spent fuel basin. The sampling kit consisted of a TOC analyzer, a YSI multiprobe, and a thickness probe. The tools were field tested to determine their ease of use, reliability, and determine the quality of data that each tool could provide. Characterization was done over a two day period in June 2011, and confirmed that the L Area basin is a well operated facility with low corrosion potential.

  12. Automated fuel pin loading system

    DOE Patents [OSTI]

    Christiansen, David W. (Kennewick, WA); Brown, William F. (West Richland, WA); Steffen, Jim M. (Richland, WA)

    1985-01-01

    An automated loading system for nuclear reactor fuel elements utilizes a gravity feed conveyor which permits individual fuel pins to roll along a constrained path perpendicular to their respective lengths. The individual lengths of fuel cladding are directed onto movable transports, where they are aligned coaxially with the axes of associated handling equipment at appropriate production stations. Each fuel pin can be reciprocated axially and/or rotated about its axis as required during handling steps. The fuel pins are inserted as a batch prior to welding of end caps by one of two disclosed welding systems.

  13. Strategy for the Management and Disposal of Used Nuclear Fuel and

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

    High-Level Radioactive Waste | Department of Energy Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste The Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste is a framework for moving toward a sustainable program to deploy an integrated system capable of transporting, storing, and disposing of used nuclear fuel

  14. Ukraine Loads U.S. Nuclear Fuel into Power Plant as Part of DOE-Ukraine

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

    Nuclear Fuel Qualification Program | Department of Energy Ukraine Loads U.S. Nuclear Fuel into Power Plant as Part of DOE-Ukraine Nuclear Fuel Qualification Program Ukraine Loads U.S. Nuclear Fuel into Power Plant as Part of DOE-Ukraine Nuclear Fuel Qualification Program April 9, 2010 - 12:11pm Addthis KYIV, UKRAINE - Officials from the U.S. Department of Energy's (DOE) Office of Nuclear Energy today (April 8, 2010) participated in a ceremony in Ukraine to mark the insertion of

  15. Retrievable fuel pin end member for a nuclear reactor

    DOE Patents [OSTI]

    Rosa, Jerry M. (Los Gatos, CA)

    1982-01-01

    A bottom end member (17b) on a retrievable fuel pin (13b) secures the pin (13b) within a nuclear reactor (12) by engaging on a transverse attachment rail (18) with a spring clip type of action. Removal and reinstallation if facilitated as only axial movement of the fuel pin (13b) is required for either operation. A pair of resilient axially extending blades (31) are spaced apart to define a slot (24) having a seat region (34) which receives the rail (18) and having a land region (37), closer to the tips (39) of the blades (31) which is normally of less width than the rail (18). Thus an axially directed force sufficient to wedge the resilient blades (31) apart is required to emplace or release the fuel pin (13b) such force being greater than the axial forces on the fuel pins (13b) which occur during operation of the reactor (12).

  16. Fuel Cycle Technologies Near Term Planning for Storage and Transportation of Used Nuclear Fuel

    Office of Environmental Management (EM)

    Implementation of Section 180(c) of the Nuclear Waste Policy Act, as amended National Transportation Stakeholder's Forum Buffalo, NY May 15, 2013 Section 180(c) Mandate "The Secretary shall provide technical assistance and funds to States for training for public safety officials of appropriate units of local government and Indian tribes through whose jurisdiction the Secretary plans to transport spent nuclear fuel or high-level radioactive waste [to an NWPA-authorized facility]. * The

  17. New Hampshire Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,247",29.8,"10,910",49.2 "Coal",546,13.1,"3,083",13.9 "Hydro and Pumped Storage",489,11.7,"1,478",6.7 "Natural

  18. New Jersey Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,108",22.3,"32,771",49.9 "Coal","2,036",11.1,"6,418",9.8 "Hydro and Pumped Storage",404,2.2,-176,-0.3 "Natural

  19. New York Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","5,271",13.4,"41,870",30.6 "Coal","2,781",7.1,"13,583",9.9 "Hydro and Pumped

  20. North Carolina Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,958",17.9,"40,740",31.7 "Coal","12,766",46.1,"71,951",55.9 "Hydro and Pumped

  1. South Carolina Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","6,486",27.0,"51,988",49.9 "Coal","7,230",30.1,"37,671",36.2 "Hydro and Pumped

  2. Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan

    SciTech Connect (OSTI)

    Adkins, Harold E.

    2013-04-01

    Under current U.S. Nuclear Regulatory Commission regulation, it is not sufficient for used nuclear fuel (UNF) to simply maintain its integrity during the storage period, it must maintain its integrity in such a way that it can withstand the physical forces of handling and transportation associated with restaging the fuel and moving it to treatment or recycling facilities, or a geologic repository. Hence it is necessary to understand the performance characteristics of aged UNF cladding and ancillary components under loadings stemming from transport initiatives. Researchers would like to demonstrate that enough information, including experimental support and modeling and simulation capabilities, exists to establish a preliminary determination of UNF structural performance under normal conditions of transport (NCT). This research, development and demonstration (RD&D) plan describes a methodology, including development and use of analytical models, to evaluate loading and associated mechanical responses of UNF rods and key structural components. This methodology will be used to provide a preliminary assessment of the performance characteristics of UNF cladding and ancillary components under rail-related NCT loading. The methodology couples modeling and simulation and experimental efforts currently under way within the Used Fuel Disposition Campaign (UFDC). The methodology will involve limited uncertainty quantification in the form of sensitivity evaluations focused around available fuel and ancillary fuel structure properties exclusively. The work includes collecting information via literature review, soliciting input/guidance from subject matter experts, performing computational analyses, planning experimental measurement and possible execution (depending on timing), and preparing a variety of supporting documents that will feed into and provide the basis for future initiatives. The methodology demonstration will focus on structural performance evaluation of Westinghouse WE 1717 pressurized water reactor fuel assemblies with a discharge burnup range of 30-58 GWd/MTU (assembly average), loaded in a representative high-capacity (?32 fuel rod assemblies) transportation package. Evaluations will be performed for representative normal conditions of rail transport involving a rail conveyance capable of meeting the Association of American Railroads (AAR) S-2043 specification. UNF modeling is anticipated to be defined to the pellet-cladding level and take in to account influences associated with spacer grids, intermediate fluid mixers, and control components. The influence of common degradation issues such as ductile-to-brittle-transition will also be accounted for. All model development and analysis will be performed with commercially available software packages exclusively. Inputs and analyses will be completely documented, all supporting information will be traceable, and bases will be defendable so as to be most useful to the U.S. Department of Energy community and mission. The expected completion date is the end of fiscal year (FY) 2013.

  3. Waste management plan for Hanford spent nuclear fuel characterization activities

    SciTech Connect (OSTI)

    Chastain, S.A. [Westinghouse Hanford Co., Richland, WA (United States); Spinks, R.L. [Pacific Northwest Lab., Richland, WA (United States)

    1994-10-17

    A joint project was initiated between Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratory (PNL) to address critical issues associated with the Spent Nuclear Fuel (SNF) stored at the Hanford Site. Recently, particular attention has been given to remediation of the SNF stored in the K Basins. A waste management plan (WMP) acceptable to both parties is required prior to the movement of selected material to the PNL facilities for examination. N Reactor and Single Pass Reactor (SPR) fuel has been stored for an extended period of time in the N Reactor, PUREX, K-East, and K-West Basins. Characterization plans call for transport of fuel material form the K Basins to the 327 Building Postirradiation Testing Laboratory (PTL) in the 300 Area for examination. However, PNL received a directive stating that no examination work will be started in PNL hot cell laboratories without an approved disposal route for all waste generated related to the activity. Thus, as part of the Characterization Program Management Plan for Hanford Spent Nuclear Fuel, a waste management plan which will ensure that wastes generated as a result of characterization activities conducted at PNL will be accepted by WHC for disposition is required. This document contains the details of the waste handling plan that utilizes, to the greatest extent possible, established waste handling and disposal practices at Hanford between PNL and WHC. Standard practices are sufficient to provides for disposal of most of the waste materials, however, special consideration must be given to the remnants of spent nuclear fuel elements following examination. Fuel element remnants will be repackaged in an acceptable container such as the single element canister and returned to the K Basins for storage.

  4. Nuclear reactor vessel fuel thermal insulating barrier

    DOE Patents [OSTI]

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  5. Portable instrument for inspecting irradiated nuclear fuel assemblies

    DOE Patents [OSTI]

    Nicholson, Nicholas (Los Alamos, NM); Dowdy, Edward J. (Los Alamos, NM); Holt, David M. (Los Alamos, NM); Stump, Jr., Charles J. (Santa Fe, NM)

    1985-01-01

    A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

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

    DOE Patents [OSTI]

    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.

  7. Method for reprocessing and separating spent nuclear fuels

    DOE Patents [OSTI]

    Krikorian, Oscar H. (Danville, CA); Grens, John Z. (Livermore, CA); Parrish, Sr., William H. (Walnut Creek, CA)

    1983-01-01

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

  8. HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER

    SciTech Connect (OSTI)

    BROWN,LC; BESENBRUCH,GE; LENTSCH,RD; SCHULTZ,KR; FUNK,JF; PICKARD,PS; MARSHALL,AC; SHOWALTER,SK

    2003-06-01

    OAK B202 HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER. Combustion of fossil fuels, used to power transportation, generate electricity, heat homes and fuel industry provides 86% of the world's energy. Drawbacks to fossil fuel utilization include limited supply, pollution, and carbon dioxide emissions. Carbon dioxide emissions, thought to be responsible for global warming, are now the subject of international treaties. Together, these drawbacks argue for the replacement of fossil fuels with a less-polluting potentially renewable primary energy such as nuclear energy. Conventional nuclear plants readily generate electric power but fossil fuels are firmly entrenched in the transportation sector. Hydrogen is an environmentally attractive transportation fuel that has the potential to displace fossil fuels. Hydrogen will be particularly advantageous when coupled with fuel cells. Fuel cells have higher efficiency than conventional battery/internal combustion engine combinations and do not produce nitrogen oxides during low-temperature operation. Contemporary hydrogen production is primarily based on fossil fuels and most specifically on natural gas. When hydrogen is produced using energy derived from fossil fuels, there is little or no environmental advantage. There is currently no large scale, cost-effective, environmentally attractive hydrogen production process available for commercialization, nor has such a process been identified. The objective of this work is to find an economically feasible process for the production of hydrogen, by nuclear means, using an advanced high-temperature nuclear reactor as the primary energy source. Hydrogen production by thermochemical water-splitting (Appendix A), a chemical process that accomplishes the decomposition of water into hydrogen and oxygen using only heat or, in the case of a hybrid thermochemical process, by a combination of heat and electrolysis, could meet these goals. Hydrogen produced from fossil fuels has trace contaminants (primarily carbon monoxide) that are detrimental to precious metal catalyzed fuel cells, as is now recognized by many of the world's largest automobile companies. Thermochemical hydrogen will not contain carbon monoxide as an impurity at any level. Electrolysis, the alternative process for producing hydrogen using nuclear energy, suffers from thermodynamic inefficiencies in both the production of electricity and in electrolytic parts of the process. The efficiency of electrolysis (electricity to hydrogen) is currently about 80%. Electric power generation efficiency would have to exceed 65% (thermal to electrical) for the combined efficiency to exceed the 52% (thermal to hydrogen) calculated for one thermochemical cycle. Thermochemical water-splitting cycles have been studied, at various levels of effort, for the past 35 years. They were extensively studied in the late 70s and early 80s but have received little attention in the past 10 years, particularly in the U.S. While there is no question about the technical feasibility and the potential for high efficiency, cycles with proven low cost and high efficiency have yet to be developed commercially. Over 100 cycles have been proposed, but substantial research has been executed on only a few. This report describes work accomplished during a three-year project whose objective is to ''define an economically feasible concept for production of hydrogen, by nuclear means, using an advanced high temperature nuclear reactor as the energy source.'' The emphasis of the first phase was to evaluate thermochemical processes which offer the potential for efficient, cost-effective, large-scale production of hydrogen from water in which the primary energy input is high temperature heat from an advanced nuclear reactor and to select one (or, at most three) for further detailed consideration. During Phase 1, an exhaustive literature search was performed to locate all cycles previously proposed. The cycles located were screened using objective criteria to determine which could benefit, in terms of efficiency and cost, from the high-temperature capabilities of advanced nuclear reactors. The more promising cycles were then analyzed in depth as to their adaptability to advanced high-temperature nuclear reactors. As a result, the Sulfur-Iodine (S-I) cycle was selected for integration into the advanced nuclear reactor system. In Phases 2 and 3, alternative flowsheets were developed and compared. This effort entailed a considerable effort into developing the solution thermodynamics pertinent to the S-I cycle.

  9. What to Expect When Readying to Move Spent Nuclear Fuel from Commercial

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

    Nuclear Power Plants | Department of Energy PDF icon What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power Plants More Documents & Publications Nuclear Fuel Storage and Transportation Planning Project Overview Indiana Department of Homeland Security - NNPP Exercise Better Security Through Discussion

  10. Fuel cycle analysis of once-through nuclear systems.

    SciTech Connect (OSTI)

    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 (LEU) fuels. Examples of systems in this class include the small modular reactors being considered internationally; e.g. 4S [Tsuboi 2009], Hyperion Power Module [Deal 2010], ARC-100 [Wade 2010], and SSTAR [Smith 2008]. (2) Systems for Resource Utilization - In recent years, interest has developed in the use of advanced nuclear designs for the effective utilization of fuel resources. Systems under this class have generally utilized the breed and burn concept in which fissile material is bred and used in situ in the reactor core. Due to the favorable breeding that is possible with fast neutrons, these systems have tended to be fast spectrum systems. In the once-through concepts (as opposed to the traditional multirecycle approach typically considered for fast reactors), an ignition (or starter) zone contains driver fuel which is fissile material. This zone is designed to last a long time period to allow the breeding of sufficient fissile material in the adjoining blanket zone. The blanket zone is initially made of fertile depleted uranium fuel. This zone could also be made of fertile thorium fuel or recovered uranium from fuel reprocessing or natural uranium. However, given the bulk of depleted uranium and the potentially large inventory of recovered uranium, it is unlikely that the use of thorium is required in the near term in the U.S. Following the breeding of plutonium or fissile U-233 in the blanket, this zone or assembly then carries a larger fraction of the power generation in the reactor. These systems tend to also have a long cycle length (or core life) and they could be with or without fuel shuffling. When fuel is shuffled, the incoming fuel is generally depleted uranium (or thorium) fuel. In any case, fuel is burned once and then discharged. Examples of systems in this class include the CANDLE concept [Sekimoto 2001], the traveling wave reactor (TWR) concept of TerraPower [Ellis 2010], the ultra-long life fast reactor (ULFR) by ANL [Kim 2010], and the BNL fast mixed spectrum reactor (FMSR) concept [Fisher 1979]. (3) Thermal systems for resource extensio

  11. Safe Advantage on Dry Interim Spent Nuclear Fuel Storage

    SciTech Connect (OSTI)

    Romanato, L.S.

    2008-07-01

    This paper aims to present the advantages of dry cask storage in comparison with the wet storage (cooling water pools) for SNF. When the nuclear fuel is removed from the core reactor, it is moved to a storage unit and it wait for a final destination. Generally, the spent nuclear fuel (SNF) remains inside water pools within the reactors facility for the radioactive activity decay. After some period of time in pools, SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing facilities, or still, wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet facilities, depending on the method adopted by the nuclear power plant or other plans of the country. Interim storage, up to 20 years ago, was exclusively wet and if the nuclear facility had to be decommissioned another storage solution had to be found. At the present time, after a preliminary cooling of the SNF elements inside the water pool, the elements can be stored in dry facilities. This kind of storage does not need complex radiation monitoring and it is safer then wet one. Casks, either concrete or metallic, are safer, especially on occurrence of earthquakes, like that occurred at Kashiwazaki-Kariwa nuclear power plant, in Japan on July 16, 2007. (authors)

  12. Molten tin reprocessing of spent nuclear fuel elements

    DOE Patents [OSTI]

    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.

  13. Detachable connection for a nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Christiansen, David W.; Karnesky, Richard A.

    1986-01-01

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engageable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  14. Detachable connection for a nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Christiansen, D.W.; Karnesky, R.A.

    1983-08-29

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engagable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  15. Integrated Used Nuclear Fuel Storage, Transportation, and Disposal Canister

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

    System - Energy Innovation Portal Storage Energy Storage Electricity Transmission Electricity Transmission Advanced Materials Advanced Materials Find More Like This Return to Search Integrated Used Nuclear Fuel Storage, Transportation, and Disposal Canister System Oak Ridge National Laboratory Contact ORNL About This Technology Publications: PDF Document Publication 11-G00239_ID2603 (2).pdf (847 KB) Technology Marketing Summary Researchers at ORNL have developed an integrated system that

  16. Letter Report: Looking Ahead at Nuclear Fuel Resources

    SciTech Connect (OSTI)

    J. Stephen Herring

    2013-09-01

    The future of nuclear energy and its ability to fulfill part of the worlds energy needs for centuries to come depend on a reliable input of nuclear fuel, either thorium or uranium. Obviously, the present nuclear fuel cycle is completely dependent on uranium. Future thorium cycles will also depend on 235U or fissile isotopes separated from used fuel to breed 232Th into fissile 233U. This letter report discusses several emerging areas of scientific understanding and technology development that will clarify and enable assured supplies of uranium and thorium well into the future. At the most fundamental level, the nuclear energy community needs to appreciate the origins of uranium and thorium and the processes of planetary accretion by which those materials have coalesced to form the earth and other planets. Secondly, the studies of geophysics and geochemistry are increasing understanding of the processes by which uranium and thorium are concentrated in various locations in the earths crust. Thirdly, the study of neutrinos and particularly geoneutrinos (neutrinos emitted by radioactive materials within the earth) has given an indication of the overall global inventories of uranium and thorium, though little indication for those materials locations. Crustal temperature measurements have also given hints of the vertical distribution of radioactive heat sources, primarily 238U and 232Th, within the continental crust. Finally, the evolving technologies for laser isotope separation are indicating methods for reducing the energy input to uranium enrichment but also for tailoring the isotopic vectors of fuels, burnable poisons and structural materials, thereby adding another tool for dealing with long-term waste management.

  17. Method for cleaning solution used in nuclear fuel reprocessing

    DOE Patents [OSTI]

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

    1980-12-17

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

  18. Naval Spent Nuclear Fuel disposal Container System Description Document

    SciTech Connect (OSTI)

    N. E. Pettit

    2001-07-13

    The Naval Spent Nuclear Fuel Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers/waste packages are loaded and sealed in the surface waste handling facilities, transferred underground through the access drifts using a rail mounted transporter, and emplaced in emplacement drifts. The Naval Spent Nuclear Fuel Disposal Container System provides long term confinement of the naval spent nuclear fuel (SNF) placed within the disposal containers, and withstands the loading, transfer, emplacement, and retrieval operations. The Naval Spent Nuclear Fuel Disposal Container System provides containment of waste for a designated period of time and limits radionuclide release thereafter. The waste package maintains the waste in a designated configuration, withstands maximum credible handling and rockfall loads, limits the waste form temperature after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Each naval SNF disposal container will hold a single naval SNF canister. There will be approximately 300 naval SNF canisters, composed of long and short canisters. The disposal container will include outer and inner cylinder walls and lids. An exterior label will provide a means by which to identify a disposal container and its contents. Different materials will be selected for the waste package inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and the natural barrier will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel while the outer cylinder and outer cylinder lids will be made of high-nickel alloy.

  19. Method for cleaning solution used in nuclear fuel reprocessing

    DOE Patents [OSTI]

    Tallent, Othar K. (Oak Ridge, TN); Crouse, David J. (Oak Ridge, TN); Mailen, James C. (Oak Ridge, TN)

    1982-01-01

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

  20. Thermal barrier and support for nuclear reactor fuel core

    DOE Patents [OSTI]

    Betts, Jr., William S. (Del Mar, CA); Pickering, J. Larry (Del Mar, CA); Black, William E. (San Diego, CA)

    1987-01-01

    A thermal barrier/core support for the fuel core of a nuclear reactor having a metallic cylinder secured to the reactor vessel liner and surrounded by fibrous insulation material. A top cap is secured to the upper end of the metallic cylinder that locates and orients a cover block and post seat. Under normal operating conditions, the metallic cylinder supports the entire load exerted by its associated fuel core post. Disposed within the metallic cylinder is a column of ceramic material, the height of which is less than that of the metallic cylinder, and thus is not normally load bearing. In the event of a temperature excursion beyond the design limits of the metallic cylinder and resulting in deformation of the cylinder, the ceramic column will abut the top cap to support the fuel core post.

  1. Report on interim storage of spent nuclear fuel. Midwestern high-level radioactive waste transportation project

    SciTech Connect (OSTI)

    Not Available

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  2. Spent fuel accumulations and arisings in 'fuel user' states: implications for a world nuclear partnership

    SciTech Connect (OSTI)

    Lineberry, M.J.

    2007-07-01

    Consider the question: In a GNEP world, what are the implications of current spent fuel inventories and possible future accumulations in 'fuel user' states? The inventory today ({approx}95,000 MTHM) and the growth rate ({approx}4,000 MTHM/yr) in likely fuel user states is roughly twice that in the U.S., and thus the problem is substantial. The magnitude of the issue increases if, as is expected, fuel user states grow their nuclear capacity. For purposes of illustration, 2% annual growth assumed in this study. There is an implied imperative to remove spent fuel inventories from fuel user states in some reasonable time (if the spent fuel backlog cannot be cleared by 2100, what motivation is there for GNEP today?). Clearing the backlog requires massive reprocessing capacity, but that makes no economic or strategic sense until the advanced burner reactors are developed and deployed. Thus, burner reactor development is the pacing item with any practical GNEP schedule. With regard to economics, cost estimates for reprocessing (which drive key GNEP costs) are much too disparate today. The disparities must be lessened in order to permit rational decision-making. (author)

  3. Used Fuel Disposition Campaign Phase I Ring Compression Testing...

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

    Phase I Ring Compression Testing of High Burnup Cladding Used Fuel Disposition Campaign Phase I Ring Compression Testing of High Burnup Cladding The purpose of ring compression...

  4. VISION -- A Dynamic Model of the Nuclear Fuel Cycle

    SciTech Connect (OSTI)

    J. J. Jacobson; A. M. Yacout; S. J. Piet; D. E. Shropshire; G. E. Matthern

    2006-02-01

    The Advanced Fuel Cycle Initiatives (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. VISION is intended to serve as a broad systems analysis and study tool applicable to work conducted as part of the AFCI (including costs estimates) and Generation IV reactor development studies.

  5. National briefing summaries: Nuclear fuel cycle and waste management

    SciTech Connect (OSTI)

    Schneider, K.J.; Bradley, D.J.; Fletcher, J.F.; Konzek, G.J.; Lakey, L.T.; Mitchell, S.J.; Molton, P.M.; Nightingale, R.E.

    1991-04-01

    Since 1976, the International Program Support Office (IPSO) at the Pacific Northwest Laboratory (PNL) has collected and compiled publicly available information concerning foreign and international radioactive waste management programs. This National Briefing Summaries is a printout of an electronic database that has been compiled and is maintained by the IPSO staff. The database contains current information concerning the radioactive waste management programs (with supporting information on nuclear power and the nuclear fuel cycle) of most of the nations (except eastern European countries) that now have or are contemplating nuclear power, and of the multinational agencies that are active in radioactive waste management. Information in this document is included for three additional countries (China, Mexico, and USSR) compared to the prior issue. The database and this document were developed in response to needs of the US Department of Energy.

  6. Electroslag Strip Cladding of Steam Generators With Alloy 690

    SciTech Connect (OSTI)

    Consonni, M.; Maggioni, F.; Brioschi, F.

    2006-07-01

    The present paper details the results of electroslag cladding and tube-to-tubesheet welding qualification tests conducted by Ansaldo-Camozzi ESC with Alloy 690 (Alloy 52 filler metal) on steel for nuclear power stations' steam generators shell, tubesheet and head; the possibility of submerged arc cladding on first layer was also considered. Test results, in terms of chemical analysis, mechanical properties and microstructure are reproducible and confidently applicable to production cladding and show that electroslag process can be used for Alloy 52 cladding with exceptionally stable and regular operation and high productivity. The application of submerged arc cladding process to the first layer leads to a higher base metal dilution, which should be avoided. Moreover, though the heat affected zone is deeper with electroslag cladding, in both cases no coarsened grain zone is found due to recrystallization effect of second cladding layer. Finally, the application of electroslag process to cladding of Alloy 52 with modified chemical composition, was proved to be highly beneficial as it strongly reduces hot cracking sensitivity, which is typical of submerged arc cladded Alloy 52, both during tube-to-tubesheet welding and first re-welding. (authors)

  7. Managing the Nuclear Fuel Cycle, The Big Picture

    SciTech Connect (OSTI)

    Brett W Carlsen

    2010-07-01

    The nuclear industry, at least in the United States, has failed to deliver on its promise of cheap, abundant energy. After pioneering the science and application and becoming a primary exporter of nuclear technologies, domestic use of nuclear power fell out-of-favor with the public and has been relatively stagnant for several decades. Recently, renewed interest has generated optimism and talk of a nuclear renaissance characterized by a new generation of safe, clean nuclear plants in this country. But, as illustrated by recent policy shifts regarding closure of the fuel cycle and geologic disposal of high-level radioactive wastes, significant hurdles have yet to be overcome. Using the principles of system dynamics, this paper will take a holistic look at the nuclear industry and the interactions between the key players to explore both the intended and unintended consequences of efforts to address the issues that have impeded the growth of the industry and also to illustrate aspects which must be effectively addressed if the renaissance of our industry is to be achieved and sustained.

  8. Verification of the BISON fuel performance code

    SciTech Connect (OSTI)

    D. M. Perez; R. J. Gardner; J. D. Hales; S. R. Novascone; G. Pastore; B. W. Spencer; R. L. Williamson

    2014-09-01

    BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Labo- ratory (USA) since 2009. The code is applicable to both steady and transient fuel behavior and is used to analyze 1D spherical, 2D axisymmetric, or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods and other well known fuel performance codes. Results from several assessment cases are reported, with emphasis on fuel centerline temperatures at various stages of fuel life, fission gas release, and clad deformation during pellet clad mechanical interaction (PCMI). BISON comparisons to fuel centerline temperature measurements are very good at beginning of life and reasonable at high burnup. Although limited to date, fission gas release comparisons are very good. Comparisons of rod diameter following significant power ramping are also good and demonstrate BISONs unique ability to model discrete pellet behavior and accurately predict clad ridging from PCMI.

  9. Fuels

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

    Fuels - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs Advanced Nuclear Energy

  10. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    SciTech Connect (OSTI)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  11. Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

    SciTech Connect (OSTI)

    Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

  12. U.S. Commits $14 million to U.S. - Ukraine Nuclear Fuel Qualification

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

    Project | Department of Energy Commits $14 million to U.S. - Ukraine Nuclear Fuel Qualification Project U.S. Commits $14 million to U.S. - Ukraine Nuclear Fuel Qualification Project March 15, 2007 - 10:55am Addthis KYIV, Ukraine - U.S. Department of Energy Deputy Secretary Clay Sell today announced that the United States will invest $14 million to provide 42 nuclear fuel assemblies to the South Ukraine Nuclear Power Plant under the U.S.-Ukraine Nuclear Fuel Qualification Project (UNFQP). In

  13. EARTHQUAKE CAUSED RELEASES FROM A NUCLEAR FUEL CYCLE FACILITY

    SciTech Connect (OSTI)

    Charles W. Solbrig; Chad Pope; Jason Andrus

    2014-08-01

    The fuel cycle facility (FCF) at the Idaho National Laboratory is a nuclear facility which must be licensed in order to operate. A safety analysis is required for a license. This paper describes the analysis of the Design Basis Accident for this facility. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. The hot cell is used to process spent metallic nuclear fuel. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities.

  14. DOE NE Used Fuel Disposition FY2015 Working Group Presentations

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

    NE Used Fuel Disposition FY2015 Working Group Presentations http://energy.sandia.gov/energy/nuclear-energy/ne-workshops/ufd-working-group-2015/ FOLDER NAME PRESENTATION TITLE AUTHOR Tuesday, June 9, 2015 UFD WG 2015-06-09 Tue Auditorium Afternoon - 1 Quantification of Cation Sorption to Engineered Barrier Materials Under Extreme Conditions Powell UFD WG 2015-06-09 Tue Auditorium Afternoon - 1 Development and Experimental Benchmark of Simulations to Predict Used Nuclear Fuel Cladding Temperatures

  15. High-Burnup BWR Fuel Behavior Under Simulated Reactivity-Initiated Accident Conditions

    SciTech Connect (OSTI)

    Nakamura, Takehiko; Kusagaya, Kazuyuki; Fuketa, Toyoshi; Uetsuka, Hiroshi

    2002-06-15

    Boiling water reactor (BWR) fuel at 56 to 61 GWd/tonne U was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated accident conditions. Current Japanese 8 x 8 type Step II BWR fuel from Fukushima Daini Unit 2 was refabricated to short segments, and thermal energy from 272 to 586 J/g (65 to 140 cal/g) was promptly inserted to the test rods. Cladding deformation of the BWR fuel by the pulse irradiation was smaller than that of pressurized water reactor (PWR) fuels. However, cladding failure occurred in tests with fuel at burnup of 61 GWd/tonne U at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g) during the early stages of transients, while the cladding remained cool. The failure was comparable to the one observed in high-burnup PWR fuel tests, in which embrittled cladding with dense hydride precipitation near the outer surface was fractured due to pellet cladding mechanical interaction. Transient fission gas release by the pulse irradiation was {approx}9.6 to 17% depending on the peak fuel enthalpy.

  16. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    SciTech Connect (OSTI)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-06

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  17. Synergistic Smart Fuel For In-pile Nuclear Reactor Measurements

    SciTech Connect (OSTI)

    James A. Smith; Dale K. Kotter; Randall A. Ali; Steven L . Garrett

    2013-10-01

    In March 2011, an earthquake of magnitude 9.0 on the Richter scale struck Japan with its epicenter on the northeast coast, near the Tohoku region. In addition to the immense physical destruction and casualties across the country, several nuclear power plants (NPP) were affected. It was the Fukushima Daiichi NPP that experienced the most severe and irreversible damage. The earthquake brought the reactors at Fukushima to an automatic shutdown and because the power transmission lines were damaged, emergency diesel generators (EDGs) were activated to ensure that there was continued cooling of the reactors and spent fuel pools. The situation was being successfully managed until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to the reactors.2 At this point, the situation became critical. There was a loss of the sensors and instrumentation within the reactor that could have provided valuable information to guide the operators to make informed decisions and avoid the unfortunate events that followed. In the light of these events, we have developed and tested a potential self-powered thermoacoustic system, which will have the ability to serve as a temperature sensor and can transmit data independently of electronic networks. Such a device is synergistic with the harsh environment of the nuclear reactor as it utilizes the heat from the nuclear fuel to provide the input power.

  18. Microstructural characteristics of PWR spent fuel relative to its leaching behavior

    SciTech Connect (OSTI)

    Wilson, C.N.

    1985-11-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data.

  19. Microstructural characteristics of PWR [pressurized water reactor] spent fuel relative to its leaching behavior

    SciTech Connect (OSTI)

    Wilson, C.N.

    1986-01-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data.

  20. Laser cutting apparatus for nuclear core fuel subassembly

    DOE Patents [OSTI]

    Walch, Allan P. (Manchester, CT); Caruolo, Antonio B. (Vernon, CT)

    1982-02-23

    The object of the invention is to provide a system and apparatus which employs laser cutting to disassemble a nuclear core fuel subassembly. The apparatus includes a gantry frame (C) which straddles the core fuel subassembly (14), an x-carriage (22) travelling longitudinally above the frame which carries a focus head assembly (D) having a vertically moving carriage (46) and a laterally moving carriage (52), a system of laser beam transferring and focusing mirrors carried by the x-carriage and focusing head assembly, and a shroud follower (F) and longitudinal follower (G) for following the shape of shroud (14) to maintain a beam focal point (44) fixed upon the shroud surface for accurate cutting.

  1. Spent nuclear fuel removal program at the West Valley Demonstration Project: Topical report

    SciTech Connect (OSTI)

    Connors, B. J.; Golden, M. P.; Valenti, P. J.; Winkel, J. J.

    1987-03-01

    The spent nuclear fuel removal program at the West Valley Demonstration Project (WVDP) consisted of removing the spent nuclear fuel (SNF) assemblies from the storage pool in the plant, loading them in shielded casks, and preparing the casks for transportation. So far, four fuel removal campaigns have been completed with the return of 625 spent nuclear fuel assemblies to their four utility owners. A fifth campaign, which is not yet completed, will transfer the remaining 125 fuel assemblies to a government site in Idaho. A spent fuel rod consolidation demonstration has been completed, and the storage canisters and their racks are being removed from the fuel receiving and storage pool to make way for installation of the size reduction equipment. A brief history of the West Valley reprocessing plant and the events leading to the storage and ownership of the spent nuclear fuel assemblies and their subsequent removal from West Valley are also recorded as background information. 3 refs., 16 figs., 9 tabs.

  2. Review of Used Nuclear Fuel Storage and Transportation Technical Gap Analyses

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development...

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

    DOE Patents [OSTI]

    Campbell, David O. (Oak Ridge, TN); Buxton, Samuel R. (Wartburg, TN)

    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.

  4. Fundamental aspects of nuclear reactor fuel elements (Technical Report) |

    Office of Scientific and Technical Information (OSTI)

    SciTech Connect elements Citation Details In-Document Search Title: Fundamental aspects of nuclear reactor fuel elements × You are accessing a document from the Department of Energy's (DOE) SciTech Connect. This site is a product of DOE's Office of Scientific and Technical Information (OSTI) and is provided as a public service. Visit OSTI to utilize additional information resources in energy science and technology. A paper copy of this document is also available for sale to the public from

  5. Method for cleaning solution used in nuclear fuel reprocessing

    DOE Patents [OSTI]

    Tallent, Othar K. (Oak Ridge, TN); Dodson, Karen E. (Knoxville, TN); Mailen, James C. (Oak Ridge, TN)

    1983-01-01

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

  6. Closure mechanism and method for spent nuclear fuel canisters

    DOE Patents [OSTI]

    Doman, Marvin J. (Monroeville, PA)

    2004-11-23

    A canister is provided for storing, transporting, and/or disposing of spent nuclear fuel. The canister includes a canister shell, a top shield plug disposed within the canister, and a leak-tight closure arrangement. The closure arrangement includes a shear ring which forms a containment boundary of the canister, and which is welded to the canister shell and top shield plug. An outer seal plate, forming an outer seal, is disposed above the shear ring and is welded to the shield plug and the canister.

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

    DOE Patents [OSTI]

    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.

  8. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    SciTech Connect (OSTI)

    Rodney C. Ewing

    2004-10-07

    Spent nuclear fuel, essentially U{sub 2}, accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO{sub 2} in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term.

  9. Fuel leak detection apparatus for gas cooled nuclear reactors

    DOE Patents [OSTI]

    Burnette, Richard D.

    1977-01-01

    Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

  10. Training implementation matrix, Spent Nuclear Fuel Project (SNFP)

    SciTech Connect (OSTI)

    EATON, G.L.

    2000-06-08

    This Training Implementation Matrix (TIM) describes how the Spent Nuclear Fuel Project (SNFP) implements the requirements of DOE Order 5480.20A, Personnel Selection, Qualification, and Training Requirements for Reactor and Non-Reactor Nuclear Facilities. The TIM defines the application of the selection, qualification, and training requirements in DOE Order 5480.20A at the SNFP. The TIM also describes the organization, planning, and administration of the SNFP training and qualification program(s) for which DOE Order 5480.20A applies. Also included is suitable justification for exceptions taken to any requirements contained in DOE Order 5480.20A. The goal of the SNFP training and qualification program is to ensure employees are capable of performing their jobs safely and efficiently.

  11. OECD/NEA Ongoing activities related to the nuclear fuel cycle

    SciTech Connect (OSTI)

    Cornet, S.M.; McCarthy, K.; Chauvin, N.

    2013-07-01

    As part of its role in encouraging international collaboration, the OECD Nuclear Energy Agency is coordinating a series of projects related to the Nuclear Fuel Cycle. The Nuclear Science Committee (NSC) Working Party on Scientific Issues of the Nuclear Fuel Cycle (WPFC) comprises five different expert groups covering all aspects of the fuel cycle from front to back-end. Activities related to fuels, materials, physics, separation chemistry, and fuel cycles scenarios are being undertaken. By publishing state-of-the-art reports and organizing workshops, the groups are able to disseminate recent research advancements to the international community. Current activities mainly focus on advanced nuclear systems, and experts are working on analyzing results and establishing challenges associated to the adoption of new materials and fuels. By comparing different codes, the Expert Group on Advanced Fuel Cycle Scenarios is aiming at gaining further understanding of the scientific issues and specific national needs associated with the implementation of advanced fuel cycles. At the back end of the fuel cycle, separation technologies (aqueous and pyrochemical processing) are being assessed. Current and future activities comprise studies on minor actinides separation and post Fukushima studies. Regular workshops are also organized to discuss recent developments on Partitioning and Transmutation. In addition, the Nuclear Development Committee (NDC) focuses on the analysis of the economics of nuclear power across the fuel cycle in the context of changes of electricity markets, social acceptance and technological advances and assesses the availability of the nuclear fuel and infrastructure required for the deployment of existing and future nuclear power. The Expert Group on the Economics of the Back End of the Nuclear Fuel Cycle (EBENFC), in particular, is looking at assessing economic and financial issues related to the long term management of spent nuclear fuel. (authors)

  12. Utilization of Used Nuclear Fuel in a Potential Future US Fuel Cycle Scenario - 13499

    SciTech Connect (OSTI)

    Worrall, Andrew

    2013-07-01

    To date, the US reactor fleet has generated approximately 68,000 MTHM of used nuclear fuel (UNF) and even with no new nuclear build in the US, this stockpile will continue to grow at approximately 2,000 MTHM per year for several more decades. In the absence of reprocessing and recycle, this UNF is a liability and needs to be dealt with accordingly. However, with the development of future fuel cycle and reactor technologies in the decades ahead, there is potential for UNF to be used effectively and efficiently within a future US nuclear reactor fleet. Based on the detailed expected operating lifetimes, the future UNF discharges from the existing reactor fleet have been calculated on a yearly basis. Assuming a given electricity demand growth in the US and a corresponding growth demand for nuclear energy via new nuclear build, the future discharges of UNF have also been calculated on a yearly basis. Using realistic assumptions about reprocessing technologies and timescales and which future fuels are likely to be reprocessed, the amount of plutonium that could be separated and stored for future reactor technologies has been determined. With fast reactors (FRs) unlikely to be commercially available until 2050, any new nuclear build prior to then is assumed to be a light water reactor (LWR). If the decision is made for the US to proceed with reprocessing by 2030, the analysis shows that the UNF from future fuels discharged from 2025 onwards from the new and existing fleet of LWRs is sufficient to fuel a realistic future demand from FRs. The UNF arising from the existing LWR fleet prior to 2025 can be disposed of directly with no adverse effect on the potential to deploy a FR fleet from 2050 onwards. Furthermore, only a proportion of the UNF is required to be reprocessed from the existing fleet after 2025. All of the analyses and conclusions are based on realistic deployment timescales for reprocessing and reactor deployment. The impact of the delay in recycling the UNF from the FRs due to time in the core, cooling time, reprocessing, and re-fabrication time is built into the analysis, along with impacts in delays and other key assumptions and sensitivities have been investigated. The results of this assessment highlight how the UNF from future reactors (LWRs and FRs) and the resulting fissile materials (U and Pu) from reprocessing can be effectively utilized, and show that the timings of future nuclear programs are key considerations (both for reactors and fuel cycle facilities). The analysis also highlights how the timings are relevant to managing the UNF and how such an analysis can therefore assist in informing the potential future R and D strategy and needs of the US fuel cycle programs and reactor technology. (authors)

  13. Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from

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

    Decommissioned Reactors | Department of Energy Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors PDF icon Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors More Documents & Publications Information Request, "THE REPORT TO THE PRESIDENT AND THE CONGRESS BY THE SECRETARY OF ENERGY ON THE NEED

  14. Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors

    Energy Savers [EERE]

    6 Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Reactor Sites December 2008 U.S. Department of Energy Office of Civilian Radioactive Waste Management Washington, D.C. Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel The picture on the cover is the Connecticut Yankee Independent Spent Fuel Storage Installation site in Haddam, Connecticut, with 43 dry storage NRC-licensed dual-purpose

  15. Evaluation of Options for Permanent Geologic Disposal of Spent NuclearFuel

    Office of Environmental Management (EM)

    and High-Level Radioactive Waste | Department of Energy Options for Permanent Geologic Disposal of Spent NuclearFuel and High-Level Radioactive Waste Evaluation of Options for Permanent Geologic Disposal of Spent NuclearFuel and High-Level Radioactive Waste This study provides a technical basis for informing policy decisions regarding strategies for the management and permanent disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) in the United States requiring geologic

  16. Strategy for the Management and Disposal of Used Nuclear Fuel and

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

    High-Level Radioactive Waste | Department of Energy Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste Issued on January 11, 2013, the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste is a framework for moving toward a sustainable program to deploy an integrated system capable of transporting, storing, and

  17. Neutronic fuel element fabrication

    DOE Patents [OSTI]

    Korton, George (Cincinnati, OH)

    2004-02-24

    This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure by encompassing the sides of the fuel element between the header plates.

  18. N.R. 20 FOSSIL-FUELED POWER PLANTS; 21 SPECIFIC NUCLEAR REACTORS...

    Office of Scientific and Technical Information (OSTI)

    20 FOSSIL-FUELED POWER PLANTS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 14 SOLAR ENERGY; 15 GEOTHERMAL ENERGY; GEOTHERMAL POWER PLANTS; COMPUTERIZED SIMULATION; HEAT...

  19. EIS-0306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    DOE prepared a EIS that evaluated the potential environmental impacts of treatment and management of DOE-owned sodium bonded spent nuclear fuel.

  20. Locations of Spent Nuclear Fuel and High-Level Radioactive Waste |

    Office of Environmental Management (EM)

    Department of Energy Locations of Spent Nuclear Fuel and High-Level Radioactive Waste Locations of Spent Nuclear Fuel and High-Level Radioactive Waste Map of the United States of America showing the locations of spent nuclear fuel and high-level radioactive waste. PDF icon Slide 1 More Documents & Publications Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel End of Year 2010 SNF & HLW Inventories Evaluation of Options for Permanent

  1. Spent nuclear fuel recycling with plasma reduction and etching

    DOE Patents [OSTI]

    Kim, Yong Ho

    2012-06-05

    A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF6 and PuF4. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF6 (gas) that is formed to decompose back to PuF4 (solid), and b) to maintain stability of the UF6. Uranium (in the form of gaseous UF6) is easily extracted and separated from the plutonium (in the form of solid PuF4). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF6 (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF4 or NF3 as the fluorine sources instead of F2 or HF.

  2. Categorical Exclusion Determinations: Nuclear Energy | Department of Energy

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

    Nuclear Energy Categorical Exclusion Determinations: Nuclear Energy Categorical Exclusion Determinations issued by Nuclear Energy. DOCUMENTS AVAILABLE FOR DOWNLOAD August 26, 2014 CX-012670: Categorical Exclusion Determination Development of Self-Healing Zirconium Silicide Coatings for Improved Performance of Zirconium- Alloy Fuel Cladding - University of Wisconsin CX(s) Applied: B3.6 Date: 41877 Location(s): Wisconsin Offices(s): Nuclear Energy August 26, 2014 CX-012687: Categorical Exclusion

  3. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    SciTech Connect (OSTI)

    Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.; Massaro, Lawrence M.; Jensen, Philip J.

    2014-10-01

    This report presents a preliminary evaluation of removing used nuclear fuel (UNF) from 12 shutdown nuclear power plant sites. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites are Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. The evaluation was divided into four components: characterization of the UNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory; a description of the on-site infrastructure and conditions relevant to transportation of UNF and GTCC waste; an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing UNF and GTCC waste, including identification of gaps in information; and, an evaluation of the actions necessary to prepare for and remove UNF and GTCC waste. The primary sources for the inventory of UNF and GTCC waste are the U.S. Department of Energy (DOE) RW-859 used nuclear fuel inventory database, industry sources such as StoreFUEL and SpentFUEL, and government sources such as the U.S. Nuclear Regulatory Commission. The primary sources for information on the conditions of site and near-site transportation infrastructure and experience included observations and information collected during visits to the Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion sites; information provided by managers at the shutdown sites; Facility Interface Data Sheets compiled for DOE in 2005; Services Planning Documents prepared for DOE in 1993 and 1994; industry publications such as Radwaste Solutions; and Google Earth. State and Regional Group representatives, a Tribal representative, and a Federal Railroad Administration representative participated in six of the shutdown site visits. Every site was found to have at least one off-site transportation mode option for removing its UNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: The 12 shutdown sites use designs from 4 different suppliers involving 9 different (horizontal and vertical) dry storage systems that would require the use of 8 different transportation cask designs to remove the UNF and GTCC waste from the shutdown sites; Although there are common aspects, each site has some unique features and/or conditions; Although some regulatory actions will be required, all UNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in licensed systems that can be transported, including a small amount of high-burnup fuel; Each site indicated that 2-3 years of advance time would be required for its preparations before shipments could begin; Most sites have more than one transportation option, e.g., rail, barge, or heavy haul truck, as well as constraints and preferences. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.

  4. Spent Nuclear Fuel Project path forward: nuclear safety equivalency to comparable NRC-licensed facilities

    SciTech Connect (OSTI)

    Garvin, L.J.

    1995-11-01

    This document includes the Technical requirements which meet the nuclear safety objectives of the NRC regulations for fuel treatment and storage facilities. These include requirements regarding radiation exposure limits, safety analysis, design and construction. This document also includes administrative requirements which meet the objectives of the major elements of the NRC licensing process. These include formally documented design and safety analysis, independent technical review, and oppportunity for public involvement.

  5. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    SciTech Connect (OSTI)

    McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel; Wirth, Brian; Kennedy, Rory

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  6. Neutronics of accelerator-driven subcritical fission for burning transuranics in used nuclear fuel

    SciTech Connect (OSTI)

    Sattarov, A.; Assadi, S.; Badgley, K.; Baty, A.; Comeaux, J.; Gerity, J.; Kellams, J.; Mcintyre, P.; Pogue, N.; Sooby, E.; Tsvetkov, P.; Rosaire, G.; Mann, T.

    2013-04-19

    We report the development of a conceptual design for accelerator-driven subcritical fission in a molten salt core (ADSMS). ADSMS is capable of destroying all of the transuranics at the same rate and proportion as they are produced in a conventional nuclear power plant. The ADSMS core is fueled solely by transuranics extracted from used nuclear fuel and reduces its radiotoxicity by a factor 10,000. ADSMS offers a way to close the nuclear fuel cycle so that the full energy potential in the fertile fuels uranium and thorium can be recovered.

  7. Investigation of Electrochemical Recovery of Zirconium from Spent Nuclear Fuels

    SciTech Connect (OSTI)

    Michael Simpson; II-Soon Hwang

    2014-06-01

    This project uses both modeling and experimental studies to design optimal electrochemical technology methods for recovery of zirconium from used nuclear fuel rods for more effective waste management. The objectives are to provide a means of efficiently separating zirconium into metallic high-level waste forms and to support development of a process for decontamination of zircaloy hulls to enable their disposal as low- and intermediate-level waste. Modeling work includes extension of a 3D model previously developed by Seoul National University for uranium electrorefining by adding the ability to predict zirconium behavior. Experimental validation activities include tests for recovery of zirconium from molten salt solutions and aqueous tests using surrogate materials. *This is a summary of the FY 2013 progress for I-NERI project # 2010-001-K provided to the I-NERI office.

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

    SciTech Connect (OSTI)

    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. Full-scale hot cell test of an acoustic sensor dedicated to measurement of the internal gas pressure and composition of a LWR nuclear fuel rod

    SciTech Connect (OSTI)

    Ferrandis, J. Y.; Rosenkrantz, E.; Leveque, G.; Baron, D.; Segura, J. C.; Cecilia, G.; Provitina, O.

    2011-07-01

    A full-scale hot cell test of the internal gas pressure and composition measurement by an acoustic sensor was carried on successfully between 2008 and 2010 on irradiated fuel rods in the LECA-STAR facility at Cadarache Centre. The acoustic sensor has been specially designed in order to provide a nondestructive technique to easily carry out the measurement of the internal gas pressure and gas composition of a LWR nuclear fuel rod. This sensor has been achieved in 2007 and is now covered by an international patent. The first positive result, concerning the device behaviour, is that the sensor-operating characteristics have not been altered by a two-year exposure in the hot cell ambient. We performed the gas characterisation contained in irradiated fuel rods. The acoustic method accuracy is now {+-}5 bars on the pressure measurement result and {+-}0.3% on the evaluated gas composition. The results of the acoustic method were compared to puncture results. Another significant conclusion is that the efficiency of the acoustic method is not altered by the irradiation time, and possible modification of the cladding properties. These results make it possible to demonstrate the feasibility of the technique on irradiated fuel rods. The transducer and the associated methodology are now operational. (authors)

  10. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    SciTech Connect (OSTI)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  11. Study on Nuclear Fuel Cycle System using Coated-particle Fuel and Hybrid Micro-capsule Separation Method

    SciTech Connect (OSTI)

    Wakabayashi, Toshio; Mimura, Hitoshi

    2007-07-01

    A new concept of nuclear fuel cycle system for the fast reactors was proposed. The concept is mainly composed of a reprocessing and separation system based on hybrid micro-capsules containing extractant, a remote fuel fabrication system for carbide coated particle fuels and a gas-cooled fast reactor. The gas-cooled fast reactor system features a fast-neutron-spectrum helium-cooled reactor and closed fuel cycle to be able to have sustainable energy resources in future. The feasibility study of the concept has been performed from the viewpoints core performance, radioactive waste reduction, separation technology, fabrication technology, etc. It was shown that the new concept of nuclear fuel cycle system including the gas-cooled fast reactor is very attractive to the future energy supply system in this study. (authors)

  12. OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

    SciTech Connect (OSTI)

    Gauld, Ian C; Sly, Nicholas C; Michel-Sendis, Franco

    2014-01-01

    Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

  13. Software Design Document for the AMP Nuclear Fuel Performance Code

    SciTech Connect (OSTI)

    Philip, Bobby; Clarno, Kevin T; Cochran, Bill

    2010-03-01

    The purpose of this document is to describe the design of the AMP nuclear fuel performance code. It provides an overview of the decomposition into separable components, an overview of what those components will do, and the strategic basis for the design. The primary components of a computational physics code include a user interface, physics packages, material properties, mathematics solvers, and computational infrastructure. Some capability from established off-the-shelf (OTS) packages will be leveraged in the development of AMP, but the primary physics components will be entirely new. The material properties required by these physics operators include many highly non-linear properties, which will be replicated from FRAPCON and LIFE where applicable, as well as some computationally-intensive operations, such as gap conductance, which depends upon the plenum pressure. Because there is extensive capability in off-the-shelf leadership class computational solvers, AMP will leverage the Trilinos, PETSc, and SUNDIALS packages. The computational infrastructure includes a build system, mesh database, and other building blocks of a computational physics package. The user interface will be developed through a collaborative effort with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Capability Transfer program element as much as possible and will be discussed in detail in a future document.

  14. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    SciTech Connect (OSTI)

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  15. International Source Book: Nuclear Fuel Cycle Research and Development Vol 1 Volume 1

    SciTech Connect (OSTI)

    Harmon, K. M.; Lakey, L. T.

    1983-07-01

    This document starts with an overview that summarizes nuclear power policies and waste management activities for nations with significant commercial nuclear fuel cycle activities either under way or planned. A more detailed program summary is then included for each country or international agency conducting nuclear fuel cycle and waste management research and development. This first volume includes the overview and the program summaries of those countries listed alphabetically from Argentina to Italy.

  16. EIS-0453: Recapitalization of Infrastructure Supporting Naval Spent Nuclear Fuel Handling at the Idaho National Laboratory

    Broader source: Energy.gov [DOE]

    The Draft EIS evaluates the potential environmental impacts associated with recapitalizing the infrastructure needed to ensure the long-term capability of the Naval Nuclear Propulsion Program (NNPP) to support naval spent nuclear fuel handling capabilities provided by the Expended Core Facility (ECF). Significant upgrades are necessary to ECF infrastructure and water pools to continue safe and environmentally responsible naval spent nuclear fuel handling until at least 2060.

  17. DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel

    Office of Environmental Management (EM)

    Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel October 2014 [This page left blank.] FOREWORD In January 2012, the final report of the Blue Ribbon Commission on America's Nuclear Future urged the Administration to conduct a review of the current policy to dispose of defense and commercial high level radioactive waste and spent nuclear fuel in a single repository or repositories. Based on that recommendation, in January 2013, the Administration's

  18. Characterization of degraded EBR-II fuel from the ICPP-603 basin: National spent nuclear fuel program, FY 1999 final report

    SciTech Connect (OSTI)

    Pahl, R. G.

    2000-04-17

    Characterization data is reported for sodium bonded Experimental Breeder Reactor II (EBR-II) fuel which had been stored underwater in containers since the late 1970's. Sixteen stainless steel storage containers were retrieved from the ICPP-603 storage pool at the Idaho National Engineering and Environmental Laboratory (INEEL) in Idaho. Ten of the containers had leaked water due to improper sealing. In the container chosen for detailed destructive analysis, the stainless steel cladding on the uranium alloy fuel had ruptured and fuel oxide particulate formed and filled the bottom of the container. Headspace gas analysis determined that greater than 99% hydrogen was present. Cesium-137, which had leached out of the fuel during the aqueous corrosion process, dominated the radionuclide source term of the water. The metallic sodium from the fuel element bond had reacted with the water, forming a caustic solution of NaOH.

  19. Nuclear Fuel Cycle Options Evaluation to Inform R&D Planning

    SciTech Connect (OSTI)

    R. Wigeland; T. Taiwo; M. Todosow; H. Ludewig; W. Halsey; J. Gehin; R. Jubin; J. Buelt; S. Stockinger; K. Jenni; B. Oakley

    2014-04-01

    An Evaluation and Screening (E&S) of nuclear fuel cycle options has been conducted in fulfilment of a Charter specified for the study by the U.S. Department of Energy (DOE) Office of Nuclear Energy. The E&S study used an objective and independently reviewed evaluation process to provide information about the potential benefits and challenges that could strengthen the basis and provide guidance for the research and development(R&D) activities undertaken by the DOE Fuel Cycle Technologies Program Office. Using the nine evaluation criteria specified in the Charter and associated evaluation metrics and processes developed during the E&S study, a screening was conducted of 40 nuclear fuel cycle evaluation groups to provide answers to the questions: (1) Which nuclear fuel cycle system options have the potential for substantial beneficial improvements in nuclear fuel cycle performance, and what aspects of the options make these improvements possible? (2)Which nuclear material management approaches can favorably impact the performance of fuel cycle options? (3)Where would R&D investment be needed to support the set of promising fuel cycle system options and nuclear material management approaches identified above, and what are the technical objectives of associated technologies?

  20. Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?

    SciTech Connect (OSTI)

    Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.; Wood, Thomas W.; Perkins, Casey J.

    2011-11-14

    For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may be constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.

  1. Cladding axial elongation models for FRAP-T6. [PWR; BWR

    SciTech Connect (OSTI)

    Shah, V.N.; Carlson, E.R.; Berna, G.A.

    1983-01-01

    This paper presents a description of the cladding axial elongation models developed at the Idaho National Engineering Laboratory (INEL) for use by the FRAP-T6 computer code in analyzing the response of fuel rods during reactor transients in light water reactors (LWR). The FRAP-T6 code contains models (FRACAS-II subcode) that analyze the structural response of a fuel rod including pellet-cladding-mechanical-interaction (PCMI). Recently, four models were incorporated into FRACAS-II to calculate cladding axial deformation: (a) axial PCMI, (b) trapped fuel stack, (c) fuel relocation, and (d) effective fuel thermal expansion. Comparisons of cladding axial elongation measurements from two experiments with the corresponding FRAP-T6 calculations are presented.

  2. Boiling water reactor fuel behavior at burnup of 26 GWd/tonne U under reactivity-initiated accident conditions

    SciTech Connect (OSTI)

    Nakamura, Takehiko; Yoshinaga, Makio . Dept. of Reactor Safety Research); Sobajima, Makoto ); Ishijima, Kiyomi; Fujishiro, Toshio . Dept. of Reactor Safety Research)

    1994-10-01

    Irradiated boiling water reactor (BWR) fuel behavior under reactivity-initiated accident (RIA) conditions was investigated in the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Research Institute. Short test fuel rods, refabricated from a commercial 7 x 7 type BWR fuel rod at a burnup of 26 GWd/ tonne U, were pulse irradiated in the NSRR under simulated cooled startup RIA conditions of the BWRs. Thermal energy from 230 J/g fuel (55 cal/g fuel) to 410 J/g fuel (98 cal/g fuel) was promptly subjected to the test fuel rods by pulse irradiation within [approximately] 10 ms. The peak fuel enthalpies are believed to be the same as the prompt energy depositions. The test fuel rods demonstrated characteristic behavior of the irradiated fuel rods under the accident conditions, such as enhanced pellet cladding mechanical interaction (PCMI) and fission gas release. However, all the fuel rods survived the accident conditions with considerable margins. Simulations by the FRAP-T6 code and fresh fuel rod tests under the same RIA conditions highlighted the burnup effects on the accident fuel performance. The tests and the simulation suggested that the BWR fuel would possibly fail by a cladding burst due to fission gas release during the cladding temperature escalation rather than the PCMI under the cold startup RIA conditions of a severe power burst.

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

    SciTech Connect (OSTI)

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

    2013-10-01

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

  4. 3D Simulation of Missing Pellet Surface Defects in Light Water Reactor Fuel Rods

    SciTech Connect (OSTI)

    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 local defect to be coupled with the global fuel rod model. This approach for modeling fuel with MPS defects is demonstrated and compared with alternative techniques. The effects of varying parameters of the MPS defect are studied using this technique and presented here.

  5. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    SciTech Connect (OSTI)

    Francis, Matthew W.; Weber, Charles F.; Pigni, Marco T.; Gauld, Ian C.

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied as much as 10% of the measured values, and 109Ag was consistently over-predicted by as much as 170%. In general, there is a larger uncertainty for modeling radioactive fission products when compared to either the actinides or the stable fission products in SNF. The relative C/E ratios ranged from a few percent for 137Cs up to 60% and 100% for 106Ru and 125Sb, respectively. Limited or no radioactive fission products data exist in the current data sets for reactor types other than PWRs and BWRs. More work is needed in obtaining a greater diversity of radioactive fission product data. While performing this survey, issues leading to inconsistencies in nuclear fission yield data were discovered that specifically impacted the fission product noble gases. Emphasis was given to this legacy data, and corrective actions were taken as described in this report. After the fission yield data were corrected, the stable xenon and krypton fission products were predicted to within 5% of their measurements. However, preliminary results not explicitly given in this report indicate that the relative C/E ratio for the radioactive isotope 85Kr varied as much as 10%. Due to the complex migration and the difficulty in measuring noble gases in the fuel, a more thorough investigation is needed to understand how accurately depletion codes can calculate these gas concentrations.

  6. The Impacts of Dry-Storage Canister Designs on Spent Nuclear Fuel Handling,

    Office of Environmental Management (EM)

    Storage, Transportation, and Disposal in the U.S. | Department of Energy The Impacts of Dry-Storage Canister Designs on Spent Nuclear Fuel Handling, Storage, Transportation, and Disposal in the U.S. The Impacts of Dry-Storage Canister Designs on Spent Nuclear Fuel Handling, Storage, Transportation, and Disposal in the U.S. PDF icon The Impacts of Dry-Storage Canister Designs on Spent Nuclear Fuel Handling, Storage, Transportation, and Disposal in the U.S. More Documents & Publications

  7. U.S. Global Threat Reduction Initiative/U.S.-Origin Nuclear Fuel Removals |

    Office of Environmental Management (EM)

    Department of Energy Global Threat Reduction Initiative/U.S.-Origin Nuclear Fuel Removals U.S. Global Threat Reduction Initiative/U.S.-Origin Nuclear Fuel Removals Presentation made by Chuck Messick for the NTSF annual meeting held from May 14-16, 2013 in Buffalo, NY. PDF icon U.S. Global Threat Reduction Initiative/U.S.-Origin Nuclear Fuel Removals More Documents & Publications Global Threat Reduction Initiative Global Threat Reduction Initiative Microsoft Word - 25 May Vienna GTR Fact

  8. Numerical simulation of gas dynamics and heat exchange tasks in fuel assemblies of the nuclear reactors

    SciTech Connect (OSTI)

    Zhuchenko, S. V.

    2014-11-12

    This report presents a PC-based program for solution gas dynamics and heat exchange mathematical tasks in fuel assemblies of the fast-neutron nuclear reactors. A fuel assembly consisting of bulk heat-generating elements, which are integrated together by the system of supply and pressure manifolds, is examined. Spherical heat-generating microelements, which contain nuclear fuel, are pulled into the heat-generating elements. Gaseous coolant proceed from supply manifolds to heat-generating elements, where it withdraws the nuclear reaction heat and assembles in pressure manifolds.

  9. Idaho Site Completes Demolition of Cold War-era Nuclear Fuel Reprocessing

    Energy Savers [EERE]

    Facility | Department of Energy Completes Demolition of Cold War-era Nuclear Fuel Reprocessing Facility Idaho Site Completes Demolition of Cold War-era Nuclear Fuel Reprocessing Facility December 22, 2011 - 11:12am Addthis Media Contact Erik Simpson (208) 360-0426 A gravel mound, larger than half a city block and several feet thick, is the only visible feature that remains at the site of a Cold War-era spent nuclear fuel reprocessing facility at the U.S. Department of Energy's Idaho site.

  10. Defective fuel rod detection in operating pressurized water reactors during periods of continuously decreasing fuel rod integrity levels

    SciTech Connect (OSTI)

    Zanker, H. )

    1989-09-01

    Periods of continuously decreasing levels of fuel rod integrity due to debris-induced cladding damage, vibration-induced fretting wear of the cladding, etc. cause difficulties in the assessment of fuel rod performance from coolant activity data. The calculational models currently in use for this purpose in nuclear power plants are not sufficiently capable of indicating cases in which they are invalid. This can mislead reactor operators by misinterpretation of the coolant activity data, especially in situations where fast reactions are necessary. A quick test of validity is suggested to check the applicability of the currently available calculational models for estimating the number and average size of fuel rod defects. This paper describes how to recognize immediately periods of continuously decreasing levels of fuel rod integrity in order to prevent complications in routine power plant maintenance as well as accident situations caused by more severe fuel rod degradation.

  11. Aerogel-clad optical fiber

    DOE Patents [OSTI]

    Sprehn, G.A.; Hrubesh, L.W.; Poco, J.F.; Sandler, P.H.

    1997-11-04

    An optical fiber is surrounded by an aerogel cladding. For a low density aerogel, the index of refraction of the aerogel is close to that of air, which provides a high numerical aperture to the optical fiber. Due to the high numerical aperture, the aerogel clad optical fiber has improved light collection efficiency. 4 figs.

  12. Aerogel-clad optical fiber

    DOE Patents [OSTI]

    Sprehn, Gregory A. (Livermore, CA); Hrubesh, Lawrence W. (Pleasanton, CA); Poco, John F. (Livermore, CA); Sandler, Pamela H. (San Marino, CA)

    1997-01-01

    An optical fiber is surrounded by an aerogel cladding. For a low density aerogel, the index of refraction of the aerogel is close to that of air, which provides a high numerical aperture to the optical fiber. Due to the high numerical aperture, the aerogel clad optical fiber has improved light collection efficiency.

  13. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    SciTech Connect (OSTI)

    N. E. Pettit

    2001-07-13

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident.

  14. Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

    SciTech Connect (OSTI)

    Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J.; Brey, R.F.; Wright, R.N.; Windes, W.F.

    1999-09-03

    Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10{sup 3} and 6 x 10{sup 4} rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10{sup 4} rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10{sup 5} rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance.

  15. Thoria-based cermet nuclear fuel : sintered microsphere fabrication by spray drying.

    SciTech Connect (OSTI)

    Solomon, A.A.; McDeavitt, S.M.; Chandrmouli, V.; Anthonysamy, S.; Kuchibhotla, S.; Downar, T.J.

    2002-01-09

    Cermet nuclear fuels have been demonstrated to have significant potential to enhance fuel performance because of low internal fuel temperatures and low stored energy. The combination of these benefits with the inherent proliferation resistance, high burnup capability, and favorable neutronic properties of the thorium fuel cycle produces intriguing options for advanced nuclear fuel cycles. This paper describes aspects of a Nuclear Energy Research Initiative (NERI) project with two primary goals: (1) Evaluate the feasibility of implementing the thorium fuel cycle in existing or advanced reactors using a zirconium-matrix cermet fuel, and (2) Develop enabling technologies required for the economic application of this new fuel form. Spray drying is a physical process of granulating fine powders that is used widely in the chemical, pharmaceutical, ceramic, and food industries. It is generally used to produce flowable fine powders. Occasionally it is used to fabricate sintered bodies like cemented carbides, but it has not, heretofore, been used to produce sintered microspheres. As a physical process, it can be adapted to many powder types and mixtures and thus, has appeal for nuclear fuels and waste forms of various compositions. It also permits easy recycling of process ''wastes'' and minimal chemical waste streams that can arise in chemical sol/gel processing. On the other hand, for radioactive powders, it presents safety challenges for processing these materials in powder form and in achieving microspheres of high density and perfection.

  16. Preliminary Evaluation of Removing Used Nuclear Fuel From Nine Shutdown Sites

    SciTech Connect (OSTI)

    Maheras, Steven J.; Best, Ralph; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul

    2013-04-30

    The Blue Ribbon Commission on America’s Nuclear Future identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses. In this report, a preliminary evaluation of removing used nuclear fuel from nine shutdown sites was conducted. The shutdown sites included Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion. At these sites a total of 7649 used nuclear fuel assemblies and a total of 2813.2 metric tons heavy metal (MTHM) of used nuclear fuel are contained in 248 storage canisters. In addition, 11 canisters containing greater-than-Class C (GTCC) low-level radioactive waste are stored at these sites. The evaluation was divided in four components: • characterization of the used nuclear fuel and GTCC low-level radioactive waste inventory at the shutdown sites • an evaluation of the onsite transportation conditions at the shutdown sites • an evaluation of the near-site transportation infrastructure and experience relevant to the shipping of transportation casks containing used nuclear fuel from the shutdown sites • an evaluation of the actions necessary to prepare for and remove used nuclear fuel and GTCC low-level radioactive waste from the shutdown sites. Using these evaluations the authors developed time sequences of activities and time durations for removing the used nuclear fuel and GTCC low-level radioactive waste from a single shutdown site, from three shutdown sites located close to each other, and from all nine shutdown sites.

  17. Advanced Nuclear Fuel | Y-12 National Security Complex

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

    refueling. Global Security Cooley joins Y-12's Global Security and Strategic Partnerships Manufacturing and Technical Services Nuclear Material Recovery Nuclear Detection and...

  18. Re-evaluation of Spent Nuclear Fuel Assay Data for the Three...

    Office of Scientific and Technical Information (OSTI)

    Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation Gauld, Ian C. Oak Ridge National Lab. (ORNL), Oak Ridge,...

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

    SciTech Connect (OSTI)

    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.

  20. EA-1117: Management of Spent Nuclear Fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of the proposal for the management of spent nuclear fuel on the U.S. Department of Energy's Oak Ridge Reservation to implement the preferred alternative...

  1. Report of the Fuel Cycle Subcommittee of the Nuclear Energy Advisory Committee

    Broader source: Energy.gov [DOE]

    The Fuel Cycle Subcommittee (FCSC) of NEAC met in Washington, August 17- 19, 2010. DOE’s new science-based approach to all matters related to nuclear energy is being implemented. The general...

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

    DOE Patents [OSTI]

    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. Molten salt considerations for accelerator-driven subcritical fission to close the nuclear fuel cycle

    SciTech Connect (OSTI)

    Sooby, Elizabeth; Baty, Austin; Gerity, James; McIntyre, Peter; Melconian, Karie; Pogue, Nathaniel; Sattarov, Akhdiyor; Adams, Marvin; Tsevkov, Pavel; Phongikaroon, Supathorn; Simpson, Michael; Tripathy, Prabhat

    2013-04-19

    The host salt selection, molecular modeling, physical chemistry, and processing chemistry are presented here for an accelerator-driven subcritical fission in a molten salt core (ADSMS). The core is fueled solely with the transuranics (TRU) and long-lived fission products (LFP) from used nuclear fuel. The neutronics and salt composition are optimized to destroy the transuranics by fission and the long-lived fission products by transmutation. The cores are driven by proton beams from a strong-focusing cyclotron stack. One such ADSMS system can destroy the transuranics in the used nuclear fuel produced by a 1GWe conventional reactor. It uniquely provides a method to close the nuclear fuel cycle for green nuclear energy.

  4. Re-evaluation of Spent Nuclear Fuel Assay Data for the Three...

    Office of Scientific and Technical Information (OSTI)

    DOE PAGES Search Results Accepted Manuscript: Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation This...

  5. Method for forming nuclear fuel containers of a composite construction and the product thereof

    DOE Patents [OSTI]

    Cheng, Bo-Ching; Rosenbaum, Herman S.; Armijo, Joseph S.

    1984-01-01

    An improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof.

  6. Supplement Analysis … Spent Nuclear Fuel and SRS H-Canyon Operations

    Energy Savers [EERE]

    DOE/EIS-0218-SA-07 SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM Highly Enriched Uranium Target Residue Material Transportation U.S. Department of Energy Washington, DC November 2015 DOE/EIS-0218-SA-07 SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM Highly Enriched Uranium Target Residue Material Transportation 1.0 INTRODUCTION The Department of Energy (DOE) has a continuing responsibility for safeguarding

  7. Safety Aspects of Wet Storage of Spent Nuclear Fuel, OAS-L-13-11

    Energy Savers [EERE]

    Safety Aspects of Wet Storage of Spent Nuclear Fuel OAS-L-13-11 July 2013 Department of Energy Washington, DC 20585 July 10, 2013 MEMORANDUM FOR THE SENIOR ADVISOR FOR ENVIRONMENTAL MANAGEMENT FROM: Daniel M. Weeber Assistant Inspector General for Audits and Administration Office of Inspector General SUBJECT: INFORMATION: Audit Report on "Safety Aspects of Wet Storage of Spent Nuclear Fuel" BACKGROUND The Department of Energy (Department) is responsible for managing and storing spent

  8. Literature on fabrication of tungsten for application in pyrochemical processing of spent nuclear fuels

    SciTech Connect (OSTI)

    Edstrom, C.M.; Phillips, A.G.; Johnson, L.D.; Corle, R.R.

    1980-10-11

    The pyrochemical processing of nuclear fuels requires crucibles, stirrers, and transfer tubing that will withstand the temperature and the chemical attack from molten salts and metals used in the process. This report summarizes the literature that pertains to fabrication (joining, chemical vapor deposition, plasma spraying, forming, and spinning) is the main theme. This report also summarizes a sampling of literature on molbdenum and the work previously performed at Argonne National Laboratory on other container materials used for pyrochemical processing of spent nuclear fuels.

  9. FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL

    SciTech Connect (OSTI)

    Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

    2012-10-30

    This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

  10. Technology Implimentation Plan - ATF FeCrAl Cladding for LWR Application

    SciTech Connect (OSTI)

    Snead, Mary A.; Snead, Lance; Terrani, Kurt A.; Field, Kevin G.; Worrall, Andrew; Robb, Kevin R.; Yamamoto, Yukinori; Powers, Jeffrey J.; Dryepondt, Sebastien N.; Pint, Bruce A.; Hu, Xunxiang

    2015-06-01

    Technology implimentation plan for FeCrAl development under the FCRD Advanced Fuel program. The document describes the activities required to get FeCrAl clad ready for LTR testing

  11. Nuclear Power Generation and Fuel Cycle Report 1996

    Reports and Publications (EIA)

    1996-01-01

    This report provides information and forecasts important to the domestic and world nuclear and uranium industries.

  12. Chemical thermodynamic assessment of the Li-U-O system for possible space nuclear applications

    SciTech Connect (OSTI)

    Besmann, T.M.; Cooper, R.H. Jr.

    1985-06-01

    A thermochemical assessment of possible oxide fuel-lithium coolant interactions in conceptual 100-kW(e) space nuclear power reactors has been performed. Results of the evaluation indicate that in the event of a cladding breach the fuel and coolant will interact with extremely negative consequences. The lithium has the potential to reduce the fuel to metallic uranium. Differences in temperature within the coolant loop can drive oxygen and uranium transport processes.

  13. Nuclear power generation and fuel cycle report 1997

    SciTech Connect (OSTI)

    1997-09-01

    Nuclear power is an important source of electric energy and the amount of nuclear-generated electricity continued to grow as the performance of nuclear power plants improved. In 1996, nuclear power plants supplied 23 percent of the electricity production for countries with nuclear units, and 17 percent of the total electricity generated worldwide. However, the likelihood of nuclear power assuming a much larger role or even retaining its current share of electricity generation production is uncertain. The industry faces a complex set of issues including economic competitiveness, social acceptance, and the handling of nuclear waste, all of which contribute to the uncertain future of nuclear power. Nevertheless, for some countries the installed nuclear generating capacity is projected to continue to grow. Insufficient indigenous energy resources and concerns over energy independence make nuclear electric generation a viable option, especially for the countries of the Far East.

  14. Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

    DOE Patents [OSTI]

    Lloyd, M.H.

    1981-01-09

    Method for direct coprocessing of nuclear fuels derived from a product stream of fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

  15. Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

    DOE Patents [OSTI]

    Lloyd, Milton H. (Oak Ridge, TN)

    1983-01-01

    Method for direct coprocessing of nuclear fuels derived from a product stream of a fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

  16. Assessment of MARMOT. A Mesoscale Fuel Performance Code

    SciTech Connect (OSTI)

    Tonks, M. R.; Schwen, D.; Zhang, Y.; Chakraborty, P.; Bai, X.; Fromm, B.; Yu, J.; Teague, M. C.; Andersson, D. A.

    2015-04-01

    MARMOT is the mesoscale fuel performance code under development as part of the US DOE Nuclear Energy Advanced Modeling and Simulation Program. In this report, we provide a high level summary of MARMOT, its capabilities, and its current state of validation. The purpose of MARMOT is to predict the coevolution of microstructure and material properties of nuclear fuel and cladding. It accomplished this using the phase field method coupled to solid mechanics and heat conduction. MARMOT is based on the Multiphysics Object-Oriented Simulation Environment (MOOSE), and much of its basic capability in the areas of the phase field method, mechanics, and heat conduction come directly from MOOSE modules. However, additional capability specific to fuel and cladding is available in MARMOT. While some validation of MARMOT has been completed in the areas of fission gas behavior and grain growth, much more validation needs to be conducted. However, new mesoscale data needs to be obtained in order to complete this validation.

  17. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix D, Part B: Naval spent nuclear fuel management

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    This volume contains the following attachments: transportation of Naval spent nuclear fuel; description of Naval spent nuclear receipt and handling at the Expended Core Facility at the Idaho National Engineering Laboratory; comparison of storage in new water pools versus dry container storage; description of storage of Naval spent nuclear fuel at servicing locations; description of receipt, handling, and examination of Naval spent nuclear fuel at alternate DOE facilities; analysis of normal operations and accident conditions; and comparison of the Naval spent nuclear fuel storage environmental assessment and this environmental impact statement.

  18. Advanced Fuels Campaign FY 2010 Accomplishments Report

    SciTech Connect (OSTI)

    Lori Braase

    2010-12-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) Accomplishment Report documents the high-level research and development results achieved in fiscal year 2010. The AFC program has been given responsibility to develop advanced fuel technologies for the Department of Energy (DOE) using a science-based approach focusing on developing a microstructural understanding of nuclear fuels and materials. The science-based approach combines theory, experiments, and multi-scale modeling and simulation aimed at a fundamental understanding of the fuel fabrication processes and fuel and clad performance under irradiation. The scope of the AFC includes evaluation and development of multiple fuel forms to support the three fuel cycle options described in the Sustainable Fuel Cycle Implementation Plan4: Once-Through Cycle, Modified-Open Cycle, and Continuous Recycle. The word “fuel” is used generically to include fuels, targets, and their associated cladding materials. This document includes a brief overview of the management and integration activities; but is primarily focused on the technical accomplishments for FY-10. Each technical section provides a high level overview of the activity, results, technical points of contact, and applicable references.

  19. Fission gas induced fuel swelling in low and medium burnup fuel during high temperature transients. [PWR

    SciTech Connect (OSTI)

    Vinjamuri, K.

    1980-01-01

    The behavior of light water reactor fuel elements under postulated accident conditions is being studied by the EG and G Idaho, Inc., Thermal Fuels Behavior Program for the Nuclear Regulatory Commission. As a part of this program, unirradiated and previously irradiated, pressurized-water-reactor type fuel rods were tested under power-cooling-mismatch (PCM) conditions in the Power Burst Facility (PBF). During these integral in-reactor experiments, film boiling was produced on the fuel rods which created high fuel and cladding temperatures. Fuel rod diameters increased in the film boiling region to a greater extent for irradiated rods than for unirradiated rods. The purpose of the study was to investigate and assess the fuel swelling which caused the fuel rod diameter increases and to evaluate the ability of an analytical code, the Gas Release and Swelling Subroutine - Steady-State and Transient (GRASS-SST), to predict the results.

  20. Apparatus and method for classifying fuel pellets for nuclear reactor

    DOE Patents [OSTI]

    Wilks, Robert S. (Plum Borough, PA); Sternheim, Eliezer (Pittsburgh, PA); Breakey, Gerald A. (Penn Township, Allegheny County, PA); Sturges, Jr., Robert H. (Plum Borough, PA); Taleff, Alexander (Churchill Borough, PA); Castner, Raymond P. (Monroeville, PA)

    1984-01-01

    Control for the operation of a mechanical handling and gauging system for nuclear fuel pellets. The pellets are inspected for diameters, lengths, surface flaws and weights in successive stations. The control includes, a computer for commanding the operation of the system and its electronics and for storing and processing the complex data derived at the required high rate. In measuring the diameter, the computer enables the measurement of a calibration pellet, stores that calibration data and computes and stores diameter-correction factors and their addresses along a pellet. To each diameter measurement a correction factor is applied at the appropriate address. The computer commands verification that all critical parts of the system and control are set for inspection and that each pellet is positioned for inspection. During each cycle of inspection, the measurement operation proceeds normally irrespective of whether or not a pellet is present in each station. If a pellet is not positioned in a station, a measurement is recorded, but the recorded measurement indicates maloperation. In measuring diameter and length a light pattern including successive shadows of slices transverse for diameter or longitudinal for length are projected on a photodiode array. The light pattern is scanned electronically by a train of pulses. The pulses are counted during the scan of the lighted diodes. For evaluation of diameter the maximum diameter count and the number of slices for which the diameter exceeds a predetermined minimum is determined. For acceptance, the maximum must be less than a maximum level and the minimum must exceed a set number. For evaluation of length, the maximum length is determined. For acceptance, the length must be within maximum and minimum limits.