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Sample records for hafnium clad fuels

  1. EPRI fuel cladding integrity program

    Energy Technology Data Exchange (ETDEWEB)

    Yang, R. [Electric Power Research Institute, Palo Alto, CA (United States)

    1997-01-01

    The objectives of the EPRI fuel program is to supplement the fuel vendor research to assure that utility economic and operational interests are met. To accomplish such objectives, EPRI has conducted research and development efforts to (1) reduce fuel failure rates and mitigate the impact of fuel failures on plant operation, (2) provide technology to extend burnup and reduce fuel cycle cost. The scope of R&D includes fuel and cladding. In this paper, only R&D related to cladding integrity will be covered. Specific areas aimed at improving fuel cladding integrity include: (1) Fuel Reliability Data Base; (2) Operational Guidance for Defective Fuel; (3) Impact of Water Chemistry on Cladding Integrity; (4) Cladding Corrosion Data and Model; (5) Cladding Mechanical Properties; and (6) Transient Fuel Cladding Response.

  2. Advanced Fuels Campaign Cladding & Coatings Meeting Summary

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Experimental assessment of fuel-cladding interactions

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-29

    A range of fuel concepts designed to better tolerate accident scenarios and reactor transients are currently undergoing fundamental development at national laboratories as well as university and industrial partners. Pellet-clad mechanical and chemical interaction can be expected to affect fuel failure rates experienced during steady state operation, as well as dramatically impact the response of the fuel form under loss of coolant and other accident scenarios. The importance of this aspect of fuel design prompted research initiated by AFC in FY14 to begin exploratory efforts to characterize this phenomenon for candidate fuelcladding systems of immediate interest. Continued efforts in FY15 and FY17 aimed to better understand and simulate initial pellet-clad interaction with little-to-no pressure on the pellet-clad interface. Reported here are the results from 1000 h heat treatments at 400, 500, and 600°C of diffusion couples pairing UN with a FeCrAl alloy, SiC, and Zr-based cladding candidate sealed in evacuated quartz ampoules. No gross reactions were observed, though trace elemental contaminants were identified.

  4. Accident-tolerant oxide fuel and cladding

    Science.gov (United States)

    Mariani, Robert D.

    2017-05-30

    Systems and methods for accident tolerant oxide fuel. One or more disks can be placed between fuel pellets comprising UO.sub.2, wherein such disks possess a higher thermal conductivity material than that of the UO.sub.2 to provide enhanced heat rejection thereof. Additionally, a cladding coating comprising zircaloy coated with a material that provides stability and high melting capability can be provided. The pellets can be configured as annular pellets having an annulus filled with the higher thermal conductivity material. The material coating the zircaloy can be, for example, Zr.sub.5Si.sub.4 or another silicide such as, for example, a Zr-Silicide that limits corrosion. The aforementioned higher thermal conductivity material can be, for example, Si, Zr.sub.xSi.sub.y, Zr, or Al.sub.2O.sub.3.

  5. Fuel/cladding compatibility of U-10Zr and U-5Fs fuels with advanced alloy cladding materials

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    Wood, E. L.; Porter, D. L.

    1985-05-01

    This study was concerned with the fuel/cladding interaction observed in diffusion couples of U-10 wt % Zr, U-5 wt % Fs fuels with HT-9, T91, D9, and 316 cladding at 650{sup 0}C and U-10 wt % Zr fuel with the cladding materials at 750{sup 0}C (U-5Fs forms a eutectic with these cladding alloys at temperatures just over 700{sup 0}C). The results presented here represent complete qualitative and semiquantitative findings for diffusion couples exposed for 720 hrs along with preliminary results of 2880 h exposures. The most important result of technological importance to IFR feasibility was that U-10 wt % Zr fuel was found to not significantly interact chemically with the cladding alloys. Moreover, energy-dispersive x-ray analyses demonstrated that metallographically observed banding in the fuel near the fuel/cladding interface was not related to chemical interdiffusion, but was related to zirconium migration towards the fuel/cladding interface. The zirconium migration may be related to oxygen availability at this location. The 2880 h samples were consistent with 720 h findings, indicating that essentially no fuel/cladding interdiffusion occurred, but showing additional zirconium migration.

  6. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

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

    Science.gov (United States)

    Vasil'ev, I. V.; Ivanov, A. S.; Churin, V. A.

    2014-12-01

    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 1500°C. 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.

  8. Development of advanced LWR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H. [and others

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out.

  9. Interdiffusion between U-Zr fuel vs selected cladding steels

    Science.gov (United States)

    Keiser, D. D.; Dayananda, M. A.

    1994-08-01

    To better understand fuel-cladding compatibility issues as affected by diffusion processes in Argonne National Laboratory’s Integral Fast Reactors, interdiffusion studies were carried out with solid-solid diffusion couples assembled with a U-23 at. pct Zr alloy and cladding steels, such as 316, D9, and HT9. All diffusion couples were annealed at 700 °C and examined metallographically and by scanning electron microscopy-energy-dispersive spectroscopy analysis for diffusion structure development. The development of diffusion layers in the couples for various cladding steels is compared and discussed in light of the relative diffusion behavior of the individual elements, intermetallic formation, and experimental diffusion paths. In the context of fuel-cladding compatibility, HT9 is considered superior to 316 and D9, as it develops the smallest diffusion zone with the fewest number of phases.

  10. POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

    Directory of Open Access Journals (Sweden)

    Vojtěch Caha

    2016-12-01

    Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

  11. Characterization Of Cladding Hull Wastes From Used Nuclear Fuels

    Directory of Open Access Journals (Sweden)

    Kang K.H.

    2015-06-01

    Full Text Available Used cladding hulls from pressurized water reactor (PWR are characterized to provide useful information for the treatment and disposal of cladding hull wastes. The radioactivity and the mass of gamma emitting nuclides increases with an increase in the fuel burn-up and their removal ratios are found to be more than 99 wt.% except Co-60 and Cs-137. In the result of measuring the concentrations of U and Pu included in the cladding hull wastes, most of the residues are remained on the surface and the removal ratio of U and Pu are revealed to be over 99.98 wt.% for the fuel burn-up of 35,000 MWd/tU. An electron probe micro-analyzer (EPMA line scanning shows that radioactive fission products are penetrated into the Zr oxide layer, which is proportional to the fuel burn-up. The oxidative decladding process exhibits more efficient removal ratio of radionuclides.

  12. Irradiation experience with HT9-clad metallic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pahl, R.G.; Lahm, C.E.; Tsai, H.; Billone, M.C.

    1991-12-31

    the safe and reliable performance of metallic fuel is currently under study and demonstration in the Integral Fast Reactor program. In-reactor tests of HT9-clad metallic fuel have now reached maturity and have all shown good performance characteristics to burnups exceeding 17.5 at. % in the lead assembly. Because this low-swelling tempered martensitic alloy is the cladding of choice for high fluence applications, the experimental observations and performance modelling efforts reported in this paper play an important role in demonstrating reliability.

  13. Material Selection for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥ 1200°C for short 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. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO2, and therefore Ti2AlC may be challenging to form 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 α´ formation. The composition effects and critical limits to retaining protective scale formation at > 1400°C are still being evaluated.

  14. Clad thickness variation N-Reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Smith, E.A.

    1966-05-12

    The current specifications for the cladding on {open_quotes}N{close_quotes} fuels were established early in the course of process development and were predicted on several basic considerations. Among these were: (a) a desire to provide an adequate safety factor in cladding thickness to insure against corrosion penetration and rupture from uranium swelling stresses; (b) an apprehension that the striations in the zircaloy cladding of the U/zircaloy interface and on the exterior surface might serve as stress-raisers, leading to untimely failures of the jacket; and (c) then existing process capability - the need to maintain a specified ratio between zircaloy and uranium in the billet assembly to effect satisfactory coextrusion. It now appears appropriate to review these specifications in an effort to determine whether some of them may be revised, with attendant gains in economy and/or operating smoothness.

  15. Novel Accident-Tolerant Fuel Meat and Cladding

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO₂ fuel and APMT-Steel Clad/UN-U₃Si₅ fuel concepts

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Galloway, Jack D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    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/UO₂ 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-U₃Si₅ 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Hongbing [Univ. of Texas, Austin, TX (United States); 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

  18. High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Emmanuel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Keiser, Jr., Dennis D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Forsmann, Bryan [Boise State Univ., ID (United States); Janney, Dawn E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Henley, Jody [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Eric C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

  19. A state of the Art report on Manufacturing technology of high burn-up fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyeong Ho; Nam, Cheol; Baek, Jong Hyuk; Choi, Byung Kwon; Park, Sang Yoon; Lee, Myung Ho; Jeong, Yong Hwan

    1999-09-01

    In order to manufacturing the prototype fuel cladding, overall manufacturing processes and technologies should be thoroughly understood on the manufacturing processes and technologies of foreign cladding tubes. Generally, the important technology related to fuel cladding tube manufacturing processes for PWRs/PHWRs is divided into three stages. The first stage is to produce the zirconium sponge from zirconium sand, the second stage is to produce the zircaloy shell or TREX from zirconium sponge ingot and finally, cladding is produced from TREX or zircaloy shell. Therefore, the manufacturing processes including the first and second stages are described in brief in this technology report in order to understand the whole fuel cladding manufacturing processes. (author)

  20. The effect of axial fuel rod power profile on fuel temperature and cladding strain

    Directory of Open Access Journals (Sweden)

    Kim Kyu-Tae

    2010-01-01

    Full Text Available The most limiting design criteria for nuclear reactor normal operating conditions (ANS Condition I are known to be rod internal pressure and cladding oxidation, while those for nuclear reactor transient operating conditions (ANS Conditon II to be fuel centerline temperature and transient cladding total tensile strain. However, the design margins against fuel temperature and transient cladding tensile strain become smaller since power uprating is being or will be utilized for the most of nuclear power reactors to enhance the economics of nuclear power. In order to secure sufficient design margins against fuel temperature and cladding total tensile strain even for power uprating, the current axial rod power profiles used in the reactor transient analysis were optimized to reduce over-conservatism, considering that 118% overpower of a steady-state peak rod average power was not exceeded during the reactor transients. The comparison of the current axial rod power profiles and the optimized ones indicates that the latter reduces the fuel centerline temperature and cladding total tensile strain by 26°C and 0.02%, respectively.

  1. Used Fuel Disposition Campaign - Baseline Studies for Ring Compression Testing of High-Burnup Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Liu, Y. Y. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2012-11-23

    Structural analyses of high-burnup 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 much higher pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). On the basis of previous test results, susceptibility to radial-hydride precipitation depends on cladding material, microstructure, and pre-drying distribution of hydrides across the cladding wall, as well as peak hoop stresses and temperatures during drying operations and storage. Susceptibility to embrittlement depends on the extent of radial-hydride precipitation and the thickness of the outer-surface hydride rim. These results highlight the importance of determining the DBTT for high-burnup cladding as a function of peak drying-storage temperatures and stresses and including the relevant mechanical properties in cask structural analyses. Additional testing is needed at lower (and perhaps more realistic) peak drying-storage temperatures and stresses, for which the DBTT is expected to decrease.

  2. Structural analysis of the SNAP-8 developmental reactor fuel element cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dalcher, A.W.

    1969-04-15

    Primary, secondary, and thermal stresses were calculated and evaluated for the SNAP-8 developmental reactor fuel element cladding. The effects of fabrication and assembly stresses, as well as test and operational stresses were included in the analysis. With the assumption that fuel-swelling-induced stresses are nil, the analytical results indicate that the cladding assembly is structurally adequate for the proposed operation.

  3. Technical basis for storage of Zircaloy-clad spent fuel in inert gases

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.; Gilbert, E.R.

    1983-09-01

    This report summarizes the technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. In addition, dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor (PWR) fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved {similar_to}5,000 fuel rods, and {similar_to}600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570{sup 0}C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at {similar_to}70{sup 0}C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the United States. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380{sup 0}C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400{sup 0}C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved.

  4. Calibration of a fuel-to-cladding gap conductance model for fast reactor fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Baker, R.B.

    1978-05-01

    The report presents refined methods for calculation of fuel temperatures in PuO/sub 2/-UO/sub 2/ fuel in Fast Breeder Reactor (FBR) fuel pins. Of primary concern is the calculation of the temperature changes across the fuel-to-cladding gap of pins with fuel burnups that range from 60 to 10,900 MWd/MTM (0.006 to 1.12 at.%). Described in particular are: (1) a proposed set of heat transfer formulations and corresponding material properties for modeling radial heat transfer through the fuel and cladding; and (2) the calibration of a fuel-to-cladding gap conductance model, as part of a thermal performance computer code (SIEX-M1) which incorporates the proposed heat transfer expressions, using integral thermal performance data from two unique in-reactor experiments. The test data used are from the HEDL P-19 and P-20 experiments which were irradiated in the Experimental Breeder Reactor Number Two (EBR-II), for the Hanford Engineering Development Laboratory (HEDL).

  5. Development of eutectic free cladding materials for metallic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tokiwai, Moriyasu; Yuda, Ryoichi [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan); Ohuchi, Atsushi [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan); Amaya, Masaki [Global Nuclear Fuel-Japan Co., Ltd, Oarai, Ibaraki (Japan)

    2002-11-01

    Historically, it is well known that U base metallic fuel has a lower eutectic temperature with stainless steel cladding. In the phase diagram for the U-Fe binary system, the eutectic temperature is 998K. The eutectic reaction is a limiting factor for raising reactor operation temperature. For the purpose of development of eutectic-free cladding materials, three kinds of diffusion-couple tests with 10 mass%Zr alloy were conducted at a temperature of 1027K for 2250 hrs. We selected the following materials: (a) nitrogen charged zirconium foils, (b) vanadium foils of commercial grade, and (c) nitrogen charged ferritic stainless steel (HT-9). The results showed that typical Zr with layer was observed in all of these materials. Zr with layer appeared to act as a barrier against inter-diffusion of U, Fe. The barrier provided immunity to the eutectic reaction. Discussion was made on C-14 problems in relation to another desirable thermodynamic characteristics of Zr such as carbon-14 immobilization. EPMA analysis indicated relatively high nitrogen concentration at the barrier. The barrier is probably composed of ZrN. (author)

  6. Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

    Energy Technology Data Exchange (ETDEWEB)

    Pasqualini, E.E. [Laboratorio de Nanotecnología Nuclear, Centro Atómico Constituyentes, Comisión Nacional de Energía Atómica, Av. General Paz 1499, B1650KNA, San Martín, Prov. Buenos Aires (Argentina); Robinson, A.B. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Wachs, D.M. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organisation, PMB 1, Menai, NSW, 2234 (Australia)

    2016-10-15

    Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm{sup 3}, 3.8E+21 (peak).

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Investigation of stainless steel clad fuel rod failures and fuel performance in the Connecticut Yankee Reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Pasupathi, V.; Klingensmith, R. W.

    1981-11-01

    Significant levels of fuel rod failures were observed in the batch 8 fuel assemblies of the Connecticut Yankee reactor. Failure of 304 stainless steel cladding in a PWR environment was not expected. Therefore a detailed poolside and hot cell examination program was conducted to determine the cause of failure and identify differences between batch 8 fuel and previous batches which had operated without failures. Hot cell work conducted consisted of detailed nondestructive and destructive examination of fuel rods from batches 7 and 8. The results indicate that the batch 8 failure mechanism was stress corrosion cracking initiating on the clad outer surface. The sources of cladding stresses are believed to be (a) fuel pellet chips wedged in the cladding gap, (b) swelling of highly nondensifying batch 8 fuel and (c) potentially harmful effects of a power change event that occurred near the end of the second cycle of irradiation for batch 8.

  9. Compatibility study between U-UO{sub 2} cermet fuel and T91 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Mishra, Sudhir, E-mail: sudhir@barc.gov.in [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Kaity, Santu; Khan, K.B. [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Sengupta, Pranesh; Dey, G.K. [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2016-12-01

    Cermet is a new fuel concept for the fast reactor system and is ideally designed to combine beneficial properties of both ceramic and metal. In order to understand fuel clad chemical compatibility, diffusion couples were prepared with U-UO{sub 2} cermet fuel and T91 cladding material. These diffusion couples were annealed at 923–1073 K for 1000 h and 1223 K for 50 h, subsequently their microstructures were examined using scanning electron microscope (SEM), X-ray energy dispersive spectroscope (EDS) and electron probe microanalyser (EPMA). It was observed that the interaction between the fuel and constituents of T91 clad was limited to a very small region up to the temperature 993 K and discrete U{sub 6}(Fe,Cr) and U(Fe,Cr){sub 2} intermetallic phases developed. Eutectic microstructure was observed in the reaction zone at 1223 K. The activation energy for reaction at the fuel clad interface was determined.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. Effective thermal conductivity method for predicting spent nuclear fuel cladding temperatures in a dry fill gas

    Energy Technology Data Exchange (ETDEWEB)

    Bahney, Robert

    1997-12-19

    This paper summarizes the development of a reliable methodology for the prediction of peak spent nuclear fuel cladding temperature within the waste disposal package. The effective thermal conductivity method replaces other older methodologies.

  14. Mechanical behavior of fast reactor fuel pin cladding subjected to simulated overpower transients

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, G.D.; Hunter, C.W.

    1978-06-01

    Cladding mechanical property data for analysis and prediction of fuel pin transient behavior were obtained under experimental conditions in which the temperature ramps of reactor transients were simulated. All cladding specimens were 20% CW Type 316 stainless steel and were cut from EBR-II irradiated fuel pins. It was determined that irradiation degraded the cladding ductility and failure strength. Specimens that had been adjacent to the fuel exhibited the poorest properties. Correlations were developed to describe the effect of neutron fluence on the mechanical behavior of the cladding. Metallographic examinations were conducted to characterize the failure mode and to establish the nature of internal and external surface corrosion. Various mechanisms for the fuel adjacency effect were examined and results for helium concentration profiles were presented. Results from the simulated transient tests were compared with TREAT test results.

  15. BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

    Energy Technology Data Exchange (ETDEWEB)

    Sweet, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-30

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. A Multi-Layered Ceramic Composite for Impermeable Fuel Cladding for COmmercial Wate Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feinroth, Herbert

    2008-03-03

    A triplex nuclear fuel cladding is developed to further improve the passive safety of commercial nuclear plants, to increase the burnup and durablity of nuclear fuel, to improve the power density and economics of nuclear power, and to reduce the amount of spent fuel requiring disposal or recycle.

  19. Vanadium as barrier to prevent inter diffusion between metallic fuel and clad material

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Soo; Lee, Seok Hee; Kalita, Deep Jyoti; Woo, Sung Pil; Yoon, Young Soo [Yonsei Univ., Shinchondong, Seoul (Korea, Republic of); Kim, Jun Hwan; Baek, Jong Hyuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Sodium cooled Fast Reactor (SFR) has been considered as next generation nuclear reactor because of its ability of recycling nuclear fuel. Specially, U Zr metal fuel in nuclear reactor has advantages such as ease of fabrication, high thermal conductivity, proliferation resistance and a good stability for sodium which have proven efficient in extending the fusion possibility. In spite of advantages, metal fuel can be inconvenient to use cladding. Actinide elements cause a FCCI (Fuel Clad Chemical Interaction) and eutectic reaction with Fe as nuclear cladding components at just above 650 .deg. C. Since nuclear cladding thickness is decreased during the combusting U Zr metal fuel, the interaction place in the cladding is brittle and less strength. It was reported that the eutectic melting between U Pu Zr and Fe occurs above 650 .deg. C. For such reasons, liner related materials and process have been studied by many research groups. In order to apply this nuclear cladding liner, Zr and V metals show better properties to preventing FCCI. Although liner materials prevent FCCI to an extent, it cannot block it perfectly. In this study, we attempt a combination of vanadium (V) and vanadium foil double layer in between a 420J2(Fe based 12Cr steel) and misch metal. The V thin film was deposited with various RF power. The results of diffusion couple tests at 660 .deg. C for 25 hours showed that a combination of the V thin films and foil exhibited a better shielding for FCCI.

  20. Fuel behavior in severe accidents and Mo-alloy based cladding designs to improve accident tolerance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Bo [Electric Power Research Institute, Palo Alto, CA (United States). Nucler Power Sector

    2013-03-15

    The severe accidents at TMI-2 and Fukushima-Daiichi led to core meltdown and hydrogen explosions. The main source of energy causing core melting is the decay heat from {beta}-, {beta}+, and {gamma} decays of short-lived isotopes following a power scram. The exothermic reaction of Zr-alloy cladding can further increase the cladding temperature leading to rapid cladding corrosion and hydrogen production. The most effective mitigation to minimize core damage in a severe accident is to extend the duration of heat removal capacity via battery-supported passive cooling for as long as practically possible. Replacing the Zr-alloy cladding with a higher heat resistant cladding with lower enthalpy release rate may also provide additional coping time for accident management. Such a heat resistant cladding may also overcome the current licensing concerns about Zr-alloy hydriding and post quench ductility issues in a design base loss of coolant accident (LOCA). Zr-alloy cladding, while has been optimized for normal operation in high pressure water and steam of light water reactors, will rapidly lose its corrosion resistance and tensile and creep strength in high pressure steam. Evaluation of alternate cladding materials and designs have been performed to search for a new fuel cladding design which will substantially improve the safety margins at elevated temperatures during a severe accident, while maintaining the excellent fuel performance attributes of the current Zr-alloy cladding. The screening criteria for the evaluation include neutronic properties, material availability, adaptability and operability in current LWRs, resistance to melting. The new designs also need to be fabricable, maintain sufficient strength and resist to attack by high pressure steam. Engineering metals, alloys and ceramics which can meet some or most of these requirements are limited. Following review of the properties of potential candidates, it is concluded that molybdenum alloys may potentially

  1. Fuel Performance Calculations for FeCrAl Cladding in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    George, Nathan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Sweet, Ryan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Maldonado, G. Ivan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Wirth, Brian D. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.

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

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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.

  3. Study on the Standard Establishment for the Integrity Assessment of Nuclear Fuel Cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S-S; Kim, S-H; Jung, Y-K; Yang, C-Y; Kim, I-G; Choi, Y-H; Kim, H-J; Kim, M-W; Rho, B-H [KINS, Daejeon (Korea, Republic of)

    2008-02-15

    Fuel cladding material plays important role as a primary structure under the high temperature, high pressure and neutron environment of nuclear power plant. According to this environment, cladding material can be experienced several type aging phenomena including the neutron irradiation embrittlement. On the other hand, although the early nuclear power plant was designed to fitting into the 40MWd/KgU burn-up, the currently power plant intends to go to the high burn-up range. In this case, the safety criteria which was established at low burn-up needs to conform the applicability at the high burn-up. In this study, the safety criteria of fuel cladding material was reviewed to assess the cladding material integrity, and the material characteristics of cladding were reviewed. The current LOCA criterial was also reviewed, and the basic study for re-establishment of LOCA criteria was performed. The time concept safety criteria was also discussed to prevent the breakaway oxidation. Through the this study, safety issues will be produced and be helpful for integrity insurance of nuclear fuel cladding material. This report is the final report.

  4. Study on the standard establishment for the integrity assessment of nuclear fuel cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S. S.; Kim, S. H.; Jung, Y. K.; Yang, C. Y.; Kim, I. G.; Choi, Y. H.; Kim, H. J.; Kim, M. W.; Rho, B. H. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2007-02-15

    Fuel cladding material plays important role as a primary structure under the high temperature, high pressure and neutron environment of nuclear power plant. According to this environment, cladding material can be experienced several type aging phenomena including the neutron irradiation embrittlement. On the other hand, although the early nuclear power plant was designed to fitting into the 40MWd/KgU burn-up, the currently power plant intends to go to the high burn-up range. In this case, the safety criteria which was established at low burn-up needs to conform the applicability at the high burn-up. In this study, the safety criteria of fuel cladding material was reviewed to assess the cladding material integrity, and the material characteristics of cladding were reviewed. The current LOCA criterial was also reviewed, and the basic study for re-establishment of LOCA criteria was performed. The time concept safety criteria was also discussed to prevent the breakaway oxidation. Through the this study, safety issues will be produced and be helpful for integrity insurance of nuclear fuel cladding material. This report is 2nd term report.

  5. Allowable peak heat-up cladding temperature for spent fuel integrity during interim-dry storage

    Directory of Open Access Journals (Sweden)

    Ki-Nam Jang

    2017-12-01

    Full Text Available To investigate allowable peak cladding temperature and hoop stress for maintenance of cladding integrity during interim-dry storage and subsequent transport, zirconium alloy cladding tubes were hydrogen-charged to generate 250 ppm and 500 ppm hydrogen contents, simulating spent nuclear fuel degradation. The hydrogen-charged specimens were heated to four peak temperatures of 250°C, 300°C, 350°C, and 400°C, and then cooled to room temperature at cooling rates of 0.3 °C/min under three tensile hoop stresses of 80 MPa, 100 MPa, and 120 MPa. The cool-down specimens showed that high peak heat-up temperature led to lower hydrogen content and that larger tensile hoop stress generated larger radial hydride fraction and consequently lower plastic elongation. Based on these out-of-pile cladding tube test results only, it may be said that peak cladding temperature should be limited to a level < 250°C, regardless of the cladding hoop stress, to ensure cladding integrity during interim-dry storage and subsequent transport.

  6. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2017-04-01

    The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world’s highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form during fabrication and are enhanced during irradiation between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding. One aspect of fuel development and qualification is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 oC). The mechanisms responsible for fission gas release events are discussed.

  7. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

    Science.gov (United States)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2017-04-01

    The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed.

  8. Modelling of the intergranular damage of fuel rod cladding under condition of fuel-cladding interactions; Modelisation de l'endommagement intergranulaire des gaines de combustibles en condition d'interaction pastille-gaine

    Energy Technology Data Exchange (ETDEWEB)

    Diard, O.; Leclercq, S.; Rousselier, G.; Cailletaud, G. [Centre des Materiaux, ENSMP, Evry (France)

    2001-04-01

    The following topics were dealt with: intergranular damage modelling for Zircaloy-4 fuel rod cladding under fuel-cladding interaction conditions, stress corrosion due to fission products (esp. iodine), inhomogeneous mechanical and chemical damaging effects, microstructural simulations with respect to constraint analysis, intergranular microcavitation and brittle phase formation.

  9. Design and fabrication of hafnium tube to control the power of the irradiation test fuel in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H

    2003-05-01

    For the irradiation test at HANARO, non-instrumentation capsule was manufactured and hafnium tube was used to control LHGR of HANARO. Hafnium tube can control the irradiation condition of HANARO similar to that of commercial reactor. Hafnium tube thickness was determined by the LHGR calculated at OR-4 irradiation hole to be installed the non-instrumented capsule. To fabricate the hafnium tube with hafnium plate, the fabrication method was determined by using the hafnium mechanical properties. And the tensile strength of hafnium was confirmed by tensile test. This report is confirmed the LHGR control at the OR-4 and the Hafnium fabrication for in used which the AFPCAP non-instrumented irradiation capsule.

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

    Energy Technology Data Exchange (ETDEWEB)

    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 Fick’s 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Courty, Olivier, E-mail: o.courty@gmail.com [Pennsylvania State University, 45 Bd Gouvion Saint Cyr, 75017 Paris (France); Motta, Arthur T., E-mail: atm2@psu.edu [Department of Mechanical and Nuclear Engineering, 227 Reber Building, Penn State University, University Park, PA 16802 (United States); Hales, Jason D., E-mail: jason.hales@inl.gov [Fuels Modeling and Simulation Department, Idaho National Laboratory (United States)

    2014-09-15

    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 Fick’s 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.

  12. Modelling of pellet-cladding interaction in thermal reactor fuel pins using the Sleuth computer code

    Energy Technology Data Exchange (ETDEWEB)

    Beatham, N.; Hughes, H.; Ellis, W.E.; Shaw, T.L. (AEA Technology, Windscale (UK))

    1990-04-01

    This Paper describes the modelling of pellet-cladding mechanical interaction (PCI) in thermal reactor fuel pins using the Sleuth Computer code. The code is based on the fundamental physical mechanisms causing PCI (differential thermal expansion, fuel swelling, cladding creep-down, etc.) coupled with an estimate of strain concentrations over fuel cracks. It uses the classical 1 1/2 dimensional method which subdivides the fuel both axially and radially. While Sleuth was originally developed to predict PCI failure/survival, it has evolved into a general fuel performance code. The latest version, Sleuth 86, which is a modular form with mnemonic variable names, has proved to be an ideal vehicle for testing new sub-models which have been required as the experimental data base has been expanded. (author).

  13. Hydrides reorientation investigation of high burn-up PWR fuel cladding

    Science.gov (United States)

    Valance, Stéphane; Bertsch, Johannes

    2015-09-01

    The direction of formation of hydride in fuel cladding tube is a major issue for the assessment of the cladding remaining ductility after service. This behavior is quite well known for fresh material, but few results exist for irradiated material. The reorientation behavior of a Zircaloy-4 fuel cladding (AREVA duplex DX-D4) at a burn-up of around 72 GWd t-1 is investigated here. The increase of the fraction of reoriented hydrides through repeated thermo-mechanical loading is inspected; as well, the possibility to recover a state with a minimized quantity of reoriented hydrides is tested using pure thermal loading cycles. The study is completed by a qualitative assessment of the hydrogen density in the duplex layer, where a dependence of the hydrides density on the hoop stress state is observed.

  14. Low temperature chemical processing of graphite-clad nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Robert A.

    2017-10-17

    A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

  15. BISON Investigation of the Effect of the Fuel- Cladding Contact Irregularities on the Peak Cladding Temperature and FCCI Observed in AFC-3A Rodlet 4

    Energy Technology Data Exchange (ETDEWEB)

    Medvedev, Pavel G. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    The primary objective of this report is to document results of BISON analyses supporting Fuel Cycle Research and Development (FCRD) activities. Specifically, the present report seeks to provide explanation for the microstructural features observed during post irradiation examination of the helium-bonded annular U-10Zr fuel irradiated during the AFC-3A experiment. Post irradiation examination of the AFC-3A rodlet revealed microstructural features indicative of the fuel-cladding chemical interaction (FCCI) at the fuel-cladding interface. Presence of large voids was also observed in the same locations. BISON analyses were performed to examine stress and temperature profiles and to investigate possible correlation between the voids and FCCI. It was found that presence of the large voids lead to a formation of circumferential temperature gradients in the fuel that may have redirected migrating lanthanides to the locations where fuel and cladding are in contact. Resulting localized increase of lanthanide concentration is expected to accelerate FCCI. The results of this work provide important guidance to the post irradiation examination studies. Specifically, the hypothesis of lanthanides being redirected from the voids to the locations where the fuel and the cladding are in contact should be verified by conducting quantitative electron microscopy or Electron Probe Micro-Analyzer (EPMA). The results also highlight the need for computer models capable of simulating lanthanide diffusion in metallic fuel and establish a basis for validation of such models.

  16. Construction of in-situ creep strain test facility for the SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Heo, Hyeong Min; Kim, Jun Hwan; Kim, Sung Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, in-situ laser inspection creep test machine was developed for the measuring the creep strain of SFR fuel cladding materials. Ferritic-martensitic steels are being considered as an attractive candidate material for a fuel cladding of a SFR due to their low expansion coefficients, high thermal conductivities and excellent irradiation resistances to a void swelling. HT9 steel (12CrMoVW) is initially developed as a material for power plants in Europe in the 1960. This steel has experienced to expose up to 200dpa in FFTE and EBR-II. Ferritic-Martensitic steel's maximum creep strength in existence is 180Mpa for 106 hour 600 .deg., but HT9 steel is 60Mpa. Because SFR is difficult to secure in developing and applying materials, HT9 steel has accumulated validated data and is suitable for SFR component. And also, because of its superior dimensional stability against fast neutron irradiation, Ferritic-martensitic steel of 9Cr and 12Cr steels, such as HT9 and FC92(12Cr-2W) are preferable to utilize in the fuel cladding of an SFR in KAERI. The pressurized thermal creep test of HT9 and FC92 claddings are being conducted in KAERI, but the change of creep strain in cladding is not easy to measure during the creep test due to its pressurized and closed conditions. In this paper, in-situ laser inspection pressurized creep test machine developed for SFR fuel cladding specimens is described. Moreover, the creep strain rate of HT9 at 650 .deg. C was examined from the in-situ laser inspection pressurized creep test machine.

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

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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. 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). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  19. Surface modification techniques for increased corrosion tolerance of zirconium fuel cladding

    Science.gov (United States)

    Carr, James Patrick, IV

    Corrosion is a major issue in applications involving materials in normal and severe environments, especially when it involves corrosive fluids, high temperatures, and radiation. Left unaddressed, corrosion can lead to catastrophic failures, resulting in economic and environmental liabilities. In nuclear applications, where metals and alloys, such as steel and zirconium, are extensively employed inside and outside of the nuclear reactor, corrosion accelerated by high temperatures, neutron radiation, and corrosive atmospheres, corrosion becomes even more concerning. The objectives of this research are to study and develop surface modification techniques to protect zirconium cladding by the incorporation of a specific barrier coating, and to understand the issues related to the compatibility of the coatings examined in this work. The final goal of this study is to recommend a coating and process that can be scaled-up for the consideration of manufacturing and economic limits. This dissertation study builds on previous accident tolerant fuel cladding research, but is unique in that advanced corrosion methods are tested and considerations for implementation by industry are practiced and discussed. This work will introduce unique studies involving the materials and methods for accident tolerant fuel cladding research by developing, demonstrating, and considering materials and processes for modifying the surface of zircaloy fuel cladding. This innovative research suggests that improvements in the technique to modify the surface of zirconium fuel cladding are likely. Three elements selected for the investigation of their compatibility on zircaloy fuel cladding are aluminum, silicon, and chromium. These materials are also currently being investigated at other labs as alternate alloys and coatings for accident tolerant fuel cladding. This dissertation also investigates the compatibility of these three elements as surface modifiers, by comparing their microstructural and

  20. Innovative coating of nanostructured vanadium carbide on the F/M cladding tube inner surface for mitigating the fuel cladding chemical interactions

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States)

    2017-11-29

    Fuel cladding chemical interactions (FCCI) have been acknowledged as a critical issue in a metallic fuel/steel cladding system due to the formation of low melting intermetallic eutectic compounds between the fuel and cladding steel, resulting in reduction in cladding wall thickness as well as a formation of eutectic compounds that can initiate melting in the fuel at lower temperature. In order to mitigate FCCI, diffusion barrier coatings on the cladding inner surface have been considered. In order to generate the required coating techniques, pack cementation, electroplating, and electrophoretic deposition have been investigated. However, these methods require a high processing temperature of above 700 oC, resulting in decarburization and decomposition of the martensites in a ferritic/martensitic (F/M) cladding steel. Alternatively, organometallic chemical vapor deposition (OMCVD) can be a promising process due to its low processing temperature of below 600 oC. The aim of the project is to conduct applied and fundamental research towards the development of diffusion barrier coatings on the inner surface of F/M fuel cladding tubes. Advanced cladding steels such as T91, HT9 and NF616 have been developed and extensively studied as advanced cladding materials due to their excellent irradiation and corrosion resistance. However, the FCCI accelerated by the elevated temperature and high neutron exposure anticipated in fast reactors, can have severe detrimental effects on the cladding steels through the diffusion of Fe into fuel and lanthanides towards into the claddings. To test the functionality of developed coating layer, the diffusion couple experiments were focused on using T91 as cladding and Ce as a surrogate lanthanum fission product. By using the customized OMCVD coating equipment, thin and compact layers with a few micron between 1.5 µm and 8 µm thick and average grain size of 200 nm and 5 µm were successfully obtained at the specimen coated between 300oC and

  1. FABRICATION AND MATERIAL ISSUES FOR THE APPLICATION OF SiC COMPOSITES TO LWR FUEL CLADDING

    Directory of Open Access Journals (Sweden)

    WEON-JU KIM

    2013-08-01

    Full Text Available The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade SiCf/SiC composites are briefly reviewed. A CVI-processed SiCf/SiC composite with a PyC or (PyC-SiCn interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.

  2. Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

    Directory of Open Access Journals (Sweden)

    Bo Cheng

    2016-02-01

    Full Text Available In severe loss of coolant accidents (LOCA, similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in 1,200–1,500°C steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstrated corrosion resistance. As these composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Mo alloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are

  3. Current status of materials development of nuclear fuel cladding tubes for light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Zhengang, E-mail: duan_zg@imr.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Yang, Huilong [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Satoh, Yuhki [Institute for Materials Research, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Abe, Hiroaki, E-mail: abe.hiroaki@n.t.u-tokyo.ac.jp [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan)

    2017-05-15

    Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

  4. Investigation and basic evaluation for ultra-high burnup fuel cladding material

    Energy Technology Data Exchange (ETDEWEB)

    Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Suga, Masataka [Kokan Keisoku Co., Kawasaki, Kanagawa (Japan)

    2001-03-01

    In ultra-high burnup of the power reactor, it is an essential problem to develop the cladding with excellent durability. First, development history and approach of the safety assessment of Zircaloy for the high burnup fuel were summarized in the report. Second, the basic evaluation and investigation were carried out on the material with high practicability in order to select the candidate materials for the ultra-high burnup fuel. In addition, the basic research on modification technology of the cladding surface was carried out from the viewpoint of the addition of safety margin as a cladding. From the development history of the zirconium alloy including the Zircaloy, it is hard to estimate the results of in-pile test from those of the conventional corrosion test (out-pile test). Therefore, the development of the new testing technology that can simulate the actual environment and the elucidation of the corrosion-controlling factor of the cladding are desired. In cases of RIA (Reactivity Initiated Accident) and LOCA (Loss of Coolant Accident), it seems that the loss of ductility in zirconium alloys under heavy irradiation and boiling of high temperature water restricts the extension of fuel burnup. From preliminary evaluation on the high corrosion-resistance materials (austenitic stainless steel, iron or nickel base superalloys, titanium alloy, niobium alloy, vanadium alloy and ferritic stainless steel), stabilized austenitic stainless steels with a capability of future improvement and high-purity niobium alloys with a expectation of the good corrosion resistance were selected as candidate materials of ultra-high burnup cladding. (author)

  5. Oxidation investigation of cladding specimens for regular and accident tolerant fuel rods under LOCA conditions

    Science.gov (United States)

    Bazyuk, S. S.; Deryabin, I. A.; Kiselev, D. S.; Kuzma-Kichta, Yu A.; Mokrushin, A. A.; Parshin, N. Ya; Popov, E. B.; Soldatkin, D. M.

    2017-11-01

    The high-temperature oxidation tests were carried out for the regular fuel rod claddings specimens made of sponge-based zirconium alloy (E110G) and for the accident tolerant fuel (ATF) ones – pure vacuum melted molybdenum (VCPM) and niobium alloy (Nb-1%Zr). The tests were carried out under the ambient pressure p ∼ 0.1 MPa in pure water steam. The experimental data on the oxidation characteristics were obtained for E110G specimens in the temperature range T = 1100 ‑ 1500 °C, that for VCPM and Nb-1%Zr are investigated under extended temperature-duration range (more than 1 hour). The thermal effects of molybdenum (QSMR) and niobium (QSNR) interactions with steam were defined and the derived oxidation rate constants for refractory metals were compared with the known ones. Based on the computations performed with PARAM-TG code the high-temperature oxidation characteristics of model fuel assemblies of large-scale facilities under LOCA conditions with regular and ATF claddings were compared. It was shown that Zr-steam interaction of fuel rod cladding (QSZR) is more intensive compared with VCPM and Nb-1%Zr ones under investigated conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  7. Modeling of laser cladding with application to fuel cell manufacturing.

    Science.gov (United States)

    2010-01-01

    Polymer electrolyte membrane (PEM) fuel cells have many advantages such as compactness, : lightweight, high power density, low temperature operation and near zero emissions. Although : many research organizations have intensified their efforts toward...

  8. Rod internal pressure of spent nuclear fuel and its effects on cladding degradation during dry storage

    Science.gov (United States)

    Kim, Ju-Seong; Hong, Jong-Dae; Yang, Yong-Sik; Kook, Dong-Hak

    2017-08-01

    Temperature and hoop stress limits have been used to prevent the gross rupture of spent nuclear fuel during dry storage. The stress due to rod internal pressure can induce cladding degradation such as creep, hydride reorientation, and delayed hydride cracking. Creep is a self-limiting phenomenon in a dry storage system; in contrast, hydride reorientation and delayed hydride cracking are potential degradation mechanisms activated at low temperatures when the cladding material is brittle. In this work, a conservative rod internal pressure and corresponding hoop stress were calculated using FRAPCON-4.0 fuel performance code. Based on the hoop stresses during storage, a study on the onset of hydride reorientation and delayed hydride cracking in spent nuclear fuel was conducted under the current storage guidelines. Hydride reorientation is hard to occur in most of the low burn-up fuel while some high burn-up fuel can experience hydride reorientation, but their effect may not be significant. On the other hand, delayed hydride cracking will not occur in spent nuclear fuel from pressurized water reactor; however, there is a lack of confirmatory data on threshold intensity factor for delayed hydride cracking and crack size distribution in the fuel.

  9. Delayed hydride cracking in zircaloy fuel cladding-an IAEA coordinated research programme

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, C. [AECL, Chalk River (Canada); Grigoriev, V. [Studsvik, Nykoeping (Sweden); Inozemtsev, V. [IAEA, Vienna (Austria); Markelov, V. [VNIINM, Moscow (Russian Federation); Roth, M. [INR, Saclay (France); Makarevicius, V. [LEI, Kaunas (Lithuania); Kim, Y. S. [KAERI, Daejeon (Korea, Republic of); Ali, Kanwar Liagat [PINS, TU Wien (Austria); Chakravartty, J. K. [BARC, Mumbai (India); Mizrahi, R. [CNEA, Buenos Aires (Argentina); Lalgudi, R. [IPEN, Rio de Janeiro (Brazil)

    2008-10-15

    The rate of delayed hydride cracking (DHC), V, has been measured in cold worked and stress relieved Zircaloy 4 fuel cladding using the Pin Loading Tension technique. At 250 .deg. C the mean value of V from 69 specimens was 3.3({+-}0.8)x10-8 m/s while the temperature dependence up to 275 .deg. C was described by Aexp(-Q/RT), where Q is 48.3 kJ/mol. No cracking or cracking at very low rates was observed at higher temperatures. The fracture surface consisted of flat fracture with no striations. The results are compared with previous results on fuel cladding and pressure tubes.

  10. Multi-component gas transport in the fuel-to-clad gap of candu fuel rods during severe accidents

    Science.gov (United States)

    Szpunar, B.; Lewis, B. J.; Arimescu, V. I.; Dickson, R. S.; Dickson, L. W.

    2001-04-01

    The multi-component transport of steam, hydrogen and stable fission gas in the fuel-to-clad gap of defective CANDU fuel rods, during severe accident conditions, is investigated in the present work based on a general Stefan-Maxwell treatment. In this analysis, incoming steam must diffuse into a breached rod against a counter-current flow of non-condensable fission gases and out-flowing hydrogen that is produced from the internal reaction of steam with the Zircaloy cladding or urania. A solution of the transport equations yields the local molar distribution of hydrogen and steam so that the internal oxygen potential can be estimated along the length of the gap as a function of time. These equations are numerically solved using finite-difference and control-volume methods with a sparse matrix approach. The model has been used to interpret the fission product release and fuel-rod (i.e., Zircaloy and urania) oxidation behavior observed in high-temperature annealing experiments that were conducted at the Chalk River Laboratories with spent fuel samples with Zircaloy cladding.

  11. Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake: Nuclear fuel durability enhancement.

    Science.gov (United States)

    Škarohlíd, Jan; Ashcheulov, Petr; Škoda, Radek; Taylor, Andrew; Čtvrtlík, Radim; Tomáštík, Jan; Fendrych, František; Kopeček, Jaromír; Cháb, Vladimír; Cichoň, Stanislav; Sajdl, Petr; Macák, Jan; Xu, Peng; Partezana, Jonna M; Lorinčík, Jan; Prehradná, Jana; Steinbrück, Martin; Kratochvílová, Irena

    2017-07-25

    In this work, we demonstrate and describe an effective method of protecting zirconium fuel cladding against oxygen and hydrogen uptake at both accident and working temperatures in water-cooled nuclear reactor environments. Zr alloy samples were coated with nanocrystalline diamond (NCD) layers of different thicknesses, grown in a microwave plasma chemical vapor deposition apparatus. In addition to showing that such an NCD layer prevents the Zr alloy from directly interacting with water, we show that carbon released from the NCD film enters the underlying Zr material and changes its properties, such that uptake of oxygen and hydrogen is significantly decreased. After 100-170 days of exposure to hot water at 360 °C, the oxidation of the NCD-coated Zr plates was typically decreased by 40%. Protective NCD layers may prolong the lifetime of nuclear cladding and consequently enhance nuclear fuel burnup. NCD may also serve as a passive element for nuclear safety. NCD-coated ZIRLO claddings have been selected as a candidate for Accident Tolerant Fuel in commercially operated reactors in 2020.

  12. Synthesis of the Novel MAX Phases for the Future Nuclear Fuel Cladding and Structural Materials

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hyeok [Kyunghee Univ., Yongin (Korea, Republic of); Kim, Taehee; Lee, Taegyu; Ryu, H. J. [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    With these properties, the MAX phases are expected to be used for the Accident Tolerant Fuel (ATF) cladding and oxidation/corrosion resistance materials. Especially, the MAX phase can be used for the Gen-IV, SFR and HTGR, component materials which have to possess the thermal and corrosion resistance. The zirconium has been used to the nuclear industry for fuel cladding because of the small thermal neutron cross-section. Zr-based MAX phase was discovered by group Lapauw et al. They observed the Zr{sub 2}AlC and Zr{sub 3}AlC{sub 2} with the X-ray diffraction (XRD) patterns and backscattered electron detector. Fabrication of the Zr-containing MAX phase was investigated for nuclear fuel cladding and structural materials applications. A MAX phase with the Zr{sub 3}AlC{sub 2} structure was synthesized by spark plasma sintering of a powder mixture targeting (Zr{sub 0.5}Cr{sub 0.5}){sub 4}AlC{sub 3}. The formation of MAX phases was confirmed by XRD and EDS of sintered samples. In the future work, the electron probe micro analyzer (EPMA) and transmission electron microscopy (TEM) are required to certain analyze the elements composition and formation of the MAX phase.

  13. Scratch Behaviors of Cr-Coated Zr-Based Fuel Claddings for Accident-Tolerant Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ho; Kim, Il-Hyun; Kim, Hyun-Gil; Kim, Hyung-Kyu; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As the progression of Fukushima accident is worsened by the runaway reaction at a high temperature above 1200 .deg. C, it is essential to ensure the stabilities of coating layers on conventional Zr-based alloys during normal operations as well as severe accident conditions. This is because the failures of coating layer result in galvanic corrosion phenomenon by potential difference between coating layer and Zr alloy. Also, it is possible to damage the coating layer during handling and manufacturing process by contacting structural components of a fuel assembly. So, adhesion strength is one of the key factors determining the reliability of the coating layer on conventional Zr-based alloy. In this study, two kinds of Cr-coated Zr-based claddings were prepared using arc ion plating (AIP) and direct laser (DL) coating methods. The objective is to evaluate the scratch deformation behaviors of each coating layers on Zr alloys. Large area spallation below normal load of about 15 N appeared to be the predominant mode of failure in the AIP coating during scratch test. However, no tensile crack were found in entire stroke length. In DL coating, small plastic deformation and grooving behavior are more dominant scratching results. It was observed that the change of the slope of the COF curve did not coincide with the failure of coating layer.

  14. Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

    Science.gov (United States)

    2013-06-01

    Be was modeled in SERPENT ; the depletion of Be at 60 MWd/kg in 5.5% 235 U enriched fuel was negligible as the difference between the SERPENT predicted...SIMULATE in the evaluation of core physics performance. 77 Comparison of ENDF-VI based CASMO results with ENDF-VII based SERPENT results for PuO2

  15. PRELIMINARY EVALUATION OF FeCrAl CLADDING AND U-Si FUEL FOR ACCIDENT TOLERANT FUEL CONCEPTS

    Energy Technology Data Exchange (ETDEWEB)

    Hales, J. D.; Gamble, K. A.

    2015-09-01

    Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations and operational transients. This paper presents early work in developing thermal and mechanical models for two materials that may have promise: U-Si for fuel, and FeCrAl for cladding. These materials would not necessarily be used together in the same fuel system, but individually have promising characteristics. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with Zr-4/UO2 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as ductility and thermal conductivity in FeCrAl and U-Si in order to provide guidance on future experiments for these materials.

  16. Evaluation of corrosion on the fuel performance of stainless steel cladding

    Directory of Open Access Journals (Sweden)

    de Souza Gomes Daniel

    2016-01-01

    Full Text Available In nuclear reactors, the use of stainless steel (SS as the cladding material offers some advantages such as good mechanical and corrosion resistance. However, its main advantage is the reduction in the amount of the hydrogen released during loss-of-coolant accident, as observed in the Fukushima Daiichi accident. Hence, research aimed at developing accident tolerant fuels should consider SS as an important alternative to existing materials. However, the available computational tools used to analyze fuel rod performance under irradiation are not capable of assessing the effectiveness of SS as the cladding material. This paper addresses the SS corrosion behavior in a modified fuel performance code in order to evaluate its effect on the global fuel performance. Then, data from the literature concerning to SS corrosion are implemented in the specific code subroutines, and the results obtained are compared to those for Zircaloy-4 (Zy-4 under the same power history. The results show that the effects of corrosion on SS are considerably different from those on Zy-4. The thickness of the oxide layer formed on the SS surface is considerably lower than that formed on Zy-4. As a consequence of this, the global fuel performance of SS under irradiation should be less affected by the corrosion.

  17. Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Weight Reactors

    Science.gov (United States)

    Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

    2017-11-01

    The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

  18. Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

    Science.gov (United States)

    Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

    2018-02-01

    The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

  19. Internal corrosion of EK164 and CHS68 fuel pin cladding steels of uranium dioxide fast power reactor

    Directory of Open Access Journals (Sweden)

    E.A. Kinev

    2015-12-01

    Full Text Available Austenitic chromium-nickel steel EK-164 is the promising material for manufacturing claddings of fuel pins of fast nuclear reactors. Physical and chemical compatibility with typical nuclear fuel compositions on the basis of uranium dioxide pellets is an important aspect of ensuring fuel cladding operability. Post-irradiation examination of irradiated combined fuel assembly with maximum burnup 9.1% FIMA and damaging dose of 77.3dpa equipped with fuel pins with claddings made of CHS-68 and EK-164 steels in cold-worked state was performed. Gamma-scanning, electric potential resistometry and optical metallography methods were applied in the examination. According to the gamma-scanning and resistometry data high-temperature sections of fuel pins are the potential centers of development of fuel pin cladding corrosion. Comparative analysis of internal corrosion of fuel pin claddings made of EK-164 and CHS-68 steels along the reactor core height was performed. In the section with maximum power density at operational temperatures below 540°С depth of corrosion of CHS-68 steel from the side of fuel did not exceed 15µm. On similar sections of fuel pin cladding made of EK-164 steel depth of internal corrosion amounted to 10µm. Maximum of corrosion damage for both steel types was registered at temperatures in the range from 600°С to 650°С. In this case depth of corrosion damages in the form of intercrystalline and general corrosion did not exceed 20µm. No significant differences in the corrosion mechanism between the steels were found. Local exacerbation of corrosion at the junctions between fuel pellets and in places of concentration of cesium fission fragments was detected. And contrariwise in place of narrowing residual gap between fuel and cladding, where cesium is not present, corrosion of EK-164 steel is minimal. Maximum thinning of cladding of the investigated fuel pins with maximum burnup of 9% FIMA amounted to not more than 5% of the

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  1. Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance

    Directory of Open Access Journals (Sweden)

    Cheng Bo

    2016-01-01

    Full Text Available Molybdenum based alloy is selected as a candidate to enhance tolerance of fuel to severe loss of coolant accidents due to its high melting temperature of ∼2600 °C and ability to maintain sufficient mechanical strength at temperatures exceeding 1200 °C. An outer layer of either a Zr-alloy or Al-containing stainless steel is designed to provide corrosion resistance under normal operation and oxidation resistance in steam exceeding 1000 °C for 24 hours under severe loss of coolant accidents. Due to its higher neutron absorption cross-sections, the Mo-alloy cladding is designed to be less than half the thickness of the current Zr-alloy cladding. A feasibility study has been undertaken to demonstrate (1 fabricability of long, thin wall Mo-alloy tubes, (2 formability of a protective outer coating, (3 weldability of Mo tube to endcaps, (4 corrosion resistance in autoclaves with simulated LWR coolant, (5 oxidation resistance to steam at 1000–1500 °C, and (6 sufficient axial and diametral strength and ductility. High purity Mo as well as Mo + La2O3 ODS alloy have been successfully fabricated into ∼2-meter long tubes for the feasibility study. Preliminary results are encouraging, and hence rodlets with Mo-alloy cladding containing fuel pellets have been under preparation for irradiation at the Advanced Test Reactor (ATR in Idaho National Laboratory. Additional efforts are underway to enhance the Mo cladding mechanical properties via process optimization. Oxidation tests to temperatures up to 1500 °C, and burst and creep tests up to 1000 °C are also underway. In addition, some Mo disks in close contact with UO2 from a previous irradiation program (to >100 GWd/MTU at the Halden Reactor have been subjected to post-irradiation examination to evaluate the chemical compatibility of Mo with irradiated UO2 and fission products. This paper will provide an update on results from the feasibility study and discuss the attributes of the

  2. Influence of operating and water-chemistry parameters on fuel cladding corrosion and deposition of corrosion products on cladding surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Kritsky, V.G.; Berezina, I.G.; Rodionov, Y.A., E-mail: kritsky@givnipiet.spb.ru, E-mail: alemaskina@givnipiet.ru [Leading Inst. ' VNIPIET' , Saint Petersburg (Russian Federation)

    2010-07-01

    data available at an NPP and correct the water chemistry so that the pressure drop across the reactor is kept at a stable level by adjusting the concentrations of KOH, H{sub 2}, and NH{sub 3}. The parameters that have been included in the model are the following: operating parameters: reactor thermal power and concentration of boric acid; standards of water chemistry; parameters determining the system redox-potential: concentrations of hydrogen and ammonia; parameters of the physicochemical model of mass transfer; and parameters characterizing the composition of corrosion products in coolant. The deposits along fuel rod bring to sub-boiling and results in acceleration of corrosion products and boron precipitation on the fuel cladding surface increase of nuclide activation period and coolant radioactivity. Activity of Co-58 is the indicator of deposits growth acceleration. (author)

  3. EVALUATION METRICS APPLIED TO ACCIDENT TOLERANT FUEL CLADDING CONCEPTS FOR VVER REACTORS

    Directory of Open Access Journals (Sweden)

    Martin Sevecek

    2016-12-01

    Full Text Available Enhancing the accident tolerance of LWRs became a topic of high interest in many countries after the accidents at Fukushima-Daiichi. Fuel systems that can tolerate a severe accident for a longer time period are referred as Accident Tolerant Fuels (ATF. Development of a new ATF fuel system requires evaluation, characterization and prioritization since many concepts have been investigated during the first development phase. For that reason, evaluation metrics have to be defined, constraints and attributes of each ATF concept have to be studied and finally rating of concepts presented. This paper summarizes evaluation metrics for ATF cladding with a focus on VVER reactor types. Fundamental attributes and evaluation baseline was defined together with illustrative scenarios of severe accidents for modeling purposes and differences between PWR design and VVER design.

  4. A cone beam computed tomography inspection method for fuel rod cladding tubes

    Science.gov (United States)

    Fu, Jian; Tan, Renbo; Wang, Qianli; Deng, Jingshan; Liu, Ming

    2012-10-01

    Fuel rods in nuclear power plants consist of UO2 pellets enclosed in Zirconium alloy (Zircaloy) cladding tube, which is composed of a body and a plug. The body is manufactured separately from the plug and, before its use, the plug is welded with the body. It is vitally important for the welding zone to remain free from defects after the fuel pellets are loaded into the cladding tube to prevent the radioactive fission products from leaking. X-ray computed tomography (CT) is in principle a feasible inspection method for the welding zone, but it faces several challenges due to the high attenuation of Zircaloy. In this paper, a cone beam CT method is proposed to address these issues and perform the welding flaw inspection. A Zircaloy compensator is adopted to narrow the signal range, a structure-based background removal technique to reveal the defects, a linear extension technique to determine the reference X-ray intensity signal and FDK algorithm to reconstruct the slice images. A prototype system, based on X-ray tube source and flat panel detector, has been developed and the experiments in this system have demonstrated that the welding void and the incomplete joint penetrations could be detected by this method. This approach may find applications in the quality control of nuclear fuel rods.

  5. FABRICATION AND MATERIAL ISSUES FOR THE APPLICATION OF SiC COMPOSITES TO LWR FUEL CLADDING

    OpenAIRE

    Kim, Weon-Ju; Kim, Daejong; Park, Ji Yeon

    2013-01-01

    The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade SiCf/SiC composites are briefly reviewed. A CVI-processed SiCf/SiC composite with a PyC or (PyC-SiC)n interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs t...

  6. ODS Ferritic/martensitic alloys for Sodium Fast Reactor fuel pin cladding

    Science.gov (United States)

    Dubuisson, Philippe; Carlan, Yann de; Garat, Véronique; Blat, Martine

    2012-09-01

    The development of ODS materials for the cladding for Sodium Fast Reactors is a key issue to achieve the objectives required for the GEN IV reactors. CEA, AREVA and EDF have launched in 2007 an important program to determine the optimal fabrication parameters, and to measure and understand the microstructure and properties before, under and after irradiation of such cladding materials. The aim of this paper is to present the French program and the major results obtained recently at CEA on Fe-9/14/18Cr1WTiY2O3 ferritic/martensitic ODS materials. The first step of the program was to consolidate Fe-9/14/18Cr ODS materials as plates and bars to study the microstructure and the mechanical properties of the new alloys. The second step consists in producing tubes at a geometry representative of the cladding of new Sodium Fast Reactors. The optimization of the fabrication route at the laboratory scale is conducted and different tubes were produced. Their microstructure depends on the martensitic (Fe-9Cr) or ferritic (Fe-14Cr) structure. To join the plug to the tube, the reference process is the welding resistance. A specific approach is developed to model the process and support the development of the welds performed within the "SOPRANO" facility. The development at CEA of Fe-9/14/18Cr new ODS materials for the cladding for GENIV Sodium Fast Reactors is in progress. The first microstructural and mechanical characterizations are very encouraging and the full assessment and qualification of this new alloys and products will pass through the irradiation of specimens, tubes, fuel pins and subassemblies up to high doses.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

    2014-09-30

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

  8. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

    2016-05-01

    Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.

  9. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, W.J., E-mail: jon.carmack@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415 (United States); Chichester, H.M., E-mail: heather.chichester@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415 (United States); Porter, D.L., E-mail: douglas.porter@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415 (United States); Wootan, D.W., E-mail: david.wootan@pnnl.gov [Pacific Northwest National Laboratory, PO Box 999, Richland, WA 99354 (United States)

    2016-05-15

    The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data. - Highlights: • Irradiation and post irradiation examination of full-length metallic fast reactor fuel. • Fuel cladding chemical interaction formation in full-length metallic fast reactor fuel. • Correlation of FCCI with temperature and burnup. • Comparison of full-length reactor fuel performance with test reactor fuel performance.

  10. Development of Modified Ring Tensile Test Technique for Fuel Cladding in Hot Cell (II)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Sik; Ahn, S. B.; Oh, W. H.; Yoo, B. O.; Choo, Y. S

    2005-12-15

    The modified ring tensile test technique was proposed in order to evaluate mechanical properties of fuel cladding under hoop loading condition in hot cell. The hoop loading grip for the modified ring tensile test is designed such that a constant specimen curvature is maintained during deformation, and the gage section of ring specimen is located at the top of the half-cylinder({phi}8.08 mm). The interface between the outer surface of the half-cylinder and the inner surface of the ring specimen was lubricated by graphite lubricant(Molykote P37) in order to minimize the friction between this contact surface. The ring specimen design for ring tensile test is conducted to limit deformation within the gauge section and to maximize uniformity of strain distribution. The dimensions of the ring specimen are 5 mm in ring width, 3 mm in gage length, 2 mm in width of the gage section and 1 mm in radius of the shoulder part. The specially designed precision grinding machine was developed to machine the gage section at the ring segment with 5 mm in length, and the optimum machining conditions are determined. From the comparisons between the test results in this study and the other researcher's test results, we ensure that the proposed ring tensile test technique is suitable to evaluate the mechanical properties of fuel cladding in hoop direction quantitatively.

  11. Sensitivity analysis of FeCrAl cladding and U3Si2 fuel under accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, Jason Dean [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    The purpose of this milestone report is to highlight the results of sensitivity analyses performed on two accident tol- erant fuel concepts: U3Si2 fuel and FeCrAl cladding. The BISON fuel performance code under development at Idaho National Laboratory was coupled to Sandia National Laboratories’ DAKOTA software to perform the sensitivity analyses. Both Loss of Coolant (LOCA) and Station blackout (SBO) scenarios were analyzed using main effects studies. The results indicate that for FeCrAl cladding the input parameters with greatest influence on the output metrics of interest (fuel centerline temperature and cladding hoop strain) during the LOCA were the isotropic swelling and fuel enrichment. For U3Si2 the important inputs were found to be the intergranular diffusion coefficient, specific heat, and fuel thermal conductivity. For the SBO scenario, Young’s modulus was found to be influential in FeCrAl in addition to the isotropic swelling and fuel enrichment. Contrarily to the LOCA case, the specific heat of U3Si2 was found to have no effect during the SBO. The intergranular diffusion coefficient and fuel thermal conductivity were still found to be of importance. The results of the sensitivity analyses have identified areas where further research is required including fission gas behavior in U3Si2 and irradiation swelling in FeCrAl. Moreover, the results highlight the need to perform the sensitivity analyses on full length fuel rods for SBO scenarios.

  12. 2nd Gen FeCrAl ODS Alloy Development For Accident-Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Massey, Caleb P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Edmondson, Philip D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    Extensive research at ORNL aims at developing advanced low-Cr high strength FeCrAl alloys for accident tolerant fuel cladding. One task focuses on the fabrication of new low Cr oxide dispersion strengthened (ODS) FeCrAl alloys. The first Fe-12Cr-5Al+Y2O3 (+ ZrO2 or TiO2) ODS alloys exhibited excellent tensile strength up to 800 C and good oxidation resistance in steam up to 1400 C, but very limited plastic deformation at temperature ranging from room to 800 C. To improve alloy ductility, several fabrication parameters were considered. New Fe-10-12Cr-6Al gas-atomized powders containing 0.15 to 0.5wt% Zr were procured and ball milled for 10h, 20h or 40h with Y2O3. The resulting powder was then extruded at temperature ranging from 900 to 1050 C. Decreasing the ball milling time or increasing the extrusion temperature changed the alloy grain size leading to lower strength but enhanced ductility. Small variations of the Cr, Zr, O and N content did not seem to significantly impact the alloy tensile properties, and, overall, the 2nd gen ODS FeCrAl alloys showed significantly better ductility than the 1st gen alloys. Tube fabrication needed for fuel cladding will require cold or warm working associated with softening heat treatments, work was therefore initiated to assess the effect of these fabrications steps on the alloy microstructure and properties. This report has been submitted as fulfillment of milestone M3FT 16OR020202091 titled, Report on 2nd Gen FeCrAl ODS Alloy Development for the Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program.

  13. Eutectic reaction analysis between TRU-50%Zr alloy fuel and HT-9 cladding, and temperature prediction of eutectic reaction under steady-state

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok

    2001-02-01

    Blanket fuel assembly for HYPER contains a bundle of pins arrayed in triangular pitch, which has hexagonal bundle structure. The reference blanket fuel pin consists of the fuel slug of TRU-50wt%Zr alloy and the cladding material of ferritic martensite steel, HT-9. Chemical interaction between fuel slug and cladding is one of the major concerns in metallic fuel rod design. The contact of metallic fuel slug and stainless steel cladding in a fuel rod forms a complex multi-component diffusion couple at elevated temperatures. The potential problem of inter-diffusion of fuel and cladding components is essentially two-fold weakening of cladding mechanical strength due to the formation of diffusion zones in the cladding, and the formation of comparatively low melting point phases in the fuel/cladding interface to develop eutectic reaction. The main components of fuel slug are composed of zirconium alloying element in plutonium matrix, including neptunium, americium and uranium additionally. Therefore basic eutectic reaction change of Pu-Fe binary system can be assessed, while it is estimated how much other elements zirconium, uranium, americium and neptunium influence on plutonium phase stability. Afterwards it is needed that eutectic reaction is verified through experimental necessarily.

  14. Design of the Capsule (13M-01K) for Irradiation of Fuel Cladding Materials in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Choo, Kee Nam; Yang, Seong Woo; Kang, Young Hwan; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Nuclear-grade zirconium alloys contain more than 95% Zr, and therefore most of their properties are similar to those of pure zirconium. ZIRLO material, used in fuel rod cladding, structural and flow mixing grids, instrumentation tubes, and guide thimbles, increases margin-to-fuel-rod-corrosion limits and enhances fuel assembly structural stability. The demonstrated corrosion resistance and enhanced structural stability of ZIRLO cladding enable longer cycle lengths at higher temperatures without reducing operating margins. An instrumented capsule (13M-01K) was designed and fabricated for evaluation of the neutron irradiation properties of Zirlo material, which is commonly used for cladding of nuclear fuel. This capsule is now being irradiated for 2 cycles at CT test hole of HANARO, which was started at Jan 27 and will be ended at Mar 31, 2014. The structure of the capsule was based on the previous capsule (11M-22K capsule) which was successfully irradiated at the same hole of HANARO. In the capsule, 182 specimens such as tensile specimens of plate type and ring type specimens were placed. Most of them are made of Zirlo, but a few are HANA material that is developed in KAERI. The irradiation test was requested by 4 universities including Dong-Kook and Han-Yang etc. The capsule is composed of 5 layers, each of which had Al holder containing several specimens and an independent electric heater, thermocouples etc. During the irradiation test, temperatures of the specimens and fast neutron fluence were measured by 14 thermocouples and 5 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. The capsule is irradiated for 2 cycles (28 days) at the CT test hole of HANARO of a 30MW thermal output at 350-390 .deg. C up to a fast neutron fluence of 9.6Χ10{sup 20} (n/cm{sup 2}) (E>1.0 Mev). In this capsule, the two kinds of irradiation tests are performed at quite different temperatures. At upper 3 layers of the capsule, specimens for irradiation at high

  15. Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

    Directory of Open Access Journals (Sweden)

    Shengli Chen

    2017-01-01

    Full Text Available Neutronic performance is investigated for a potential accident tolerant fuel (ATF, which consists of U3Si2 fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2 has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly. These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod, and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U3Si2-FeCrAl system is a potential ATF candidate from a neutronic view.

  16. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  17. Evaluation of tantalum-alloy-clad uranium mononitride fuel specimens from 7500-hour, 1040 C pumped-lithium-loop test

    Science.gov (United States)

    Watson, G. K.

    1974-01-01

    Simulated nuclear fuel element specimens, consisting of uranium mononitride (UN) fuel cylinders clad with tungsten-lined T-111, were exposed for up to 7500 hr at 1040 C (1900 F) in a pumped-lithium loop. The lithium flow velocity was 1.5 m/sec (5 ft/sec) in the specimen test section. No evidence of any compatibility problems between the specimens and the flowing lithium was found based on appearance, weight change, chemistry, and metallography. Direct exposure of the UN to the lithium through a simulated cladding crack resulted in some erosion of the UN in the area of the defect. The T-111 cladding was ductile after lithium exposure, but it was sensitive to hydrogen embrittlement during post-test handling.

  18. Non-destructive evaluation of the cladding thickness in LEU fuel plates by accurate ultrasonic scanning technique

    Energy Technology Data Exchange (ETDEWEB)

    Borring, J.; Gundtoft, H.E.; Borum, K.K.; Toft, P. [Riso National Lab. (Denmark)

    1997-08-01

    In an effort to improve their ultrasonic scanning technique for accurate determination of the cladding thickness in LEU fuel plates, new equipment and modifications to the existing hardware and software have been tested and evaluated. The authors are now able to measure an aluminium thickness down to 0.25 mm instead of the previous 0.35 mm. Furthermore, they have shown how the measuring sensitivity can be improved from 0.03 mm to 0.01 mm. It has now become possible to check their standard fuel plates for DR3 against the minimum cladding thickness requirements non-destructively. Such measurements open the possibility for the acceptance of a thinner nominal cladding than normally used today.

  19. Advanced materials for solid oxide fuel cells: Hafnium-Praseodymium-Indium Oxide System

    Energy Technology Data Exchange (ETDEWEB)

    Bates, J.L.; Griffin, C.W.; Weber, W.J.

    1988-06-01

    The HfO/sub 2/-PrO/sub 1.83/-In/sub 2/O/sub 3/ system has been studied at the Pacific Northwest Laboratory to develop alternative, highly electrically conducting oxides as electrode and interconnection materials for solid oxide fuel cells. A coprecipitation process was developed for synthesizing single-phase, mixed oxide powders necessary to fabricate powders and dense oxides. A ternary phase diagram was developed, and the phases and structures were related to electrical transport properties. Two new phases, an orthorhombic PrInO/sub 3/ and a rhombohedral Hf/sub 2/In/sub 2/O/sub 7/ phase, were identified. The highest electronic conductivity is related to the presence of a bcc, In/sub 2/O/sub 3/ solid solution (ss) containing HfO/sub 2/ and PrO/sub 1.83/. Compositions containing more than 35 mol % of the In/sub 2/O/sub 3/ ss have electrical conductivities greater than 10/sup /minus/1/ (ohm-cm)/sup /minus/1/, and the two or three phase structures that contain this phase appear to exhibit mixed electronic-ionic conduction. The high electrical conductivities and structures similar to the Y/sub 2/O/sub 3/-stabilized ZrO/sub 2/(HfO/sub 2/) electrolyte give these oxides potential for use as cathodes in solid oxide fuel cells. 21 refs.

  20. Nanoindentation measurements of the mechanical properties of zirconium matrix and hydrides in unirradiated pre-hydrided nuclear fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Rico, A., E-mail: alvaro.rico@urjc.es [DIMME, Departamento de Tecnología Mecánica, Universidad Rey Juan Carlos, c/Tulipán s/n, E-28933 Móstoles, Madrid (Spain); Martin-Rengel, M.A., E-mail: mamartin@mater.upm.es [Departamento de Ciencia de los Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren SN, E-28040 Madrid (Spain); Ruiz-Hervias, J., E-mail: jesus.ruiz@upm.es [Departamento de Ciencia de los Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren SN, E-28040 Madrid (Spain); Rodriguez, J. [DIMME, Departamento de Tecnología Mecánica, Universidad Rey Juan Carlos, c/Tulipán s/n, E-28933 Móstoles, Madrid (Spain); Gomez-Sanchez, F.J., E-mail: javier.gomez@amsimulation.com [Advanced Material Simulation, S.L, Madrid (Spain)

    2014-09-15

    It is well known that the mechanical properties of the nuclear fuel cladding may be affected by the presence of hydrides. The average mechanical properties of hydrided cladding have been extensively investigated from a macroscopic point of view. In addition, the mechanical and fracture properties of bulk hydride samples fabricated from zirconium plates have also been reported. In this paper, Young’s modulus, hardness and yield stress are measured for each phase, namely zirconium hydrides and matrix, of pre-hydrided nuclear fuel cladding. To this end, nanoindentation tests were performed on ZIRLO samples in as-received state, on a hydride blister and in samples with 150 and 1200 ppm of hydrogen homogeneously distributed along the hoop direction of the cladding. The results show that the measured mechanical properties of the zirconium hydrides and ZIRLO matrix (Young’s modulus, hardness and yield stress) are rather similar. From the experimental data, the hydride volume fraction in the cladding samples with 150 and 1200 ppm was estimated and the average mechanical properties were calculated by means of the rule of mixtures. These values were compared with those obtained from ring compression tests. Good agreement between the results obtained by both methods was found.

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

    Energy Technology Data Exchange (ETDEWEB)

    Berta, V.T.; Hanson, R.G.; Johnsen, G.W.; Schultz, R.R. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    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.

  2. Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Third semiannual report, January-June 1980

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbaum, H.S. (comp.)

    1980-09-01

    Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the work scope of this program one of these concepts is to be selected for demonstration in a commercial power reactor. It was decided to demonstrate Zr-liner in 132 bundles which have liners of either crystal-bar zirconium or of low-oxygen sponge zirconium in the reload for Quad Cities Unit 2, Cycle 6. Irradiation testing or barrier fuel was continued, and the superior PCI resistance of Zr-liner fuel was further substantiated in the current report period. Furthermore, an irradiation experiment in which Zr-liner fuel, having a deliberately fabricated cladding perforation, was operated at a linear heat generation rate of 35 kW/m to a burnup of approx. 3 MWd/kg U showed no unusual signs of degradation compared with a similarly defected reference fuel rod. Four lead test assemblies of barrier fuel (two of Zr-liner and two of Cu-barrier), presently under irradiation in Quad Cities Unit 1, have achieved a burnup of 11 MWd/kg U.

  3. Capabilities to improve corrosion resistance of fuel claddings by using powerful laser and plasma sources

    Science.gov (United States)

    Borisov, V. M.; Trofimov, V. N.; Sapozhkov, A. Yu.; Kuzmenko, V. A.; Mikhaylov, V. B.; Cherkovets, V. Ye.; Yakushkin, A. A.; Yakushin, V. L.; Dzhumayev, P. S.

    2016-12-01

    The treatment conditions of fuel claddings of the E110 alloy by using powerful UV or IR laser radiation, which lead to the increase in the corrosion resistance at the high-temperature ( T = 1100°C) oxidation simulating a loss-of-coolant accident, are determined. The possibility of the complete suppression of corrosion under these conditions by using pulsed laser deposition of a Cr layer is demonstrated. The behavior of protective coatings of Al, Al2O3, and Cr planted on steel EP823 by pulsed laser deposition, which is planned to be used in the BREST-OD-300, is studied. The methods of the almost complete suppression of corrosion in liquid lead to the temperature of 720°C are shown.

  4. Capabilities to improve corrosion resistance of fuel claddings by using powerful laser and plasma sources

    Energy Technology Data Exchange (ETDEWEB)

    Borisov, V. M., E-mail: borisov@triniti.ru; Trofimov, V. N.; Sapozhkov, A. Yu.; Kuzmenko, V. A.; Mikhaylov, V. B.; Cherkovets, V. Ye.; Yakushkin, A. A. [Troitsk Institute for Innovation and Fusion Research (Russian Federation); Yakushin, V. L.; Dzhumayev, P. S. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

    2016-12-15

    The treatment conditions of fuel claddings of the E110 alloy by using powerful UV or IR laser radiation, which lead to the increase in the corrosion resistance at the high-temperature (T = 1100°C) oxidation simulating a loss-of-coolant accident, are determined. The possibility of the complete suppression of corrosion under these conditions by using pulsed laser deposition of a Cr layer is demonstrated. The behavior of protective coatings of Al, Al{sub 2}O{sub 3}, and Cr planted on steel EP823 by pulsed laser deposition, which is planned to be used in the BREST-OD-300, is studied. The methods of the almost complete suppression of corrosion in liquid lead to the temperature of 720°C are shown.

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

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan [Univ. of Tennessee, Knoxville, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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)

  6. Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Korkut, Turgay; Kara, Ayhan; Korkut, Hatun [Sinop Univ. (Turkey). Dept. of Nuclear Energy Engineering

    2016-12-15

    Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.

  7. Effect of nitrogen on tensile properties and structures of T-111 (tantalum, 8 percent tungsten, 2 percent hafnium) tubing

    Science.gov (United States)

    Buzzard, R. J.; Metroka, R. R.

    1973-01-01

    The effect of controlled nitrogen additions was evaluated on the mechanical properties of T-111 (Ta-8W-2Hf) fuel pin cladding material proposed for use in a lithium-cooled nuclear reactor concept. Additions of 80 to 1125 ppm nitrogen resulted in increased strengthening of T-111 tubular section test specimens at temperatures of 25 to 1200 C. Homogeneous distributions of up to 500 ppm nitrogen did not seriously decrease tensile ductility. Both single and two-phase microstructures, with hafnium nitride as the second phase, were evaluated in this study.

  8. Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Park, Dong Jun, E-mail: pdj@kaeri.re.kr; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

    2016-12-15

    This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility. - Highlights: • Cr and FeCrAl were coated onto Zr fuel cladding for light water nuclear reactors. • Mo layer between FeCrAl and Zr prevented inter-diffusion at high temperatures. • Coated claddings were tested under loss-of-cooling accident conditions. • Coating improved high-temperature oxidation resistance and mechanical properties.

  9. Results of High-Temperature Heating Test for Irradiated U-10Zr(-5Ce with T92 Cladding Fuel

    Directory of Open Access Journals (Sweden)

    June-Hyung Kim

    2016-11-01

    Full Text Available A microstructure observation using an optical microscope, SEM and EPMA was performed for the irradiated U-10Zr and U-10Zr-5Ce fuel slugs with a T92 cladding specimen after a high-temperature heating test. Also, the measured eutectic penetration rate was compared with the value predicted by the existing eutectic penetration correlation being used for design and modeling purposes. The heating temperature and duration time for the U-10Zr/T92 specimen were 750 °C and 1 h, and those for the U-10Zr-5Ce/T92 specimen were 800 °C and 1 h. In the case of the U-10Zr/T92 specimen, the migration phenomena of U, Zr, Fe, and Cr as well as the Nd lanthanide fission product were observed at the eutectic melting region. The measured penetration rate was similar to the value predicted by the existing eutectic penetration rate correlation. In addition, when comparing with measured eutectic penetration rates for the unirradiated U-10Zr fuel slug with FMS (ferritic martensitic steel, HT9 or Gr.91 cladding specimens which had been reported in the literature, the measured eutectic penetration rate for the irradiated fuel specimen was higher than that for the unirradiated U-10Zr specimen. In the case of the U-10Zr-5Ce/T92 specimen in which there had been a gap between the fuel slug and cladding after the irradiation test, the eutectic melting region was not found because contact between the fuel slug and cladding did not take place during the heating test.

  10. Development of nuclear fuel for the future -Development of performance improvement of the cladding by ion beam-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Byung Hoh; Jung, Moon Kyoo; Jung, Kee Suk; Kim, Wan; Lee, Jae Hyung; Song, Tae Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Han, Jun Kun [Sung Kyoon Kwan Univ., Seoul (Korea, Republic of); Kwon, Hyuk Sang [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1995-07-01

    In this research we analyzed the state of art related to the surface treatment method of nuclear fuel cladding for the development of the surface treatment technique of nuclear fuel cladding by ion beam while investigating major causes of the leakage of fuel rods. Ion implantation simulation code called TRIM-95 was used to decide basic parameters of ion beams and setup an appropriate process for ion implantation. Performance of the ion beam extraction was measured after adding the needed vacuum and cooling system to the existing gas and metal ion implanters. Target system for the ion implantation of fuel cladding improved and a plasma accelerator was installed on the target chamber of the metal ion implanter. The plasma accelerator is used to produce low energy, high current ion beams. The mechanical and chemical properties of the implanted Zircaloy-4 such as micro hardness, wear resistance, fretting wear, friction coefficient and corrosion resistance was measured under the room temperature and atmosphere. A micro structure and composition analysis of Zircaloy-4 sample was performed before and after the implantation to study the cause of the improvement in the mechanical and chemical characteristics. 94 figs, 11 tabs, 51 refs. (Author).

  11. Development and Experimental Benchmark of Simulations to Predict Used Nuclear Fuel Cladding Temperatures during Drying and Transfer Operations

    Energy Technology Data Exchange (ETDEWEB)

    Greiner, Miles [Univ. of Nevada, Reno, NV (United States)

    2017-03-31

    Radial hydride formation in high-burnup used fuel cladding has the potential to radically reduce its ductility and suitability for long-term storage and eventual transport. To avoid this formation, the maximum post-reactor temperature must remain sufficiently low to limit the cladding hoop stress, and so that hydrogen from the existing circumferential hydrides will not dissolve and become available to re-precipitate into radial hydrides under the slow cooling conditions during drying, transfer and early dry-cask storage. The objective of this research is to develop and experimentallybenchmark computational fluid dynamics simulations of heat transfer in post-pool-storage drying operations, when high-burnup fuel cladding is likely to experience its highest temperature. These benchmarked tools can play a key role in evaluating dry cask storage systems for extended storage of high-burnup fuels and post-storage transportation, including fuel retrievability. The benchmarked tools will be used to aid the design of efficient drying processes, as well as estimate variations of surface temperatures as a means of inferring helium integrity inside the canister or cask. This work will be conducted effectively because the principal investigator has experience developing these types of simulations, and has constructed a test facility that can be used to benchmark them.

  12. Critical Analysis of Dry Storage Temperature Limits for Zircaloy-Clad Spent Nuclear Fuel Based on Diffusion Controlled Cavity Growth

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, T.A.; Rosen, R.S.; Kassner, M.E.

    1999-12-01

    Interim dry storage of spent nuclear fuel (SNF) rods is of critical concern because a shortage of existing SNF wet storage capacity combined with delays in the availability of a permanent disposal repository has led to an increasing number of SNF rods being placed into interim dry storage. Safe interim dry storage must be maintained for a minimum of twenty years according to the Standard Review Plan for Dry Cask Storage Systems [1] and the Code of Federal Regulations, 10 CFR Part 72 [2]. Interim dry storage licensees must meet certain safety conditions when storing SNF rods to ensure that there is a ''very low probability (e.g. 0.5%) of cladding breach during long-term storage'' [1]. Commercial SNF typically consists of uranium oxide pellets surrounded by a thin cladding. The cladding is usually an {alpha}-zirconium based alloy know as ''Zircaloy''. In dry storage, the SNF rods are confined in one of several types of cask systems approved by the Nuclear Regulatory Commission (NRC). ''The cask system must be designed to prevent degradation of fuel cladding that results in a type of cladding breach, such as axial-splits or ductile fracture, where irradiated UO{sub 2} particles may be released. In addition, the fuel cladding should not degrade to the point where more than one percent of the fuel rods suffer pinhole or hairline crack type failure under normal storage conditions [1].'' The NRC has approved two models [3,4] for use by proposed dry storage licensees to determine the maximum initial temperature limit for nuclear fuel rods in dry storage that supposedly meet the above criteria and yield consistent temperature limits. Though these two models are based on the same fundamental failure theory, different assumptions have been made including the choice of values for material constants in the failure equation. This report will examine and compare the similarities and inconsistencies of these two models

  13. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Heuser, Brent [Univ. of Illinois, Urbana-Champaign, IL (United States); Stubbins, James [Univ. of Illinois, Urbana-Champaign, IL (United States); Kozlowski, Tomasz [Univ. of Illinois, Urbana-Champaign, IL (United States); Uddin, Rizwan [Univ. of Illinois, Urbana-Champaign, IL (United States); Trinkle, Dallas [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, Thoms [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); ang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States); Sabharwall, piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-25

    The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

  14. High Temperature Steam Oxidation Testing of Candidate Accident Tolerant Fuel Cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nelson, Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parker, Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parkison, Adam [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2013-12-23

    The Fuel Cycle Research and Development (FCRD) program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels in order to overcome the inherent shortcomings of light water reactor (LWR) fuels when exposed to beyond design basis accident conditions. The campaign has invested in development of experimental infrastructure within the Department of Energy complex capable of chronicling the performance of a wide range of concepts under prototypic accident conditions. This report summarizes progress made at Oak Ridge National Laboratory (ORNL) and Los Alamos National Laboratory (LANL) in FY13 toward these goals. Alternative fuel cladding materials to Zircaloy for accident tolerance and a significantly extended safety margin requires oxidation resistance to steam or steam-H2 environments at ≥1200°C for short times. At ORNL, prior work focused attention on SiC, FeCr and FeCrAl as the most promising candidates for further development. Also, it was observed that elevated pressure and H2 additions had minor effects on alloy steam oxidation resistance, thus, 1 bar steam was adequate for screening potential candidates. Commercial Fe-20Cr-5Al alloys remain protective up to 1475°C in steam and CVD SiC up to 1700°C in steam. Alloy development has focused on Fe-Cr-Mn-Si-Y and Fe-Cr-Al-Y alloys with the aluminaforming alloys showing more promise. At 1200°C, ferritic binary Fe-Cr alloys required ≥25% Cr to be protective for this application. With minor alloy additions to Fe-Cr, more than 20%Cr was still required, which makes the alloy susceptible to α’ embrittlement. Based on current results, a Fe-15Cr-5Al-Y composition was selected for initial tube fabrication and welding for irradiation experiments in FY14. Evaluations of chemical vapor deposited (CVD) SiC were conducted up to 1700°C in steam. The reaction of H2O with the alumina reaction tube at 1700°C resulted in Al(OH)3

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

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    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

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    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

    Energy Technology Data Exchange (ETDEWEB)

    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. Characterization of Zircaloy-4 tubing procured for fuel cladding research programs

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, R.H. (comp.)

    1976-06-14

    A quantity of Zircaloy-4 tubing (10.92 mm outside diameter by 0.635 mm wall thickness) was purchased specifically for use in a number of related fuel cladding research programs sponsored by the Division of Reactor Safety Research, Nuclear Regulatory Commission (NRC/RSR). Identical tubing (produced simultaneously and from the same ingot) was purchased concurrently by the Electric Power Research Institute (EPRI) for use in similar research programs sponsored by that organization. In this way, source variability and prior fabrication history were eliminated as parameters, thus permitting direct comparison (as far as as-received material properties are concerned) of experimental results from the different programs. The tubing is representative of current reactor technology. Consecutive serial numbers assigned to each tube identify the sequence of the individual tubes through the final tube wall reduction operation. The report presented documents the procurement activities, provides a convenient reference source of manufacturer's data and tubing distribution to the various users, and presents some preliminary characterization data. The latter have been obtained routinely in various research programs and are not complete. Although the number of analyses, tests, and/or examinations performed to date are insufficient to draw statistically valid conclusions with regard to material characterization, the data are expected to be representative of the as-received tubing. It is anticipated that additional characterizations will be performed and reported routinely by the various research programs that use the tubing.

  19. Thermochemistry of metal-rich manganese telluride and its role in fuel-clad interactions

    Science.gov (United States)

    Baba, M. Sai; Narasimhan, T. S. Lakshmi; Balasubramanian, R.; Mathews, C. K.

    Vaporisation of Mn-Te alloys was studied by Knudsen-effusion mass spectrometry. The partial pressures of Te(g) over the two-phase field, Mn-MnTe, were determined in the temperature range 1120-1250 K. Two samples of initial composition of 29.1 and 40.1 at% Te were used in the experiments. The vapour phase consists of Mn(g) and Te(g). Partial pressure-temperature relations for Te(g) were found to follow the equation log( {(p)}/{Pa}) = - {(16099±285)}/{T(K)}+(10.525±0.240) . Mn-rich phase boundary of MnTe was determined from continuous vaporisation experiments starting with a sample from two-phase field Mn + MnTe. The boundary composition was found to be 44.3 ± 0.5 at% Te in the temperature range 1205-1280 K. Enthalpy of the following reactions was obtained: MnTe0.8( s) / aiMn( s) + 0.8 Te( g), MnTe0.8( s) / aiMn( g) + 0.8 Te( g) and Mn( s) / aiMn( g). The standard molar enthalpy and Gibbs energy of formation of MnTe 0.8 were arrived at. The tellurium potential which would be required for the formation of MnTe 0.8 in AISI 316 stainless steel was calculated and the possibility of such a telluride formation in the fuel-cladding gap of a mixed-oxide fuel pin is discussed.

  20. Thermochemistry of metal-rich chromium telluride and its role in fuel-clad chemical interactions

    Science.gov (United States)

    Viswanathan, R.; Sai Baba, M.; Albert Raj, D. Darwin; Balasubramanian, R.; Saha, B.; Mathews, C. K.

    1989-09-01

    Vaporisation of Cr-Te alloys was studied by Knudsen-effusion mass spectrometry. The partial pressures of Te 2(g) and Te(g) over the two-phase field, Cr + CrTex, were determined in the temperature ranges 1015 to 1138 K and 1180 to 1285 K, respectively. The temperature dependencies of the partial pressures have indicated that nearly equimolar proportions of Te and Te 2 are present in the vapor phase and that there is a phase transformation in CrTe x at 1160 ± 20 K. The Cr-rich phase boundaries of the nonstoichiometric CrTe x were delineated at 1075 K (50.72 ± 0.7 at% Te) as well as at 1235 K (48.25 ± 0.9 at% Te) by a continuous monitoring of the intensities of Te + and Te +2 as a function of time, starting with samples having 55.63 and 57.22 at% Te. Enthalpies and Gibbs energy changes were derived for the equilibria. CrTe x(s) ai Cr(s) + ( {x}/{i})Te i(g) [ x = 1.029 and 0.932; i = 1 and 2] and Te2( g) ai 2 Te( g). Enthalpies and Gibbs energies of formation of CrTe 1.0.29 and CrTe 0.932 were arrived at. The tellurium potentials which would be required for the formation of MTe x ( M = Fe, Cr, andNi) in Type 316 stainless steel and those likely to exist in the fuel-cladding gap of a mixed-oxide fuel pin were computed.

  1. Carbon 14 distribution in irradiated BWR fuel cladding and released carbon 14 after aqueous immersion of 6.5 years

    Energy Technology Data Exchange (ETDEWEB)

    Sakuragi, T. [Radioactive Waste Management Funding and Research Center, Tsukishima 1-15-7, Chuo City, Tokyo, 104-0052 (Japan); Yamashita, Y.; Akagi, M.; Takahashi, R. [TOSHIBA Corporation, Ukishima Cho 4-1, Kawasaki Ward, Kawasaki, 210-0862 (Japan)

    2016-07-01

    Spent fuel cladding which is highly activated and strongly contaminated is expected to be disposed of in an underground repository. A typical activation product in the activated metal waste is carbon 14 ({sup 14}C), which is mainly generated by the {sup 14}N(n,p){sup 14}C reaction and produces a significant exposure dose due to the large inventory, long half-life (5730 years), rapid release rate, and the speciation and consequent migration parameters. In the preliminary Japanese safety case, the release of radionuclides from the metal matrix is regarded as the corrosion-related congruent release, and the cladding oxide layer is regarded as a source of instant release fraction (IRF). In the present work, specific activity of {sup 14}C was measured using an irradiated BWR fuel cladding (Zircaloy-2, average rod burnup of 41.6 GWd/tU) which has an external oxide film having a thickness of 25.3 μm. The {sup 14}C specific activity of the base metal was 1.49*10{sup 4} Bq/g, which in the corresponding burnup is comparable to values in the existing literature, which were obtained from various irradiated claddings. Although the specific activity in oxide was 2.8 times the base metal activity due to the additive generation by the {sup 17}O(n,α){sup 14}C reaction, the {sup 14}C abundance in oxide was less than 10% of total inventory. A static leaching test using the cladding tube was carried out in an air-tight vessel filled with a deoxygenated dilute NaOH solution (pH of 12.5) at room temperature. After 6.5 years, {sup 14}C was found in each leachate fraction of gas phase and dissolved organics and inorganics, the total of which was less than 0.01% of the {sup 14}C inventory of the immersed cladding tube. A simple calculation based on the congruent release with Zircaloy corrosion has suggested that the 96.7% of released {sup 14}C was from the external oxide layer and 3.3% was from the base Zircaloy metal. However, both the {sup 14}C abundance and the low leaching rate

  2. Modelling the role of pellet crack motion in the (r-θ) plane upon pellet-clad interaction in advanced gas reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Haynes, T.A. [Centre for Nuclear Engineering & Department of Materials, Imperial College London, Exhibition Rd., London SW7 2AZ (United Kingdom); Ball, J.A. [EDF Energy, Barnett Way, Gloucester GL4 3RS (United Kingdom); Wenman, M.R., E-mail: m.wenman@imperial.ac.uk [Centre for Nuclear Engineering & Department of Materials, Imperial College London, Exhibition Rd., London SW7 2AZ (United Kingdom)

    2017-04-01

    Highlights: • Finite element modelling of pellet relocation in the (r-θ) plane of nuclear fuel. • ‘Soft’ and ‘hard’ PCI have been predicted in a cracked nuclear fuel pellet. • Stress concentration in the cladding ahead of radial pellet cracks is predicted. • The model is very sensitive to the coefficient of friction and power ramp duration. • The model is less sensitive to the number of cracks assumed. - Abstract: A finite element model of pellet fragment relocation in the r-θ plane of advanced gas-cooled reactor (AGR) fuel is presented under conditions of both ‘hard’ and ‘soft’ pellet-clad interaction. The model was able to predict the additional radial displacement of fuel fragments towards the cladding as well as the stress concentration on the inner surface resulting from the azimuthal motion of pellet fragments. The model was subjected to a severe ramp in power from both full power and after a period of reduced power operation; in the former, the maximum hoop stress in the cladding was found to be increased by a factor of 1.6 as a result of modelling the pellet fragment motion. The pellet-clad interaction was found to be relatively insensitive to the number of radial pellet crack. However, it was very sensitive to both the coefficient of friction used between the clad and pellet fragments and power ramp duration.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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 (Faraday’s 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

  4. Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Laboratory

    2016-09-01

    U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

  5. Initial report on stress-corrosion-cracking experiments using Zircaloy-4 spent fuel cladding C-rings

    Energy Technology Data Exchange (ETDEWEB)

    Smith, H.D.

    1988-09-01

    The Nevada Nuclear Waste Storage Investigations (NNWSI) Project is sponsoring C-ring stress corrosion cracking scoping experiments as a first step in evaluating the potential for stress corrosion cracking of spent fuel cladding in a potential tuff repository environment. The objective is to scope the approximate behavior so that more precise pressurized tube testing can be performed over an appropriate range of stress, without expanding the long-term effort needlessly. The experiment consists of stressing, by compression with a dead weight load, C-rings fabricated from spent fuel cladding exposed to an environment of Well J-13 water held at 90{degree}C. The results indicate that stress corrosion cracking occurs at the high stress levels employed in the experiments. The cladding C-rings, tested at 90% of the stress at which elastic behavior is obtained in these specimens, broke in 25 to 64 d when tested in water. This was about one third of the time required for control tests to break in air. This is apparently the first observation of stress corrosion under the test conditions of relatively low temperature, benign environment but very high stress. The 150 ksi test stress could be applied as a result of the particular specimen geometry. By comparison, the uniaxial tensile yield stress is about 100 to 120 ksi and the ultimate stress is about 150 ksi. When a general model that fits the high stress results is extrapolated to lower stress levels, it indicates that the C-rings in experiments now running at {approximately}80% of the yield strength should take 200 to 225 d to break. 21 refs., 24 figs., 5 tabs.

  6. SEPARATING HAFNIUM FROM ZIRCONIUM

    Science.gov (United States)

    Lister, B.A.J.; Duncan, J.F.

    1956-08-21

    A dilute aqueous solution of zirconyl chloride which is 1N to 2N in HCl is passed through a column of a cation exchange resin in acid form thereby absorbing both zirconium and associated hafnium impurity in the mesin. The cation exchange material with the absorbate is then eluted with aqueous sulfuric acid of a O.8N to 1.2N strength. The first portion of the eluate contains the zirconium substantially free of hafnium.

  7. Eddy current examination of the nuclear fuel elements with aluminum 1100-F cladding of IPR-R1 research reactor: An initial study

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Roger F. da; Silva Júnior, Silvério F. da; Frade, Rangel T. [Centro de Desenvolvimento da Tecnologia Nucelar (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Rodrigues, Juliano S., E-mail: rfs@cdtn.br, E-mail: silvasf@cdtn.br, E-mail: rtf@cdtn.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2017-07-01

    Tubes of aluminum 1100-F as well as tubes of AISI 304 stainless steel are used as cladding of the fuel elements of TRIGA IPR-R1 nuclear research reactor. Usually, these tubes are inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements whose cladding has failed, but it is not able to determine the place where the discontinuity is located. On the other hand, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In previous works, the application of eddy current testing to evaluate the AISI 304 cladding fuel elements of TRIGA IPR-R1 was studied. In this paper, it is proposed an initial study about the use of eddy current testing for detection and characterization of discontinuities in the aluminum 1100-F fuel elements cladding. The study includes the development of probes and the design and manufacture of reference standards. (author)

  8. Zircaloy cladding performance under spent fuel disposal conditions; Progress report, May 1--October 31, 1989

    Energy Technology Data Exchange (ETDEWEB)

    Pescatore, C.; Cowgill, M.G.; Sullivan, T.M.

    1990-04-01

    The Brookhaven National Laboratory (BNL) Waste Materials and Environment Modeling (WMEM) Program has been assigned the task of helping the DOE formulate and certify analytical tools needed to support and/or strengthen the Waste Package Licensing Strategy. One objective of the WMEM program is to perform qualitative and quantitative analyses of irradiated Zircaloy cladding. This progress report presents the early findings of an on-going literature evaluation and the results of the numerical implementation of two models of Zircaloy creep. The report only addresses cladding degradation modes within intact, dry waste containers. Additional degradation modes will be considered when the study is expanded to include moist environments and partly failed containers. Further updates of the present analyses will also be provided.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Hydride reorientation and its impact on ambient temperature mechanical properties of high burn-up irradiated and unirradiated recrystallized Zircaloy-2 nuclear fuel cladding with an inner liner

    Science.gov (United States)

    Auzoux, Q.; Bouffioux, P.; Machiels, A.; Yagnik, S.; Bourdiliau, B.; Mallet, C.; Mozzani, N.; Colas, K.

    2017-10-01

    Precipitation of radial hydrides in zirconium-based alloy cladding concomitant with the cooling of spent nuclear fuel during dry storage can potentially compromise cladding integrity during its subsequent handling and transportation. This paper investigates hydride reorientation and its impact on ductility in unirradiated and irradiated recrystallized Zircaloy-2 cladding with an inner liner (cladding for boiling water reactors) subjected to hydride reorientation treatments. Cooling from 400 °C, hydride reorientation occurs in recrystallized Zircaloy-2 with liner at a lower effective stress in irradiated samples (below 40 MPa) than in unirradiated specimens (between 40 and 80 MPa). Despite significant hydride reorientation, unirradiated recrystallized Zircaloy-2 with liner cladding containing ∼200 wppm hydrogen shows a high diametral strain at fracture (>15%) during burst tests at ambient temperature. This ductile behavior is due to (1) the lower yield stress of the recrystallized cladding materials in comparison to hydride fracture strength (corrected by the compression stress arising from the precipitation) and (2) the hydride or hydrogen-depleted zone as a result of segregation of hydrogen into the liner layer. In irradiated Zircaloy-2 with liner cladding containing ∼340 wppm hydrogen, the conservation of some ductility during ring tensile tests at ambient temperature after reorientation treatment at 400 °C with cooling rates of ∼60 °C/h is also attributed to the existence of a hydride-depleted zone. Treatments at lower cooling rates (∼6 °C/h and 0.6 °C/h) promote greater levels of hydrogen segregation into the liner and allow for increased irradiation defect annealing, both of which result in a significant increase in ductility. Based on this investigation, given the very low cooling rates typical of dry storage systems, it can be concluded that the thermal transients associated with dry storage should not degrade, and more likely should actually

  11. Simulation with DIONISIO 1.0 of thermal and mechanical pellet-cladding interaction in nuclear fuel rods

    Science.gov (United States)

    Soba, Alejandro; Denis, Alicia

    2008-02-01

    The code DIONISIO 1.0 describes most of the main phenomena occurring in a fuel rod throughout its life under normal operation conditions of a nuclear thermal reactor. Starting from the power history, DIONISIO predicts the temperature distribution in the domain, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the rod free volume, gas mixing, pressure increase, restructuring and grain growth in the UO 2 pellet, irradiation growth of the Zircaloy cladding, oxide layer growth on its surface, hydrogen uptake and the effects of a corrosive atmosphere either internal or external. In particular, the models of thermal conductance of the gap and of pellet-cladding mechanical interaction incorporated to the code constitute two realistic tools. The possibility of gap closure (including partial contact between rough surfaces) and reopening during burnup is allowed. The non-linear differential equations are integrated by the finite element method in two-dimensions assuming cylindrical symmetry. Good results are obtained for the simulation of several irradiation tests.

  12. Scanning electron microscopy examination of a Fast Flux Test Facility irradiated U-10Zr fuel cross section clad with HT-9

    Science.gov (United States)

    Harp, Jason M.; Porter, Douglas L.; Miller, Brandon D.; Trowbridge, Tammy L.; Carmack, William J.

    2017-10-01

    Observations from a scanning electron microscopy examination of irradiated U-10Zr fuel are presented. The sample studied had a local burnup of 5.7 atom percent and a local inner cladding temperature of 615 °C. This examination by electron microscopy has concentrated on producing data relevant to facilitating a better understanding of Zr redistribution in irradiated U-10Zr fuel and on a better understanding of the complex microstructure present in fuel cladding chemical interaction (FCCI) layers. The presented zirconium redistribution data supplements the existing literature by providing a data set at these particular local conditions. In addition to FCCI layers that are readily visible in optical microscopy, this examination has revealed lanthanide degradation of the cladding by what appears to be a grain boundary facilitated pathway. Precipitates of fission produced Pd-lanthanide compounds were observed in the fuel. Precipitated regions with elevated Mo and elevated W content were also observed in the HT-9 cladding of this sample.

  13. Experiments for evaluation of corrosion to develop storage criteria for interim dry storage of aluminum-alloy clad spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, H.B.; Sindelar, R.L.; Lam, P.S.; Murphy, T.H.

    1994-11-01

    The technical bases for specification of limits to environmental exposure conditions to avoid excessive degradation are being developed for storage criteria for dry storage of highly-enriched, aluminum-clad spent nuclear fuels owned by the US Department of Energy. Corrosion of the aluminum cladding is a limiting degradation mechanism (occurs at lowest temperature) for aluminum exposed to an environment containing water vapor. Attendant radiation fields of the fuels can lead to production of nitric acid in the presence of air and water vapor and would exacerbate the corrosion of aluminum by lowering the pH of the water solution. Laboratory-scale specimens are being exposed to various conditions inside an autoclave facility to measure the corrosion of the fuel matrix and cladding materials through weight change measurements and metallurgical analysis. In addition, electrochemical corrosion tests are being performed to supplement the autoclave testing by measuring differences in the general corrosion and pitting corrosion behavior of the aluminum cladding alloys and the aluminum-uranium fuel materials in water solutions.

  14. Three-dimensional fuel pin model validation by prediction of hydrogen distribution in cladding and comparison with experiment

    Energy Technology Data Exchange (ETDEWEB)

    Aly, A. [North Carolina State Univ., Raleigh, NC (United States); Avramova, Maria [North Carolina State Univ., Raleigh, NC (United States); Ivanov, Kostadin [Pennsylvania State Univ., University Park, PA (United States); Motta, Arthur [Pennsylvania State Univ., University Park, PA (United States); Lacroix, E. [Pennsylvania State Univ., University Park, PA (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Walter, D. [Univ. of Michigan, Ann Arbor, MI (United States); Williamson, R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gamble, K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-10-29

    To correctly describe and predict this hydrogen distribution there is a need for multi-physics coupling to provide accurate three-dimensional azimuthal, radial, and axial temperature distributions in the cladding. Coupled high-fidelity reactor-physics codes with a sub-channel code as well as with a computational fluid dynamics (CFD) tool have been used to calculate detailed temperature distributions. These high-fidelity coupled neutronics/thermal-hydraulics code systems are coupled further with the fuel-performance BISON code with a kernel (module) for hydrogen. Both hydrogen migration and precipitation/dissolution are included in the model. Results from this multi-physics analysis is validated utilizing calculations of hydrogen distribution using models informed by data from hydrogen experiments and PIE data.

  15. Structural cladding /clad structures

    DEFF Research Database (Denmark)

    Beim, Anne

    2012-01-01

    Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure...... of materials, the structural features and the construction details of building systems in selected architectural works. With a particular focus at heavy constructions made of solid wood and masonry, and light weight constructions made of wooden frame structures and steel profiles, it is the intention...... tightness in constructions. At the same time a need for longevity and effortless maintenance have lead to contemporary architectural structures, where the exterior walls and the building envelope most often are made of several layers of advanced materials and separate building elements. In most contemporary...

  16. Nondestructive Evaluation on Hydrided LWR Fuel Cladding by Small Angle Incoherent Neutron Scattering of Hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Yong [ORNL; Qian, Shuo [ORNL; Littrell, Ken [ORNL; Parish, Chad M [ORNL; Bell, Gary L [ORNL; Plummer, Lee K [ORNL

    2013-01-01

    A non-destructive neutron scattering method was developed to precisely measure the uptake of total hydrogen in nuclear grade Ziraloy-4 cladding. The hydriding apparatus consists of a closed stainless steel vessel that contains Zr alloy specimens and H gas. By controlling the initial H gas pressure in the vessel and the temperature profile, target H concentrations from tens of ppm to a few thousands of wppm have been successfully achieved. Following H charging, the H content of the hydrided specimens was measured using the vacuum hot extraction method (VHE), by which the samples with desired H concentration were selected for the neutron study. Small angle neutron incoherent scattering (SANIS) were performed in the High Flux Isotope Reactor at Oak Ridge national Laboratory (ORNL). Our study indicates that a very small amount ( 20 ppm) H in commercial Zr cladding can be measured very accurately in minutes for a wide range of H concentration by a nondestructive method. The H distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor, which is determined by calibration process with direct chemical analysis method on the specimen. This scale factor can be used for future test with unknown H concentration, thus provide a nondestructive method for absolute H concentration determination.

  17. Improving 6061-Al Grain Growth and Penetration across HIP-Bonded Clad Interfaces in Monolithic Fuel Plates: Initial Studies

    Energy Technology Data Exchange (ETDEWEB)

    Hackenberg, Robert E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McCabe, Rodney J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Montalvo, Joel D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Clarke, Kester D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dvornak, Matthew J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Edwards, Randall L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Crapps, Justin M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trujillo, R. Ralph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aikin, Beverly [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vargas, Victor D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hollis, Kendall J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lienert, Thomas J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Forsyth, Robert T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Harada, Kiichi L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2013-05-06

    Grain penetration across aluminum-aluminum cladding interfaces in research reactor fuel plates is desirable and was obtained by a legacy roll-bonding process, which attained 20-80% grain penetration. Significant grain penetration in monolithic fuel plates produced by Hot Isostatic Press (HIP) fabrication processing is equally desirable but has yet to be attained. The goal of this study was to modify the 6061-Al in such a way as to promote a much greater extent of crossinterface grain penetration in monolithic fuel plates fabricated by the HIP process. This study documents the outcomes of several strategies attempted to attain this goal. The grain response was characterized using light optical microscopy (LOM) electron backscatter diffraction (EBSD) as a function of these prospective process modifications done to the aluminum prior to the HIP cycle. The strategies included (1) adding macroscopic gaps in the sandwiches to enhance Al flow, (2) adding engineering asperities to enhance Al flow, (3) adding stored energy (cold work), and (4) alternative cleaning and coating. Additionally, two aqueous cleaning methods were compared as baseline control conditions. The results of the preliminary scoping studies in all the categories are presented. In general, none of these approaches were able to obtain >10% grain penetration. Recommended future work includes further development of macroscopic grooving, transferred-arc cleaning, and combinations of these with one another and with other processes.

  18. Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

  19. Past research and fabrication conducted at SCK•CEN on ferritic ODS alloys used as cladding for FBR's fuel pins

    Science.gov (United States)

    De Bremaecker, Anne

    2012-09-01

    In the 1960s in the frame of the sodium-cooled fast breeders, SCK•CEN decided to develop claddings made with ferritic stainless materials because of their specific properties, namely a higher thermal conductivity, a lower thermal expansion, a lower tendency to He-embrittlement, and a lower swelling than the austenitic stainless steels. To enhance their lower creep resistance at 650-700 °C arose the idea to strengthen the microstructure by oxide dispersions. This was the starting point of an ambitious programme where both the matrix and the dispersions were optimized. A purely ferritic 13 wt% Cr matrix was selected and its mechanical strength was improved through addition of ferritizing elements. Results of tensile and stress-rupture tests showed that Ti and Mo were the most beneficial elements, partly because of the chi-phase precipitation. In 1973 the optimized matrix composition was Fe-13Cr-3.5Ti-2Mo. To reach creep properties similar to those of AISI 316, different dispersions and methods were tested: internal oxidation (that was not conclusive), and the direct mixing of metallic and oxide powders (Al2O3, MgO, ZrO2, TiO2, ZrSiO4) followed by pressing, sintering, and extrusion. The compression and extrusion parameters were determined: extrusion as hollow at 1050 °C, solution annealing at 1050 °C/15 min, cleaning, cold drawing to the final dimensions with intermediate annealings at 1050 °C, final annealing at 1050 °C, straightening and final aging at 800 °C. The choice of titania and yttria powders and their concentrations were finalized on the basis of their out-of-pile and in-pile creep and tensile strength. As soon as a resistance butt welding machine was developed and installed in a glove-box, fuel segments with PuO2 were loaded in the Belgian MTR BR2. The fabrication parameters were continuously optimized: milling and beating, lubrication, cold drawing (partial and final reduction rates, temperature, duration, atmosphere and furnace). Specific non

  20. Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

    Energy Technology Data Exchange (ETDEWEB)

    Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  1. Obtention and physical and microstructural characterization of monolhitic U-2.5Zr-7.5Zr alloy fuel plate cladded in zircaloy

    Energy Technology Data Exchange (ETDEWEB)

    Machado, Geraldo Correa; Torres, Felipe Silva; Santos, Ana Maria Matildes dos; Paula, Joao Bosco de; Brina, Jose Giovanni Mascarenhas; Lameiras, Fernando Soares; Ferraz, Wilmar Barbosa, E-mail: ferrazw@cdtn.br [Centro de Desenvolvimento da Tecnologia Nucelar (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    A major advantage of monolithic metallic nuclear fuel is due to the higher density of uranium achieved when compared with the usual dispersion fuel. The benefit of utilizing a fuel with a higher density of uranium is that it provides a higher neutron flux which, in turn, allows the replacement of high enrichment uranium by low enriched one in order to obtain the maximum fuel enrichment of 20% U-235, as established under international agreements. Some transition metals such as Zr, Mo, Nb have been used in developing uranium alloys for apply in nuclear reactors in order to retain the gamma phase of high structural stability. The ternary alloy elements uranium, zirconium and niobium have the advantage of combining a greater mechanical resistance due to zirconium with high corrosion resistance afforded by niobium, allowing of a high performance fuel. Added to this, the use of zircaloy cladding, replacing the aluminum alloys, allows the development of a higher burn-up fuel throughout reactor operation. This may be explained due to increased chemical stability of the zircaloy when in contact with uranium alloys, instead of cladding with aluminum alloys. The fuel plate cladded with aluminum alloys favors the formation of deleterious chemical reactions products between aluminum and uranium alloys with serious consequences to the fuel performance. In this work, the development of monolithic fuel plate is done using a technique called 'picture-frame'. In this way, it is obtained a sandwich comprising of a U-2.5Zr-7.5NB monolithic alloy, instead of the usual dispersion of the alloy in a metal matrix, coupled to a frame and cladded by zircaloy plates. The sandwich thus obtained was then hot and cold rolled. Samples cut from the fuel plate were prepared by the usual metallography techniques, and then analyzed by physical and microstructural techniques of X-ray diffraction, optical and electron microscopy, microanalysis of energy dispersive (EDS), Vickers microhardness

  2. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  3. Investigations of Aluminum-Doped Self-Healing Zircaloy Surfaces in Context of Accident-Tolerant Fuel Cladding Research

    Science.gov (United States)

    Carr, James; Vasudevamurthy, Gokul; Snead, Lance; Hinderliter, Brian; Massey, Caleb

    2016-06-01

    We present here some important results investigating aluminum as an effective surface dopant for increased oxidation resistance of zircaloy nuclear fuel cladding. At first, the transport behavior of aluminum into reactor grade zircaloy was studied using simple diffusion couples at temperatures greater than 770 K. The experiments revealed the formation of tens of microns thick graded Zr-Al layers. The activation energy of aluminum in zircaloy was found to be ~175 kJ/mol (~1.8 eV), indicating the high mobility of aluminum in zircaloy. Subsequently, aluminum sputter-coated zircaloy coupons were heat-treated to achieve surface doping and form compositionally graded layers. These coupons were then tested in steam environments at 1073 and 1273 K. The microstructure of the as-fabricated and steam-corroded specimens was compared to those of pure zircaloy control specimens. Analysis of data revealed that aluminum effectively competed with zircaloy for oxygen up until 1073 K blocking oxygen penetration, with no traces of large scale spalling, indicating mechanically stable interfaces and surfaces. At the highest steam test temperatures, aluminum was observed to segregate from the Zr-Al alloy under layers and migrate to the surface forming discrete clusters. Although this is perceived as an extremely desirable phenomenon, in the current experiments, oxygen was observed to penetrate into the zirconium-rich under layers, which could be attributed to formation of surface defects such as cracks in the surface alumina layers.

  4. Study of corrosion and deposit on fuel cladding under heat flux. Adjustment of an original experimental device

    Energy Technology Data Exchange (ETDEWEB)

    Foucault, M.; Albinet, B.; Lucotte, J.P.; Thomazet, J. [Framatome ANP (France)

    2002-07-01

    I INTRODUCTION The cycle extension imposes an increase of the boric acid concentration of the primary environment at the beginning of the cycles. This evolution modifies the chemical and electrochemical conditions in the primary system. The main effect is to reduce the pH of the environment. The solubility of metal cations is dependent on the temperature and the pH of the solution. The solubility limit of iron is decreasing with the temperature in the range 300 - 360 C if the pH is lowered than a critical value which is near 5.8 [1]. The figure 1 gives for instance the evolution for iron solubility limit with temperature in this pH condition. Thus the metal release of the primary system can be the source of deposit on fuel cladding during cycles. In parallel, in the modern PWR's, the surface power density is increased, leading to higher surface temperature with risk of boiling. (authors)

  5. Chemical compatibility between UO2 fuel and SiC cladding for LWRs. Application to ATF (Accident-Tolerant Fuels)

    Science.gov (United States)

    Braun, James; Guéneau, Christine; Alpettaz, Thierry; Sauder, Cédric; Brackx, Emmanuelle; Domenger, Renaud; Gossé, Stéphane; Balbaud-Célérier, Fanny

    2017-04-01

    Silicon carbide-silicon carbide (SiC/SiC) composites are considered to replace the current zirconium-based cladding materials thanks to their good behavior under irradiation and their resistance under oxidative environments at high temperature. In the present work, a thermodynamic analysis of the UO2±x/SiC system is performed. Moreover, using two different experimental methods, the chemical compatibility of SiC towards uranium dioxide, with various oxygen contents (UO2±x) is investigated in the 1500-1970 K temperature range. The reaction leads to the formation of mainly uranium silicides and carbides phases along with CO and SiO gas release. Knudsen Cell Mass Spectrometry is used to measure the gas release occurring during the reaction between UO2+x and SiC powders as function of time and temperature. These experimental conditions are representative of an open system. Diffusion couple experiments with pellets are also performed to study the reaction kinetics in closed system conditions. In both cases, a limited chemical reaction is observed below 1700 K, whereas the reaction is enhanced at higher temperature due to the decomposition of SiC leading to Si vaporization. The temperature of formation of the liquid phase is found to lie between 1850 < T < 1950 K.

  6. Preliminary developments of miniplate-type fuel of U-2.5Zr-7.5Nb alloy dispersed and cladded in zircaloy

    Energy Technology Data Exchange (ETDEWEB)

    Cantagalli, Natalia Mattar; Faeda, Kelly Cristina Martins; Braga, Daniel Martins; Paula, Joao Bosco de; Ferraz, Wilmar Barbosa, E-mail: natalia.cantagalli@prof.una.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    Nuclear fuel based on uranium metal alloys is utilized in research and test reactors. For the purpose of the reduction of fuel enrichment, high densities of uranium-235 in this kind of fuel are needed. This can be achieved when uranium alloys are used containing elements such as Zr, Mo and Nb. The construction of fuel element with high-uranium density requires materials with low cross sections for neutron absorption, stability under irradiation and thermal cycling, as also absence of the chemical interactions between the fuel and cladding elements. In case of U-Zr-Nb alloys, zircaloy (Zry) cladding is a better option due to the fact that they have a higher chemical compatibility with zirconium alloys when compared with the use of aluminum alloys. This study aims to develop plate type nuclear fuel using the U-2.5Zr-7.5Nb alloy dispersed in Zry. Powders of the U-2.5Zr-7.5Nb alloy were obtained by hydriding-dehydriding process. Uranium alloy and Zry powders were homogenized, compacted in pellets, placed between two Zry plates as a sandwich. This assembly was hot rolled forming the dispersion fuel plate which was characterized by density and microhardness measurements, phases evaluation, grain sizes, pores and precipitates presences. It was observed by visual inspection that the fuel plate showed no failures and a perfect metallurgical bonding. Results obtained by energy dispersive spectrometry (EDS) show that there is no interaction between the U-2.5Zr-7.5Nb alloy and Zry matrix revealing a fuel with properties of high stability. (author)

  7. Mineral resource of the month: zirconium and hafnium

    Science.gov (United States)

    Gambogi, Joseph

    2007-01-01

    Zirconium and hafnium are corrosion-resistant metals that are grouped in the same family as titanium on the periodic table. The two elements commonly occur in oxide and silicate minerals and have significant economic importance in everything from ink, ceramics and golf shoes to nuclear fuel rods.

  8. Zirconium and hafnium

    Science.gov (United States)

    Jones, James V.; Piatak, Nadine M.; Bedinger, George M.; Schulz, Klaus J.; DeYoung, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Zirconium and hafnium are corrosion-resistant metals that are widely used in the chemical and nuclear industries. Most zirconium is consumed in the form of the main ore mineral zircon (ZrSiO4, or as zirconium oxide or other zirconium chemicals. Zirconium and hafnium are both refractory lithophile elements that have nearly identical charge, ionic radii, and ionic potentials. As a result, their geochemical behavior is generally similar. Both elements are classified as incompatible because they have physical and crystallochemical properties that exclude them from the crystal lattices of most rock-forming minerals. Zircon and another, less common, ore mineral, baddeleyite (ZrO2), form primarily as accessory minerals in igneous rocks. The presence and abundance of these ore minerals in igneous rocks are largely controlled by the element concentrations in the magma source and by the processes of melt generation and evolution. The world’s largest primary deposits of zirconium and hafnium are associated with alkaline igneous rocks, and, in one locality on the Kola Peninsula of Murmanskaya Oblast, Russia, baddeleyite is recovered as a byproduct of apatite and magnetite mining. Otherwise, there are few primary igneous deposits of zirconium- and hafnium-bearing minerals with economic value at present. The main ore deposits worldwide are heavy-mineral sands produced by the weathering and erosion of preexisting rocks and the concentration of zircon and other economically important heavy minerals, such as ilmenite and rutile (for titanium), chromite (for chromium), and monazite (for rare-earth elements) in sedimentary systems, particularly in coastal environments. In coastal deposits, heavy-mineral enrichment occurs where sediment is repeatedly reworked by wind, waves, currents, and tidal processes. The resulting heavy-mineral-sand deposits, called placers or paleoplacers, preferentially form at relatively low latitudes on passive continental margins and supply 100 percent of

  9. Studies of electrochemical oxidation of Zircaloy nuclear reactor fuel cladding using time-of-flight-energy elastic recoil detection analysis

    Science.gov (United States)

    Whitlow, H. J.; Zhang, Y.; Wang, Y.; Winzell, T.; Simic, N.; Ahlberg, E.; Limbäck, M.; Wikmark, G.

    2000-03-01

    The trend towards increased fuel burn-up and higher operating temperatures in order to achieve more economic operation of nuclear power plants places demands on a better understanding of oxidative corrosion of Zircaloy (Zry) fuel rod cladding. As part of a programme to study these processes we have applied time-of-flight-energy elastic recoil detection (ToF-E ERD), electrochemical impedance measurements and scanning electron microscopy to quantitatively characterise thin-oxide films corresponding to the pre-transition oxidation regime. Oxide films of different nominal thickness in the 9-300 nm range were grown on a series of rolled Zr and Zry-2 plates by anodisation in dilute H 2SO 4 with applied voltages. The dielectric thickness of the oxide layer was determined from the electrochemical impedance measurements and the surface topography characterised by scanning electron microscopy. ToF-E ERD with a 60 MeV 127I 11+ ion beam was used to determine the oxygen content and chemical composition of the oxide layer. In the Zr samples, the oxygen content (O atom cm -2) that was determined by ERD was closely similar to the O content derived from impedance measurements from the dielectric film. The absolute agreement was well within the uncertainty associated with the stopping powers. Moreover, the measured composition of the thick oxide layers corresponded to ZrO 2 for the films thicker than 65 nm where the oxide layer was resolved in the ERD depth profile. Zry-2 samples exhibited a similar behaviour for small thickness ( ⩽130 nm) but had an enhanced O content at larger thicknesses that could be associated either with enhanced rough surface topography or porous oxide formation that was correlated with the presence of Second Phase Particles (SPP) in Zry-2. The concentration of SPP elements (Fe, Cr, Ni) in relation to Zr was the same in the outer 9×10 17 atom cm -2 of oxide as in the same thickness of metal. The results also revealed the presence of about 1 at.% 32S in the

  10. Effects of heat transfer coefficient treatments on thermal shock fracture prediction for LWR fuel claddings in water quenching

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho; Lee, Jeong Ik; Cheon, Hee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Accurate modeling of thermal shock induced stresses has become ever most important to emerging accident-tolerant ceramic cladding concepts, such as silicon carbide (SiC) and SiC coated zircaloy. Since fractures of ceramic (entirely ceramic or coated) occur by excessive tensile stresses with linear elasticity, modeling transient stress distribution in the material provides a direct indication of the structural integrity. Indeed, even for the current zircaloy cladding material, the oxide layer formed on the surface - where cracks starts to develop upon water quenching - essentially behaves as a brittle ceramic. Hence, enhanced understanding of thermal shock fracture of a brittle material would fundamentally contribute to safety of nuclear reactors for both the current fuel design and that of the coming future. Understanding thermal shock fracture of a brittle material requires heat transfer rate between the solid and the fluid for transient temperature fields of the solid, and structural response of the solid under the obtained transient temperature fields. In water quenching, a solid experiences dynamic time-varying heat transfer rates with phase changes of the fluid over a short quenching period. Yet, such a dynamic change of heat transfer rates during the water quenching transience has been overlooked in assessments of mechanisms, predictability, and uncertainties for thermal shock fracture. Rather, a time-constant heat transfer coefficient, named 'effective heat transfer coefficient' has become a conventional input to thermal shock fracture analysis. No single constant heat transfer could suffice to depict the actual stress evolution subject to dynamic heat transfer coefficient changes with fluid phase changes. Use of the surface temperature dependent heat transfer coefficient will remarkably increase predictability of thermal shock fracture of brittle materials and complete the picture of stress evolution in the quenched solid. The presented result

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

    Energy Technology Data Exchange (ETDEWEB)

    Tome, Carlos N [Los Alamos National Laboratory; Caro, J A [Los Alamos National Laboratory; Lebensohn, R A [Los Alamos National Laboratory; Unal, Cetin [Los Alamos National Laboratory; Arsenlis, A [LLNL; Marian, J [LLNL; Pasamehmetoglu, K [INL

    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.

  12. SEPARATION OF HAFNIUM FROM ZIRCONIUM

    Science.gov (United States)

    Overholser, L.B.; Barton, C.J. Sr.; Ramsey, J.W.

    1960-05-31

    The separation of hafnium impurities from zirconium can be accomplished by means of organic solvent extraction. The hafnium-containing zirconium feed material is dissolved in an aqueous chloride solution and the resulting solution is contacted with an organic hexone phase, with at least one of the phases containing thiocyanate. The hafnium is extracted into the organic phase while zirconium remains in the aqueous phase. Further recovery of zirconium is effected by stripping the onganic phase with a hydrochloric acid solution and commingling the resulting strip solution with the aqueous feed solution. Hexone is recovered and recycled by means of scrubbing the onganic phase with a sulfuric acid solution to remove the hafnium, and thiocyanate is recovered and recycled by means of neutralizing the effluent streams to obtain ammonium thiocyanate.

  13. Hafnium germanium telluride

    Science.gov (United States)

    Jang, Gyung-Joo; Yun, Hoseop

    2008-01-01

    The title hafnium germanium telluride, HfGeTe4, has been synthesized by the use of a halide flux and structurally characterized by X-ray diffraction. HfGeTe4 is isostructural with stoichiometric ZrGeTe4 and the Hf site in this compound is also fully occupied. The crystal structure of HfGeTe4 adopts a two-dimensional layered structure, each layer being composed of two unique one-dimensional chains of face-sharing Hf-centered bicapped trigonal prisms and corner-sharing Ge-centered tetra­hedra. These layers stack on top of each other to complete the three-dimensional structure with undulating van der Waals gaps. PMID:21202163

  14. Hafnium germanium telluride

    Directory of Open Access Journals (Sweden)

    Hoseop Yun

    2008-05-01

    Full Text Available The title hafnium germanium telluride, HfGeTe4, has been synthesized by the use of a halide flux and structurally characterized by X-ray diffraction. HfGeTe4 is isostructural with stoichiometric ZrGeTe4 and the Hf site in this compound is also fully occupied. The crystal structure of HfGeTe4 adopts a two-dimensional layered structure, each layer being composed of two unique one-dimensional chains of face-sharing Hf-centered bicapped trigonal prisms and corner-sharing Ge-centered tetrahedra. These layers stack on top of each other to complete the three-dimensional structure with undulating van der Waals gaps.

  15. On the effect of temperature on the threshold stress intensity factor of delayed hydride cracking in light water reactor fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Holston, Anna-MariaAlvarez; Stjarnsater, Johan [Studsvik Nuclear AB, Nykoping (Sweden)

    2017-06-15

    Delayed hydride cracking (DHC) was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR) fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below 300°C. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor (K{sub IH}) to initiate DHC as a function of temperature in Zry-4 for temperatures between 227°C and 315°C. The experimental technique used in this study was the pin-loading testing technique. To determine the K{sub IH}, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around 300°C, there was a sharp increase in K{sub IH} indicating the upper temperature limit for DHC. The value for K{sub IH} at 227°C was determined to be 2.6 ± 0.3 MPa √m.

  16. Ablation Resistant Zirconium and Hafnium Ceramics

    Science.gov (United States)

    Bull, Jeffrey (Inventor); White, Michael J. (Inventor); Kaufman, Larry (Inventor)

    1998-01-01

    High temperature ablation resistant ceramic composites have been made. These ceramics are composites of zirconium diboride and zirconium carbide with silicon carbide, hafnium diboride and hafnium carbide with silicon carbide and ceramic composites which contain mixed diborides and/or carbides of zirconium and hafnium. along with silicon carbide.

  17. Hot Isostatic Press Can Optimization for Aluminum Cladding of U-10Mo Reactor Fuel Plates: FY12 Final Report and FY13 Update

    Energy Technology Data Exchange (ETDEWEB)

    Clarke, Kester D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Crapps, Justin M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Scott, Jeffrey E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aikin, Beverly [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vargas, Victor D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dvornak, Matthew J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Duffield, Andrew N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Weinberg, Richard Y. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alexander, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Montalvo, Joel D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hudson, Richard W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mihaila, Bogdan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Liu, Cheng [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lovato, Manuel L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2013-08-26

    Currently, the proposed processing path for low enriched uranium – 10 wt. pct. molybdenum alloy (LEU-10Mo) monolithic fuel plates for high power research and test reactors includes hot isostatic pressing (HIP) to bond the aluminum cladding that encapsulates the fuel foil. Initial HIP experiments were performed at Idaho National Laboratory (INL) on approximately ¼ scale “mini” fuel plate samples using a HIP can design intended for these smaller experimental trials. These experiments showed that, with the addition of a co-rolled zirconium diffusion barrier on the LEU-10Mo alloy fuel foil, the HIP bonding process is a viable method for producing monolithic fuel plates. Further experimental trials at Los Alamos National Laboratory (LANL) effectively scaled-up the “mini” can design to produce full-size fuel prototypic plates. This report summarizes current efforts at LANL to produce a HIP can design that is further optimized for higher volume production runs. The production-optimized HIP can design goals were determined by LANL and Babcock & Wilcox (B&W) to include maintaining or improving the quality of the fuel plates produced with the baseline scaled-up mini can design, while minimizing material usage, improving dimensional stability, easing assembly and disassembly, eliminating machining, and significantly reducing welding. The initial small-scale experiments described in this report show that a formed-can approach can achieve the goals described above. Future work includes scaling the formed-can approach to full-size fuel plates, and current progress toward this goal is also summarized here.

  18. Amphoteric Aqueous Hafnium Cluster Chemistry.

    Science.gov (United States)

    Goberna-Ferrón, Sara; Park, Deok-Hie; Amador, Jenn M; Keszler, Douglas A; Nyman, May

    2016-05-17

    Selective dissolution of hafnium-peroxo-sulfate films in aqueous tetramethylammonium hydroxide enables extreme UV lithographic patterning of sub-10 nm HfO2 structures. Hafnium speciation under these basic conditions (pH>10), however, is unknown, as studies of hafnium aqueous chemistry have been limited to acid. Here, we report synthesis, crystal growth, and structural characterization of the first polynuclear hydroxo hafnium cluster isolated from base, [TMA]6 [Hf6 (μ-O2 )6 (μ-OH)6 (OH)12 ]⋅38 H2 O. The solution behavior of the cluster, including supramolecular assembly via hydrogen bonding is detailed via small-angle X-ray scattering (SAXS) and electrospray ionization mass spectrometry (ESI-MS). The study opens a new chapter in the aqueous chemistry of hafnium, exemplifying the concept of amphoteric clusters and informing a critical process in single-digit-nm lithography. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  19. Zirconium and hafnium in meteorites

    Science.gov (United States)

    Ehmann, W. D.; Chyi, L. L.

    1974-01-01

    The abundances of zirconium and hafnium have been determined in nine stony meteorites by a new, precise neutron-activation technique. The Zr/Hf abundance ratios for the chondrites vary in a rather narrow range, consistent with previously published observations from our group. Replicate analyses of new, carefully selected clean interior samples of the Cl chondrite Orgueil yield mean zirconium and hafnium abundances of 5.2 and 0.10 ppm, respectively. These abundances are lower than we reported earlier for two Cl chondrite samples which we now suspect may have suffered contamination. The new Cl zirconium and hafnium abundances are in closer agreement with predictions based on theories of nucleosynthesis than the earlier data.

  20. Long period gratings coated with hafnium oxide by plasma-enhanced atomic layer deposition for refractive index measurements.

    Science.gov (United States)

    Melo, Luis; Burton, Geoff; Kubik, Philip; Wild, Peter

    2016-04-04

    Long period gratings (LPGs) are coated with hafnium oxide using plasma-enhanced atomic layer deposition (PEALD) to increase the sensitivity of these devices to the refractive index of the surrounding medium. PEALD allows deposition at low temperatures which reduces thermal degradation of UV-written LPGs. Depositions targeting three different coating thicknesses are investigated: 30 nm, 50 nm and 70 nm. Coating thickness measurements taken by scanning electron microscopy of the optical fibers confirm deposition of uniform coatings. The performance of the coated LPGs shows that deposition of hafnium oxide on LPGs induces two-step transition behavior of the cladding modes.

  1. Modeling of the PWR fuel mechanical behaviour and particularly study of the pellet-cladding interaction in a fuel rod; Contribution a la modelisation du comportement mecanique des combustibles REP sous irradiation, avec en particulier le traitement de l`interaction pastille-gaine dans un crayon combustible

    Energy Technology Data Exchange (ETDEWEB)

    Hourdequin, N.

    1995-05-01

    In Pressurized Water Reactor (PWR) power plants, fuel cladding constitutes the first containment barrier against radioactive contamination. Computer codes, developed with the help of a large experimental knowledge, try to predict cladding failures which must be limited in order to maintain a maximal safety level. Until now, fuel rod design calculus with unidimensional codes were adequate to prevent cladding failures in standard PWR`s operating conditions. But now, the need of nuclear power plant availability increases. That leads to more constraining operating condition in which cladding failures are strongly influenced by the fuel rod mechanical behaviour, mainly at high power level. Then, the pellet-cladding interaction (PCI) becomes important, and is characterized by local effects which description expects a multidimensional modelization. This is the aim of the TOUTATIS 2D-3D code, that this thesis contributes to develop. This code allows to predict non-axisymmetric behaviour too, as rod buckling which has been observed in some irradiation experiments and identified with the help of TOUTATIS. By another way, PCI is influenced by under irradiation experiments and identified with the help of TOUTATIS which includes a densification model and a swelling model. The latter can only be used in standard operating conditions. However, the processing structure of this modulus provides the possibility to include any type of model corresponding with other operating conditions. In last, we show the result of these fuel volume variations on the cladding mechanical conditions. (author). 25 refs., 89 figs., 2 tabs., 12 photos., 5 appends.

  2. Hafnium implanted in iron .2.Isolated Hafnium Nitrogen Complexes

    NARCIS (Netherlands)

    de Bakker, J.M.G.J.; Pleiter, F; Smulders, P.J M

    1993-01-01

    We have used the perturbed angular correlation technique to study the interaction of interstitially diffusing nitrogen atoms with substitutional hafnium atoms implanted in iron. It was found that after post-implantation of 250 eV nitrogen ions at 450 K, isolated HfVN(x) complexes with x = 1, 2 are

  3. Hafnium radioisotope recovery from irradiated tantalum

    Science.gov (United States)

    Taylor, Wayne A.; Jamriska, David J.

    2001-01-01

    Hafnium is recovered from irradiated tantalum by: (a) contacting the irradiated tantalum with at least one acid to obtain a solution of dissolved tantalum; (b) combining an aqueous solution of a calcium compound with the solution of dissolved tantalum to obtain a third combined solution; (c) precipitating hafnium, lanthanide, and insoluble calcium complexes from the third combined solution to obtain a first precipitate; (d) contacting the first precipitate of hafnium, lanthanide and calcium complexes with at least one fluoride ion complexing agent to form a fourth solution; (e) selectively adsorbing lanthanides and calcium from the fourth solution by cationic exchange; (f) separating fluoride ion complexing agent product from hafnium in the fourth solution by adding an aqueous solution of ferric chloride to obtain a second precipitate containing the hafnium and iron; (g) dissolving the second precipitate containing the hafnium and iron in acid to obtain an acid solution of hafnium and iron; (h) selectively adsorbing the iron from the acid solution of hafnium and iron by anionic exchange; (i) drying the ion exchanged hafnium solution to obtain hafnium isotopes. Additionally, if needed to remove residue remaining after the product is dried, dissolution in acid followed by cation exchange, then anion exchange, is performed.

  4. Synchrotron X-ray diffraction investigations on strains in the oxide layer of an irradiated Zircaloy fuel cladding

    Science.gov (United States)

    Chollet, Mélanie; Valance, Stéphane; Abolhassani, Sousan; Stein, Gene; Grolimund, Daniel; Martin, Matthias; Bertsch, Johannes

    2017-05-01

    For the first time the microstructure of the oxide layer of a Zircaloy-2 cladding after 9 cycles of irradiation in a boiling water reactor has been analyzed with synchrotron micro-X-ray diffraction. Crystallographic strains of the monoclinic and to some extent of the tetragonal ZrO2 are depicted through the thick oxide layer. Thin layers of sub-oxide at the oxide-metal interface as found for autoclave-tested samples and described in the literature, have not been observed in this material maybe resulting from irradiation damage. Shifts of selected diffraction peaks of the monoclinic oxide show that the uniform strain produced during oxidation is orientated in the lattice and displays variations along the oxide layer. Diffraction peaks and their shifts from families of diffracting planes could be translated into a virtual tensor. This virtual tensor exhibits changes through the oxide layer passing by tensile or compressive components.

  5. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, Jason Dean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barani, Tommaso [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pizzocri, Davide [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  6. Hafnium isotope stratigraphy of ferromanganese crusts

    Science.gov (United States)

    Lee, D.-C.; Halliday, A.N.; Hein, J.R.; Burton, K.W.; Christensen, J.N.; Gunther, D.

    1999-01-01

    A Cenozoic record of hafnium isotopic compositions of central Pacific deep water has been obtained from two ferromanganese crusts. The crusts are separated by more than 3000 kilometers but display similar secular variations. Significant fluctuations in hafnium isotopic composition occurred in the Eocene and Oligocene, possibly related to direct advection from the Indian and Atlantic oceans. Hafnium isotopic compositions have remained approximately uniform for the past 20 million years, probably reflecting increased isolation of the central Pacific. The mechanisms responsible for the increase in 87Sr/86Sr in seawater through the Cenozoic apparently had no effect on central Pacific deep-water hafnium.

  7. Hafnium isotope stratigraphy of ferromanganese crusts

    Science.gov (United States)

    Lee; Halliday; Hein; Burton; Christensen; Gunther

    1999-08-13

    A Cenozoic record of hafnium isotopic compositions of central Pacific deep water has been obtained from two ferromanganese crusts. The crusts are separated by more than 3000 kilometers but display similar secular variations. Significant fluctuations in hafnium isotopic composition occurred in the Eocene and Oligocene, possibly related to direct advection from the Indian and Atlantic oceans. Hafnium isotopic compositions have remained approximately uniform for the past 20 million years, probably reflecting increased isolation of the central Pacific. The mechanisms responsible for the increase in (87)Sr/(86)Sr in seawater through the Cenozoic apparently had no effect on central Pacific deep-water hafnium.

  8. Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection; Avaliacao de integridade de revestimentos de combustiveis de reatores de pesquisa e teste de materiais utilizando o ensaio de correntes parasitas

    Energy Technology Data Exchange (ETDEWEB)

    Alencar, Donizete Anderson de

    2004-07-01

    A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

  9. Silver-hafnium braze alloy

    Science.gov (United States)

    Stephens, Jr., John J.; Hosking, F. Michael; Yost, Frederick G.

    2003-12-16

    A binary allow braze composition has been prepared and used in a bonded article of ceramic-ceramic and ceramic-metal materials. The braze composition comprises greater than approximately 95 wt % silver, greater than approximately 2 wt % hafnium and less than approximately 4.1 wt % hafnium, and less than approximately 0.2 wt % trace elements. The binary braze alloy is used to join a ceramic material to another ceramic material or a ceramic material, such as alumina, quartz, aluminum nitride, silicon nitride, silicon carbide, and mullite, to a metal material, such as iron-based metals, cobalt-based metals, nickel-based metals, molybdenum-based metals, tungsten-based metals, niobium-based metals, and tantalum-based metals. A hermetic bonded article is obtained with a strength greater than 10,000 psi.

  10. Transactions of the second technical exchange meeting on fuel- and clad-motion diagnostics for LMFBR safety test facilities

    Energy Technology Data Exchange (ETDEWEB)

    DeVolpi, A. (comp.)

    1976-01-01

    Papers are presented which deal with diagnostic requirements and fuel motion monitoring capabilities of hodoscopes, coded aperture systems, x-ray radiography, and in-core detectors. Separate abstracts and indexing were prepared for each paper. (DG)

  11. Past research and fabrication conducted at SCK-CEN on ferritic ODS alloys used as cladding for FBR's fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    De Bremaecker, Anne, E-mail: adbremae@sckcen.be [Studiecentrum voor Kernenergie-Centre d' Etude de l' Energie Nucleaire (SCK-CEN), NMS, Mol (Belgium)

    2012-09-15

    In the 1960s in the frame of the sodium-cooled fast breeders, SCK-CEN decided to develop claddings made with ferritic stainless materials because of their specific properties, namely a higher thermal conductivity, a lower thermal expansion, a lower tendency to He-embrittlement, and a lower swelling than the austenitic stainless steels. To enhance their lower creep resistance at 650-700 Degree-Sign C arose the idea to strengthen the microstructure by oxide dispersions. This was the starting point of an ambitious programme where both the matrix and the dispersions were optimized. A purely ferritic 13 wt% Cr matrix was selected and its mechanical strength was improved through addition of ferritizing elements. Results of tensile and stress-rupture tests showed that Ti and Mo were the most beneficial elements, partly because of the chi-phase precipitation. In 1973 the optimized matrix composition was Fe-13Cr-3.5Ti-2Mo. To reach creep properties similar to those of AISI 316, different dispersions and methods were tested: internal oxidation (that was not conclusive), and the direct mixing of metallic and oxide powders (Al{sub 2}O{sub 3}, MgO, ZrO{sub 2}, TiO{sub 2}, ZrSiO{sub 4}) followed by pressing, sintering, and extrusion. The compression and extrusion parameters were determined: extrusion as hollow at 1050 Degree-Sign C, solution annealing at 1050 Degree-Sign C/15 min, cleaning, cold drawing to the final dimensions with intermediate annealings at 1050 Degree-Sign C, final annealing at 1050 Degree-Sign C, straightening and final aging at 800 Degree-Sign C. The choice of titania and yttria powders and their concentrations were finalized on the basis of their out-of-pile and in-pile creep and tensile strength. As soon as a resistance butt welding machine was developed and installed in a glove-box, fuel segments with PuO{sub 2} were loaded in Belgian MTR BR2. The fabrication parameters were continuously optimized: milling and beating, lubrication, cold drawing (partial

  12. Formulation and method for preparing gels comprising hydrous hafnium oxide

    Science.gov (United States)

    Collins, Jack L; Hunt, Rodney D; Montgomery, Frederick C

    2013-08-06

    Formulations useful for preparing hydrous hafnium oxide gels contain a metal salt including hafnium, an acid, an organic base, and a complexing agent. Methods for preparing gels containing hydrous hafnium oxide include heating a formulation to a temperature sufficient to induce gel formation, where the formulation contains a metal salt including hafnium, an acid, an organic base, and a complexing agent.

  13. Process for separating hafnium and zirconium

    NARCIS (Netherlands)

    Xiao, Y.; Van Sandwijk, A.

    2010-01-01

    The invention is directed to a process for separating a mixture comprising hafnium and zirconium. The process of the present invention comprises a step in which a molten metal phase comprising zirconium and hafnium dissolved in a first metal M1 and a second metal M2 is contacted with a molten salt

  14. Preliminary results for the Co-Rolling process fabrication of plate-type nuclear fuel based in U-10Mo monolithic meat and zircaloy-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Pedrosa, Tercio A.; Brina, Jose Giovanni M.; Paula, Joao Bosco de; Lameiras, Fernando S.; Ferraz, Wilmar B., E-mail: tap@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    The fabrication process of plate-type nuclear fuel with monolithic meat is under development at CDTN. The U-10Mo alloy was chosen as the meat material due to its high density, corrosion resistance and the higher dimensional stability proportioned by the metastable gamma phase, which presents a lesser extension of the breakaway swelling phenomena occurrence during irradiation tests. The monolithic meat was cut from an U-10Mo ingot, that was induction melted at CDTN. The co-rolling process was adopted due to the higher mechanical properties and melting point of the Zircalloy-4 cladding material, which presents a lesser discrepancy in relation to the meat material properties, when compared to the aluminum 6061 alloy. Preliminary plates were obtained by means of the co-rolling process, performed at 650, 800, 950 deg C with total thickness reduction of 80%, followed by a pickling step and cold co-rolling passes. The plates were characterized through bending tests, optical microscopy and radiography. The co-rolling temperature of 800 deg C presented the best results, with a homogeneous distribution of the total thickness reduction between the cover plates and the meat, and the absence of delamination in the bending test samples. It was observed the occurrence of meat thickening in its ends, according to its longitudinal axle, parallel to the rolling direction, that is known as the {sup d}og bone{sup ,} for the three co-rolling temperatures. (author)

  15. Study of the uniform corrosion of an aluminium alloy used for the fuel cladding of the Jules Horowitz experimental reactor; Etude de la corrosion uniforme d'un alliage d'aluminium utilise comme gainage du combustible nucleaire du reacteur experimental Jules Horowitz

    Energy Technology Data Exchange (ETDEWEB)

    Wintergerst, M. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMI), 91 - Gif-sur-Yvette (France)

    2008-07-01

    For the Jules Horowitz new material testing reactor, an aluminium base alloy, AlFeNi, will be used for the cladding of the fuel plates. Taking into account the thermal properties of the alloy and of its oxide, the corrosion of the fuel cans presents many problems. The aim of this thesis is to provide a growing kinetic of the oxide layer at the surface of the AlFeNi fuel can in order to predict the life time of fuel element. Thus the mechanism of degradation of the cladding will be describe in order to integrate the different parameters of the operating reactor. (A.L.B.)

  16. In-situ Monitoring of Sub-cooled Nucleate Boiling on Fuel Cladding Surface in Water at 1 bar and 130 bars using Acoustic Emission Method

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Seung Heon; Wu, Kaige; Shim, Hee-Sang; Lee, Deok Hyun; Hur, Do Haeng [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Crud deposition increases through a sufficient corrosion product supply around the steam-liquid interface of a boiling bubble. Therefore, the understanding of this SNB phenomenon is important for effective and safe operation of nuclear plants. The experimental SNB studies have been performed in visible conditions at a low pressure using a high speed video camera. Meanwhile, an acoustic emission (AE) method is an on-line non-destructive evaluation method to sense transient elastic wave resulting from a rapid release of energy within a dynamic process. Some researchers have investigated boiling phenomena using the AE method. However, their works were performed at atmospheric pressure conditions. Therefore, the objective of this work is for the first time to detect and monitor SNB on fuel cladding surface in simulated PWR primary water at 325 .deg. C and 130 bars using an AE technique. We successfully observed the boiling AE signals in primary water at 1 bar and 130 bars using AE technique. Visualization test was performed effectively to identify a correlation between water boiling phenomenon and AE signals in a transparent glass cell at 1 bar, and the boiling AE signals were in good agreement with the boiling behavior. Based on the obtained correlations at 1 bar, the AE signals obtained at 130 bars were analyzed. The boiling density and size of the AE signals at 130 bars were decreased by the flow parameters. However, overall AE signals showed characteristics and a trend similar to the AE signals at 1 bar. This indicates that boiling AE signals are detected successfully at 130 bars, and the AE technique can be effectively implemented in non-visualized condition at high pressures.

  17. Thermal Expansion of Hafnium Carbide

    Science.gov (United States)

    Grisaffe, Salvatore J.

    1960-01-01

    Since hafnium carbide (HfC) has a melting point of 7029 deg. F, it may have many high-temperature applications. A literature search uncovered very little information about the properties of HfC, and so a program was initiated at the Lewis Research Center to determine some of the physical properties of this material. This note presents the results of the thermal expansion investigation. The thermal-expansion measurements were made with a Gaertner dilatation interferometer calibrated to an accuracy of +/- 1 deg. F. This device indicates expansion by the movement of fringes produced by the cancellation and reinforcement of fixed wave-length light rays which are reflected from the surfaces of two parallel quartz glass disks. The test specimens which separate these disks are three small cones, each approximately 0.20 in. high.

  18. Metallurgical and mechanical behaviours of PWR fuel cladding tube oxidised at high temperature; Comportements metallurqigue et mecanique des materiaux de gainage du combustible REP oxydes a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Stern, A

    2007-12-15

    Zirconium alloys are used as cladding materials in Pressurized Water Reactors (PWR). As they are submitted to very extreme conditions, it is necessary to check their behaviour and especially to make sure they meet the safety criteria. They are therefore studied under typical in service-loadings but also under accidental loadings. In one of these accidental scenarios, called Loss of Coolant Accident (LOCA) the cladding temperature may increase above 800 C, in a steam environment, and decrease before a final quench of the cladding. During this temperature transient, the cladding is heavily oxidised, and the metallurgical changes lead to a decrease of the post quench mechanical properties. It is then necessary to correlate this drop in residual ductility to the metallurgical evolutions. This is the problem we want to address in this study: the oxidation of PWR cladding materials at high temperature in a steam environment and its consequences on post quench mechanical properties. As oxygen goes massively into the metallic part - a zirconia layer grows at the same time - during the high temperature oxidation, the claddings tubes microstructure shows three different phases that are the outer oxide layer (zirconia) and the inner metallic phases ({alpha}(O) and 'ex {beta}') - with various mechanical properties. In order to reproduce the behaviour of this multilayered material, the first part of this study consisted in creating samples with different - but homogeneous in thickness - oxygen contents, similar to those observed in the different phases of the real cladding. The study was especially focused on the {beta}-->{alpha} phase transformation upon cooling and on the resulting microstructures. A mechanism was proposed to describe this phase transformation. For instance, we conclude that for our oxygen enriched samples, the phase transformation kinetics upon cooling are ruled by the oxygen partitioning between the two allotropic phases. Then, these materials

  19. Formation and control of stoichiometric hafnium nitride thin films by direct sputtering of hafnium nitride target

    CERN Document Server

    Gotoh, Y; Ishikawa, J; Liao, M Y

    2003-01-01

    Hafnium nitride thin films were prepared by radio-frequency sputter deposition with a hafnium nitride target. Deposition was performed with various rf powers, argon pressures, and substrate temperatures, in order to investigate the influences of these parameters on the film properties, particularly the nitrogen composition. It was found that stoichiometric hafnium nitride films were formed at an argon gas pressure of less than 2 Pa, irrespective of the other deposition parameters within the range investigated. Maintaining the nitrogen composition almost stoichiometric, orientation, stress, and electrical resistivity of the films could be controlled with deposition parameters. (author)

  20. Ceramic Coatings for Clad (The C3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sickafus, Kurt E. [Univ. of Tennessee, Knoxville, TN (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Miller, Larry [Univ. of Tennessee, Knoxville, TN (United States); Weber, Bill [Univ. of Tennessee, Knoxville, TN (United States); Zhang, Yanwen [Univ. of Tennessee, Knoxville, TN (United States); Patel, Maulik [Univ. of Tennessee, Knoxville, TN (United States); Motta, Arthur [Pennsylvania State Univ., University Park, PA (United States); Wolfe, Doug [Pennsylvania State Univ., University Park, PA (United States); Fratoni, Max [Univ. of California, Berkeley, CA (United States); Raj, Rishi [Univ. of Colorado, Boulder, CO (United States); Plunkett, Kenneth [Univ. of Colorado, Boulder, CO (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Hollis, Kendall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Andy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Chris [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Comstock, Robert [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Partezana, Jonna [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Whittle, Karl [Univ. of Sheffield (United Kingdom); Preuss, Michael [Univ. of Manchester (United Kingdom); Withers, Philip [Univ. of Manchester (United Kingdom); Wilkinson, Angus [Univ. of Oxford (United Kingdom); Donnelly, Stephen [Univ. of Huddersfield (United Kingdom); Riley, Daniel [Australian Nuclear Science and Technology Organisation, Syndney (Australia)

    2017-02-14

    The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectives of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as

  1. Investigation of Melting Dynamics of Hafnium Clusters.

    Science.gov (United States)

    Ng, Wei Chun; Lim, Thong Leng; Yoon, Tiem Leong

    2017-03-27

    Melting dynamics of hafnium clusters are investigated using a novel approach based on the idea of the chemical similarity index. Ground state configurations of small hafnium clusters are first derived using Basin-Hopping and Genetic Algorithm in the parallel tempering mode, employing the COMB potential in the energy calculator. These assumed ground state structures are verified by using the Low Lying Structures (LLS) method. The melting process is carried out either by using the direct heating method or prolonged simulated annealing. The melting point is identified by a caloric curve. However, it is found that the global similarity index is much more superior in locating premelting and total melting points of hafnium clusters.

  2. Hafnium transistor process design for neural interfacing.

    Science.gov (United States)

    Parent, David W; Basham, Eric J

    2009-01-01

    A design methodology is presented that uses 1-D process simulations of Metal Insulator Semiconductor (MIS) structures to design the threshold voltage of hafnium oxide based transistors used for neural recording. The methodology is comprised of 1-D analytical equations for threshold voltage specification, and doping profiles, and 1-D MIS Technical Computer Aided Design (TCAD) to design a process to implement a specific threshold voltage, which minimized simulation time. The process was then verified with a 2-D process/electrical TCAD simulation. Hafnium oxide films (HfO) were grown and characterized for dielectric constant and fixed oxide charge for various annealing temperatures, two important design variables in threshold voltage design.

  3. Use of hafnium in control bars of nuclear reactors; Uso de hafnio en barras de control de reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin-mx

    2003-07-01

    Recently the use of hafnium as neutron absorber material in nuclear reactors has been reason of investigation by virtue of that this material has nuclear properties as to the neutrons absorption and structural that can prolong the useful life of the control mechanisms of the nuclear reactors. In this work some of those more significant hafnium properties are presented like nuclear material. Also there are presented calculations carried out with the HELIOS code for fuel cells of uranium oxide and of uranium and plutonium mixed oxides under controlled conditions with conventional bars of boron carbide and also with similar bars to which are substituted the absorbent material by metallic hafnium, the results are presented in this work. (Author)

  4. Cladding embrittlement during postulated loss-of-coolant accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  5. Quantification of hydrogen distribution with the nuclear microprobe of the Pierre Sue Laboratory in the thickness of the PWR fuel cladding in zirconium alloy; Quantification de la repartition de l'hydrogene a la microsonde nucleaire du Laboratoire Pierre Sue dans l'epaisseur du tube de gainage du combustible des REP en alliage de zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Raepsaet, C. [Laboratoire Pierre Sue (DSM/DRECAM/LPS) - CEA Saclay, 91 - Gif-sur-Yvette (France); Bossis, Ph. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMI/LM2E), 91 - Gif sur Yvette (France); Hamon, D.; Bechade, J.L.; Brachet, J.C. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SRMA/LA2M), 91 - Gif sur Yvette (France)

    2007-07-01

    In a first part of this study, are detailed the general principles of the specific technique ERDA (Elastic Recoil Detection Analysis) used in the Pierre Sue Laboratory. Then, the contribution of this technique is illustrated with two studies examples on the behaviour of PWR nuclear fuel cladding. (O.M.)

  6. Work Function Calculation For Hafnium- Barium System

    Directory of Open Access Journals (Sweden)

    K.A. Tursunmetov

    2015-08-01

    Full Text Available The adsorption process of barium atoms on hafnium is considered. A structural model of the system is presented and on the basis of calculation of interaction between ions dipole system the dependence of the work function on the coating.

  7. Calibration of the Lutetium-Hafnium Clock

    National Research Council Canada - National Science Library

    Erik Scherer; Carsten Münker; Klaus Mezger

    2001-01-01

    ... −1 , in agreement with the two most recent decay-counting experiments. Lutetium-hafnium ages that are based on the previously used λ 176 Lu of 1.93 × 10 −11 to 1.94 × 10 −11 year −1 are thus ∼4...

  8. Percolation conductivity in hafnium sub-oxides

    Energy Technology Data Exchange (ETDEWEB)

    Islamov, D. R., E-mail: damir@isp.nsc.ru; Gritsenko, V. A., E-mail: grits@isp.nsc.ru [Rzhanov Institute of Semiconductor Physics, Siberian Branch of Russian Academy of Sciences, Novosibirsk 630090 (Russian Federation); Novosibirsk State University, Novosibirsk 630090 (Russian Federation); Cheng, C. H. [Department of Mechatronic Technology, National Taiwan Normal University, Taipei 106, Taiwan (China); Chin, A., E-mail: albert-achin@hotmail.com [National Chiao Tung University, Hsinchu 300, Taiwan (China)

    2014-12-29

    In this study, we demonstrated experimentally that formation of chains and islands of oxygen vacancies in hafnium sub-oxides (HfO{sub x}, x < 2) leads to percolation charge transport in such dielectrics. Basing on the model of Éfros-Shklovskii percolation theory, good quantitative agreement between the experimental and theoretical data of current-voltage characteristics was achieved. Based on the percolation theory suggested model shows that hafnium sub-oxides consist of mixtures of metallic Hf nanoscale clusters of 1–2 nm distributed onto non-stoichiometric HfO{sub x}. It was shown that reported approach might describe low resistance state current-voltage characteristics of resistive memory elements based on HfO{sub x}.

  9. Development of Mechanical Loading Device for testing the zirconium cladding under the pellet-cladding interaction conditions

    Directory of Open Access Journals (Sweden)

    V. I. Solonin

    2014-01-01

    Full Text Available Currently, there is a tendency of transition to the long-term cycles of operation with fuel and to the new transitional modes. This fact requires extra experimental validation for design of fuel rods. New operating conditions are expanding operability requirements of claddings.To implement the experimental techniques the Mechanical Loading Device (MLD was developed, capable of providing the conditions of stress-strain state similar to the pellet-cladding interaction (PCI during operation of the reactor.Complex strain state of a fuel rod cladding is simulated by the impacting force on the plunger and then on the simulator of the fuel pellet. The simulator is made of interposer of zirconium and the inset made of ceramic - aluminum oxide. Mechanical properties of the aluminum oxide are similar to the material of the fuel pellet - uranium dioxide. Experiments conducted on the layout and the MLD as such have shown that a stress-strain state matches with that of under operating conditions of the fuel rod in the reactor.The developed device and test method allows us to simulate a wide range of reactor transient modes. Claddings can be used both in the delivered state, and with the further preparation, including the exposure in nuclear reactor. MLD design enables us to carry out experiments with the presence of an aggressive environment inside the cladding, simulating the presence of gaseous fission products in the fuel rod.For further the development of this research it is necessary to design the laboratory complex for MLD. Extra computational verification experiment is needed as well. In particular, stresses in the cladding achieved during the experiment ought to be calculated. Calculated stresses are required to make project justification on the performance capability of fuel rods.

  10. Hafnium implanted in iron 1. Lattice location and annealing behavior

    NARCIS (Netherlands)

    de Bakker, J.M.G.J.; Pleiter, F; Smulders, P.J M

    1993-01-01

    Perturbed angular correlation, Rutherford backscattering and channelling experiments were conducted to study the lattice location and annealing behaviour of 110 keV hafnium ions implanted into iron single crystals. It was found that a fraction of 11-25% of the implanted hafnium atoms are located at

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

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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.

  12. Mechanical and fracture behavior of nuclear fuel cladding in terms of transport and temporary dry storage; Comportamiento mecanio y en fractura de vainas de combustible nuclear en condiciones de transporte y almacenamiento temporal en seco

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz Hervias, J.; Martin Rengel, M. A.; Gomez, F. J.

    2012-11-01

    In this work, the most relevant results of a research project on the mechanical and fracture behavior of cladding in transport and dry storage conditions are summarized. the project is being carried out at Universidad Politecnica de Madrid in collaboration with ENUSA, ENRESA and CSN. Non-irradiated cladding is investigated. The main objective is to determine a failure criterion of cladding as a function of hydrogen content, temperature and strain rate. (Author)

  13. Influence of texture on fracture toughness of zircaloy cladding

    Energy Technology Data Exchange (ETDEWEB)

    Grigoriev, V. [Studsvik Material AB, Nykoeping (Sweden); Andersson, Stefan [Royal Inst. of Tech., Stockholm (Sweden)

    1997-06-01

    The correlation between texture and fracture toughness of Zircaloy 2 cladding has been investigated in connection with axial cracks in fuel rods. The texture of the cladding determines the anisotropy of plasticity of the cladding which, in turn, should influence the strain conditions at the crack-tip. Plastic strains in the cladding under uniaxial tension were characterised by means of the anisotropy constants F, G and H calculated according to Hill`s theory. Test temperatures between 20 and 300 deg C do not influence the F, G and H values. Any significant effect of hydrogen (about 500 wtppm) on the anisotropy constants F, G and H has not been revealed at a test temperature of 300 deg C. The results, obtained for stress-relieved and recrystallized cladding with different texture, show an obvious influence of texture on the fracture toughness of Zircaloy cladding. A higher fracture toughness has been found for cladding with more radial texture. With a 2 page summary in Swedish. 32 refs, 18 figs.

  14. Chemical Dissolution of Simulant FCA Cladding and Plates

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pierce, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-11-08

    The Savannah River Site (SRS) has received some fast critical assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) for disposition. Among the JAEA FCA fuel are approximately 7090 rectangular Stainless Steel clad fuel elements. Each element has an internal Pu-10.6Al alloy metal wafer. The thickness of each element is either 1/16 inch or 1/32 inch. The dimensions of each element ranges from 2 inches x 1 inch to 2 inches x 4 inches. This report discusses the potential chemical dissolution of the FCA clad material or stainless steel. This technology uses nitric acid-potassium fluoride (HNO3-KF) flowsheets of H-Canyon to dissolve the FCA elements from a rack of materials. Historically, dissolution flowsheets have aimed to maximize Pu dissolution rates while minimizing stainless steel dissolution (corrosion) rates. Because the FCA cladding is made of stainless steel, this work sought to accelerate stainless steel dissolution.

  15. Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors; Quantification de la repartition de l'hydrogene a la microsonde nucleaire du Laboratoire Pierre Sue dans l'epaisseur de tubes de gainage du combustible des REP en alliage de zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Raepsaet, C. [CEA Saclay, Dept. de Recherche sur l' Etat Condense, les Atomes et les Molecules (DSM/DRECAM/LPS-CNRS) UMR9956, 91 - Gif sur Yvette (France); Bossis, Ph. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMULM2E), 91 - Gif-sur-Yvette (France); Hamon, D.; Bechade, J.L.; Brachet, J.C. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SRMALA2M), 91 - Gif-sur-Yvette (France)

    2007-07-01

    Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

  16. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  17. Non-destructive Residual Stress Analysis Around The Weld-Joint of Fuel Cladding Materials of ZrNbMoGe Alloys

    Directory of Open Access Journals (Sweden)

    Parikin

    2003-08-01

    Full Text Available The residual stress measurements around weld-joint of ZrNbMoGe alloy have been carried out by using X-ray diffraction technique in PTBIN-BATAN. The research was performed to investigate the structure of a cladding material with high temperature corrosion resistance and good weldability. The equivalent composition of the specimens (in %wt. was 97.5%Zr1%Nb1%Mo½%Ge. Welding was carried out by using TIG (tungsten inert gas technique that completed butt-joint with a current 20 amperes. Three region tests were taken in specimen while diffraction scanning, While diffraction scanning, tests were performed on three regions, i.e., the weldcore, the heat-affected zone (HAZ and the base metal. The reference region was determined at the base metal to be compared with other regions of the specimen, in obtaining refinement structure parameters. Base metal, HAZ and weldcore were diffracted by X-ray, and lattice strain changes were calculated by using Rietveld analysis program. The results show that while the quantity of minor phases tend to increase in the direction from the base metal to the HAZ and to the weldcore, the quantity of the ZrGe phase in the HAZ is less than the quantity of the ZrMo2 phase due to tGe element evaporation. The residual stress behavior in the material shows that minor phases, i.e., Zr3Ge and ZrMo2, are more dominant than the Zr matrix. The Zr3Ge and ZrMo2 experienced sharp straining, while the Zr phase was weak-lined from HAZ to weldcore. The hydrostatic residual stress ( in around weld-joint of ZrNbMoGe alloy is compressive stress which has minimum value at about -2.73 GPa in weldcore region

  18. Adsorbsi Hafnium (Hf) Dalam Resin Penukar Anion Dowex-1x8 Adsorbtion of Hafnium (Hf) in Dowex - 1x8 Anion Exchange Resin

    OpenAIRE

    Susiantini, Endang; Setyadji, Moch

    2014-01-01

    Hafnium memiliki titik lebur yang tinggi dan kemampuan menyerap neutron per luas penampang 600 kali lebih besar dari Zr sehingga berpotensi untuk dimanfaatkan sebagai salah satu bahan batang pengendali reaksi fisi nuklir. Berbagai metode pemurnian Hf dari Zr telah dikembangkan salah satunya adalah dengan menggunakan resin penukar ion. Pada penelitian ini digunakan umpan berbentuk sulfat dari hafnium murni dan hafnium-zirkonium campuran hasil proses pengolahan pasir zirkon. Umpan hafnium sulfa...

  19. Thermal hydraulic-Mechanic Integrated Simulation for Advanced Cladding Thermal Shock Fracture Analysis during Reflood Phase in LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Son, Seong Min; Lee, You Ho; Cho, Jae Wan; Lee, Jeong Ik [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This study suggested thermal hydraulic-mechanical integrated stress based methodology for analyzing the behavior of ATF type claddings by SiC-Duplex cladding LBLOCA simulation. Also, this paper showed that this methodology could predict real experimental result well. That concept for enhanced safety of LWR called Advanced Accident-Tolerance Fuel Cladding (ATF cladding, ATF) is researched actively. However, current nuclear fuel cladding design criteria for zircaloy cannot be apply to ATF directly because those criteria are mainly based on limiting their oxidation. So, the new methodology for ATF design criteria is necessary. In this study, stress based analysis methodology for ATF cladding design criteria is suggested. By simulating LBLOCA scenario of SiC cladding which is the one of the most promising candidate of ATF. Also we'll confirm our result briefly through comparing some facts from other experiments. This result is validating now. Some of results show good performance with 1-D failure analysis code for SiC fuel cladding that already developed and validated by Lee et al,. It will present in meeting. Furthermore, this simulation presented the possibility of understanding the behavior of cladding deeper. If designer can predict the dangerous region and the time precisely, it may be helpful for designing nuclear fuel cladding geometry and set safety criteria.

  20. Development of data base with mechanical properties of un- and pre-irradiated VVER cladding

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L.; Kaplar, E.; Lioutov, K. [Nuclear Safety Inst. of Russian Research Centre, Moscow (Russian Federation). Kurchatov Inst.; Smirnov, V.; Prokhorov, V.; Goryachev, A. [State Research Centre, Dimitrovgrad (Russian Federation). Research Inst. of Atomic Reactors

    1998-03-01

    Analysis of recent RIA test with PWR and VVER high burnup fuel, performed at CABRI, NSRR, IGR reactors has shown that the data base with mechanical properties of the preirradiated cladding is necessary to interpret the obtained results. During 1997 the corresponding cycle of investigations for VVER clad material was performed by specialists of NSI RRC KI and RIAR in cooperation with NRC (USA), IPSN (France) in two directions: measurements of mechanical properties of Zr-1%Nb preirradiated cladding versus temperature and strain rate; measurements of failure parameters for gas pressurized cladding tubes. Preliminary results of these investigations are presented in this paper.

  1. Hafnium transistor design for neural interfacing.

    Science.gov (United States)

    Parent, David W; Basham, Eric J

    2008-01-01

    A design methodology is presented that uses the EKV model and the g(m)/I(D) biasing technique to design hafnium oxide field effect transistors that are suitable for neural recording circuitry. The DC gain of a common source amplifier is correlated to the structural properties of a Field Effect Transistor (FET) and a Metal Insulator Semiconductor (MIS) capacitor. This approach allows a transistor designer to use a design flow that starts with simple and intuitive 1-D equations for gain that can be verified in 1-D MIS capacitor TCAD simulations, before final TCAD process verification of transistor properties. The DC gain of a common source amplifier is optimized by using fast 1-D simulations and using slower, complex 2-D simulations only for verification. The 1-D equations are used to show that the increased dielectric constant of hafnium oxide allows a higher DC gain for a given oxide thickness. An additional benefit is that the MIS capacitor can be employed to test additional performance parameters important to an open gate transistor such as dielectric stability and ionic penetration.

  2. Evaluation of missing pellet surface geometry on cladding stress distribution and magnitude

    Energy Technology Data Exchange (ETDEWEB)

    Capps, Nathan [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Montgomery, Robert [Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Sunderland, Dion [Pacific Northwest National Laboratory, Richland, WA 99354 (United States); ANATECH Corp, San Diego, CA 92121 (United States); Pytel, Martin [Electric Power Research Institute, Palo Alto, CA 94304 (United States); Wirth, Brian D. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States)

    2016-08-15

    Highlights: • Stress concentrations are related to pellet defect geometries. • The presence of radial cracks cause increases in stress concentration. • Increasing the size of MPS causes an increase hoop stress concentrations. - Abstract: Missing pellet surface (MPS) defects are local geometric defects in nuclear fuel pellets that result from pellet mishandling or manufacturing. The presence of MPS defects can cause significant clad stress concentrations that can lead to through-wall cladding failure for certain combinations of fuel burnup, and reactor power level or power change. Consequently, the impact of MPS defects has limited the rate of power increase, or ramp rate, in both pressurized and boiling water reactors (PWRs and BWRs, respectively). Improved three-dimensional (3-D) fuel performance models of MPS defect geometry can provide better understanding of the probability for pellet clad mechanical interaction (PCMI), and correspondingly the available margin against cladding failure by stress corrosion cracking (SCC). The Consortium of Advanced Simulations of Light Water Reactors (CASL) has been developing the Bison-CASL fuel performance code to consider the inherently multi-physics and multi-dimensional mechanisms that control fuel behavior, including cladding stress concentrations resulting from MPS defects. This paper evaluates the cladding hoop stress distributions as a function of MPS defect geometry with discrete pellet radial cracks for a set of typical operating conditions in a PWR fuel rod. The results provide a first step toward a probabilistic approach to assess cladding failure during power maneuvers. This analysis provides insight into how varying pellet defect geometries affect the distribution of the cladding stress, as well as the temperature distributions within the fuel and clad; and are used to develop stress concentration factors for comparing 2-D and 3-D models.

  3. Results of irradiated cladding tests and clad plate experiments

    Energy Technology Data Exchange (ETDEWEB)

    Haggag, F.M.; Iskander, S.K.

    1988-01-01

    Two aspects critical to the fracture behavior of three-wire stainless steel cladding were investigated by the Heavy-Section Steel Technology (HSST) Program: (1) radiation effects on cladding strength and toughness, and (2) the response of mechanically loaded, flawed structures in the presence of cladding (clad plate experiments). Postirradiation testing results show that, in the test temperature range from /minus/125 to 288/degree/C, the yield strength increased, and ductility insignificantly increased, while there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing. Radiation damage decreased the Charpy upper-shelf energy by 15 to 20% and resulted in up to 28/degree/C shifts of the Charpy impact transition temperature. Results of irradiated 12.5-mm-thick compact specimens (0.5TCS) show consistent decreases in the ductile fracture toughness, J/sub Ic/, and the tearing modulus. Results from clad plate tests have shown that (1) a tough surface layer composed of cladding and/or heat-affected zone has arrested running flaws under conditions where unclad plates have ruptured, and (2) the residual load-bearing capacity of clad plates with large subclad flaws significantly exceeded that of an unclad plate. 13 figs., 1 tab.

  4. Hafnium binary alloys from experiments and first principles

    OpenAIRE

    Levy, Ohad; Hart, Gus L. W.; Curtarolo, Stefano

    2009-01-01

    Despite the increasing importance of hafnium in numerous technological applications, experimental and computational data on its binary alloys is sparse. In particular, data is scant on those binary systems believed to be phase separating. We performed a comprehensive study of 44 hafnium binary systems with alkali metals, alkaline earths, transition metals and metals, using high-throughput first principles calculations. These computations predict novel unsuspected compounds in six binary syste...

  5. COMPARISON OF CLADDING CREEP RUPTURE MODELS

    Energy Technology Data Exchange (ETDEWEB)

    P. Macheret

    2000-06-12

    The objective of this calculation is to compare several creep rupture correlations for use in calculating creep strain accrued by the Zircaloy cladding of spent nuclear fuel when it has been emplaced in the repository. These correlations are used to calculate creep strain values that are then compared to a large set of experimentally measured creep strain data, taken from four different research articles, making it possible to determine the best fitting correlation. The scope of the calculation extends to six different creep rupture correlations.

  6. Aerogel-clad optical fiber

    Science.gov (United States)

    Sprehn, Gregory A.; Hrubesh, Lawrence W.; Poco, John F.; Sandler, Pamela H.

    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.

  7. Nanoscale radiotherapy with hafnium oxide nanoparticles.

    Science.gov (United States)

    Maggiorella, Laurence; Barouch, Gilles; Devaux, Corinne; Pottier, Agnès; Deutsch, Eric; Bourhis, Jean; Borghi, Elsa; Levy, Laurent

    2012-09-01

    There is considerable interest in approaches that could improve the therapeutic window of radiotherapy. In this study, hafnium oxide nanoparticles were designed that concentrate in tumor cells to achieve intracellular high-energy dose deposit. Conventional methods were used, implemented in different ways, to explore interactions of these high-atomic-number nanoparticles and ionizing radiation with biological systems. Using the Monte Carlo simulation, these nanoparticles, when exposed to high-energy photons, were shown to demonstrate an approximately ninefold radiation dose enhancement compared with water. Importantly, the nanoparticles show satisfactory dispersion and persistence within the tumor and they form clusters in the cytoplasm of cancer cells. Marked antitumor activity is demonstrated in human cancer models. Safety is similar in treated and control animals as demonstrated by a broad program of toxicology evaluation. These findings, supported by good tolerance, provide the basis for developing this new type of nanoparticle as a promising anticancer approach in human patients.

  8. Cladding tube manufacturing technology

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, B.J.; Kim, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report 'Alloy Development for High Burnup Cladding.' Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs.

  9. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  10. A new cladding embrittlement criterion derived from ring compression tests

    Energy Technology Data Exchange (ETDEWEB)

    Herb, Joachim, E-mail: Joachim.Herb@grs.de; Sievers, Jürgen, E-mail: Juergen.Sievers@grs.de; Sonnenburg, Heinz-Günther, E-mail: Heinz-Guenther.Sonnenburg@grs.de

    2014-07-01

    Highlights: • Using FEM it was possible to simulate measured ring compression test data. • The FEM provides burst stresses from Zry-4, M5 and ZIRLO cladding. • The ratio of burst stresses to yield stresses was correlated. • The ratio depends linearly on the state of oxidation and hydriding. • The ratio of stresses at unity can be applied as embrittlement criterion. - Abstract: It is of regulatory interest to prevent the breaking of fuel rods in LOCA transients. In current regulations this is accomplished by limiting the oxidation during LOCA to such an extent that still some residual ductility is preserved in the fuel rod cladding. The current oxidation limit in German as well as in US regulations is set to 17% ECR (Equivalent Cladding Reacted) which aims at maintaining a residual ductility for oxidized claddings. Recent ANL tests have shown that the combination of both oxidation and additionally hydrogen up-take affects the transition to zero-ductility. Furthermore, the oxidation during LOCA transient is accompanied by a significant up-take of hydrogen (secondary hydriding) if the fuel rod bursts during this transient. This secondary hydriding affects the cladding in the vicinity of the burst opening. These findings necessitate a new criterion for preserving cladding's strength. This paper describes a method how to derive a criterion which assures the required residual mechanical strength of the cladding for LOCA transients. This method utilizes the experimental results of 102 ring compression tests (RCT) conducted at ANL and KIT. RCTs of various cladding materials, oxidation levels and hydrogen content were considered. The basic approach was to compare the RCT test data with finite element analyses using the code ADINA. Starting with the cladding oxidation model of Leistikov, both the layer structure of the cladding and the distribution of the oxygen among these layers were determined. The mechanical properties of these layers were taken from

  11. Cladding material, tube including such cladding material and methods of forming the same

    Science.gov (United States)

    Garnier, John E.; Griffith, George W.

    2016-03-01

    A multi-layered cladding material including a ceramic matrix composite and a metallic material, and a tube formed from the cladding material. The metallic material forms an inner liner of the tube and enables hermetic sealing of thereof. The metallic material at ends of the tube may be exposed and have an increased thickness enabling end cap welding. The metallic material may, optionally, be formed to infiltrate voids in the ceramic matrix composite, the ceramic matrix composite encapsulated by the metallic material. The ceramic matrix composite includes a fiber reinforcement and provides increased mechanical strength, stiffness, thermal shock resistance and high temperature load capacity to the metallic material of the inner liner. The tube may be used as a containment vessel for nuclear fuel used in a nuclear power plant or other reactor. Methods for forming the tube comprising the ceramic matrix composite and the metallic material are also disclosed.

  12. Stone cladding engineering

    CERN Document Server

    Sousa Camposinhos, Rui de

    2014-01-01

    This volume presents new methodologies for the design of dimension stone based on the concepts of structural design while preserving the excellence of stonemasonry practice in façade engineering. Straightforward formulae are provided for computing action on cladding, with special emphasis on the effect of seismic forces, including an extensive general methodology applied to non-structural elements. Based on the Load and Resistance Factor Design Format (LRDF), minimum slab thickness formulae are presented that take into consideration stress concentrations analysis based on the Finite Element Method (FEM) for the most commonly used modern anchorage systems. Calculation examples allow designers to solve several anchorage engineering problems in a detailed and objective manner, underlining the key parameters. The design of the anchorage metal parts, either in stainless steel or aluminum, is also presented.

  13. Ice-clad volcanoes

    Science.gov (United States)

    Waitt, Richard B.; Edwards, B.R.; Fountain, Andrew G.; Huggel, C.; Carey, Mark; Clague, John J.; Kääb, Andreas

    2015-01-01

    An icy volcano even if called extinct or dormant may be active at depth. Magma creeps up, crystallizes, releases gas. After decades or millennia the pressure from magmatic gas exceeds the resistance of overlying rock and the volcano erupts. Repeated eruptions build a cone that pokes one or two kilometers or more above its surroundings - a point of cool climate supporting glaciers. Ice-clad volcanic peaks ring the northern Pacific and reach south to Chile, New Zealand, and Antarctica. Others punctuate Iceland and Africa (Fig 4.1). To climb is irresistible - if only “because it’s there” in George Mallory’s words. Among the intrepid ascents of icy volcanoes we count Alexander von Humboldt’s attempt on 6270-meter Chimborazo in 1802 and Edward Whymper’s success there 78 years later. By then Cotopaxi steamed to the north.

  14. 40 CFR 471.90 - Applicability; description of the zirconium-hafnium forming subcategory.

    Science.gov (United States)

    2010-07-01

    ... zirconium-hafnium forming subcategory. 471.90 Section 471.90 Protection of Environment ENVIRONMENTAL... POINT SOURCE CATEGORY Zirconium-Hafnium Forming Subcategory § 471.90 Applicability; description of the zirconium-hafnium forming subcategory. This subpart applies to discharges of pollutants to waters of the...

  15. 40 CFR 421.330 - Applicability: Description of the primary zirconium and hafnium subcategory.

    Science.gov (United States)

    2010-07-01

    ... primary zirconium and hafnium subcategory. 421.330 Section 421.330 Protection of Environment ENVIRONMENTAL... CATEGORY Primary Zirconium and Hafnium Subcategory § 421.330 Applicability: Description of the primary zirconium and hafnium subcategory. The provisions of this subpart are applicable to discharges resulting...

  16. Friction surface cladding: development of a solid state cladding process

    NARCIS (Netherlands)

    van der Stelt, A.A.

    2014-01-01

    Many industries including automotive, aerospace, electronics, shipbuilding, offshore, railway and heavy equipment employ surface modification technologies to change the surface properties of a manufactured product. Often, the surface is covered (coated) with a dissimilar clad layer for this purpose

  17. Complete Non-Radioactive Operability Tests for Cladding Hull Chlorination

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Emory D [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Johnson, Jared A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hylton, Tom D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brunson, Ronald Ray [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunt, Rodney Dale [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); DelCul, Guillermo Daniel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bradley, Eric Craig [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Spencer, Barry B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    Non-radioactive operability tests were made to test the metal chlorination reactor and condenser and their accessories using batch chlorinations of non-radioactive cladding samples and to identify optimum operating practices and components that need further modifications prior to installation of the equipment into the hot cell for tests on actual used nuclear fuel (UNF) cladding. The operability tests included (1) modifications to provide the desired heating and reactor temperature profile; and (2) three batch chlorination tests using, respectively, 100, 250, and 500 g of cladding. During the batch chlorinations, metal corrosion of the equipment was assessed, pressurization of the gas inlet was examined and the best method for maintaining solid salt product transfer through the condenser was determined. Also, additional accessing equipment for collection of residual ash and positioning of the unit within the hot cell were identified, designed, and are being fabricated.

  18. Pulsed Magnetic Welding for Advanced Core and Cladding Steel

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Guoping [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong [Univ. of Florida, Gainesville, FL (United States)

    2013-12-19

    To investigate a solid-state joining method, pulsed magnetic welding (PMW), for welding the advanced core and cladding steels to be used in Generation IV systems, with a specific application for fuel pin end-plug welding. As another alternative solid state welding technique, pulsed magnetic welding (PMW) has not been extensively explored on the advanced steels. The resultant weld can be free from microstructure defects (pores, non-metallic inclusions, segregation of alloying elements). More specifically, the following objectives are to be achieved: 1. To design a suitable welding apparatus fixture, and optimize welding parameters for repeatable and acceptable joining of the fuel pin end-plug. The welding will be evaluated using tensile tests for lap joint weldments and helium leak tests for the fuel pin end-plug; 2 Investigate the microstructural and mechanical properties changes in PMW weldments of proposed advanced core and cladding alloys; 3. Simulate the irradiation effects on the PWM weldments using ion irradiation.

  19. Learning from corrosion. What are the demands of the operation, what is feasible weldingly at the corrosion protection by cladding?; Von Korrosion lernen. Welche Herausforderungen stellt der Betrieb, was ist schweisstechnisch beim Korrosionsschutz durch Cladding machbar?

    Energy Technology Data Exchange (ETDEWEB)

    Herzog, Thomas; Trotha, Ghita von; Molitor, Dominik [CheMin GmbH, Augsburg (Germany)

    2013-03-01

    The weld cladding of pipe walls and single tubes with nickel-base alloys is a precautionary measure for the new construction of evaporators or super-heaters in incinerators with difficult fuels. Under this aspect, the authors of this contribution report on some aspects of corrosion of claddings in order to learn about the corrosion protection. Welding optimizations are presented.

  20. Structural Analysis of Surface-Modified Oxidation-Resistant Zirconium Alloy Cladding for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho; No, Hee Cheon; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    While the current zirconium-based alloy cladding (Zircaloy, here after) has served well for fission-product barrier and heat transfer medium for the nuclear fuel of light water reactors (LWRs) in steady-states, concerns surrounding its mechanical behavior during accidents have drawn serious attentions. In accidents, strength degradation of the current-zirconium based alloy cladding manifests at temperature around ∼800 .deg. C, which results in fuel ballooning. Above 1000 .deg. C, zircaloy undergoes rapid oxidation with steam. Formation of brittle oxide (ZrO{sub 2}) and underlying oxygen-saturated α-zircaloy as a consequence of steam oxidation leads to loss of cladding ductility. Indeed, the loss of zircaloy ductility upon the steam oxidation has been taken as a measure of fuel failure criteria as stated in 10 CFR 50.46. In addition, zircaloy steam oxidation is an exothermic reaction, which is an energy source that sharply accelerates temperature increase under loss of coolant accidents, decreasing allowable coping time for loss of coolant accidents, decreasing allowable coping time for loss of coolant accidents (LOCA) before significant fuel/core melting starts. Hydrogen generated as a result of zircaloy oxidation could cause an explosion if certain conditions are met. In steady-state operation, zircaloy embrittlement limits the burnup of the fuel rod to assure remaining cladding ductility to cope with accidents. As a way to suppress hydrogen generation and cladding embrittlement by oxidation, ideas of cladding coating with an oxidation-preventive layer have emerged. Indeed, among a numbers of 'accident-tolerant-fuel (ATF)' concepts, the concept of coating the current fuel rod. Some signs of success on the lab-scale oxidation-prevention have been confirmed with a few coating candidates. Yet, relatively less attention has been given to structural integrity of coated zirconium-based alloy cladding. It is important to note that oxidation

  1. Hafnium isotope variations in oceanic basalts

    Science.gov (United States)

    Patchett, P. J.; Tatsumoto, M.

    1980-01-01

    Hafnium isotope ratios generated by the beta(-) decay of Lu-176 are investigated in volcanic rocks derived from the suboceanic mantle. Hf-176/Hf-177 and Lu/Hf ratios were determined to precisions of 0.01-0.04% and 0.5%, respectively, by routine, low-blank chemistry. The Hf-176/Hf-177 ratio is found to be positively correlated with the Nd-143/Nd-144 ratio and negatively correlated with the Sr-87/Sr-86 and Pb-206/Pb-204 ratios, and to increase southwards along the Iceland-Reykjanes ridge traverse. An approximate bulk earth Hf-176/Hf-177 ratio of 0.28295 is inferred from the bulk earth Nd-143/Nd-144 ratio, which requires a bulk earth Lu/Hf ratio of 0.25, similar to the Juvinas eucrite. Midocean ridge basalts are shown to account for 60% of the range of Hf isotope ratios, and it is suggested that Lu-Hf fractionation is decoupled from Sm-Nd and Rb-Sr fractionation in very trace-element-depleted source regions as a result of partial melting.

  2. Nuclear reactor fuel element having improved heat transfer

    Science.gov (United States)

    Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

    1982-03-03

    A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

  3. Multispectral pyrometry for surface temperature measurement of oxidized Zircaloy claddings

    Science.gov (United States)

    Bouvry, B.; Cheymol, G.; Ramiandrisoa, L.; Javaudin, B.; Gallou, C.; Maskrot, H.; Horny, N.; Duvaut, T.; Destouches, C.; Ferry, L.; Gonnier, C.

    2017-06-01

    Non-contact temperature measurement in a nuclear reactor is still a huge challenge because of the numerous constraints to consider, such as the high temperature, the steam atmosphere, and irradiation. A device is currently developed at CEA to study the nuclear fuel claddings behavior during a Loss-of-Coolant Accident. As a first step of development, we designed and tested an optical pyrometry procedure to measure the surface temperature of nuclear fuel claddings without any contact, under air, in the temperature range 700-850 °C. The temperature of Zircaloy-4 cladding samples was retrieved at various temperature levels. We used Multispectral Radiation Thermometry with the hypothesis of a constant emissivity profile in the spectral ranges 1-1.3 μm and 1.45-1.6 μm. To allow for comparisons, a reference temperature was provided by a thermocouple welded on the cladding surface. Because of thermal losses induced by the presence of the thermocouple, a heat transfer simulation was also performed to estimate the bias. We found a good agreement between the pyrometry measurement and the temperature reference, validating the constant emissivity profile hypothesis used in the MRT estimation. The expanded measurement uncertainty (k = 2) of the temperature obtained by the pyrometry method was ±4 °C, for temperatures between 700 and 850 °C. Emissivity values, between 0.86 and 0.91 were obtained.

  4. International symposium on fuel rod simulators: development and application

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W. (comp.)

    1981-05-01

    Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

  5. Clad Treatment in KARMA Code and Library

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong-yeup; Lee, Hae-chan; Woo, Hae-seuk [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of)

    2016-05-15

    Zirconium is the main components in clad materials. The subgroup parameters of zirconium were generated with effective cross section which obtained by using flux distribution in clad region. It decreases absorption reaction rate differences with reference MCNP results. Use of composite nuclide is acceptable to increase efficiency but should be limited to specific target composition. Therefore, the use of the composite nuclide of Zircaloy-2 should be limited when HANA clad material is used for clad. Either using explicit components or generating composite nuclide for HANA is suggested. This paper investigates the clad analysis model for KARMA whether current method is applicable to HANA clad material.

  6. The Development of Expansion Plug Wedge Test for Clad Tubing Structure Mechanical Property Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Jiang, Hao [ORNL

    2016-01-12

    To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at the Oak Ridge National Laboratory (ORNL) and is described fully in US Patent Application 20060070455, “Expanded plug method for developing circumferential mechanical properties of tubular materials.” This method is designed for testing fuel rod cladding ductility in a hot cell using an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are the simplicity of measuring the test component assembly in the hot cell and the direct measurement of the specimen’s strain. It was also found that cladding strength could be determined from the test results.

  7. Results of the Gallium-Clad Phase 3 and Phase 4 tasks (canceled prior to completion)

    Energy Technology Data Exchange (ETDEWEB)

    Morris, R.N.

    1998-08-01

    This report summarizes the results of the Gallium-Clad interactions Phase 3 and 4 tasks. Both tasks were to involve examining the out-of-pile stability of residual gallium in short fuel rods with an imposed thermal gradient. The thermal environment was to be created by an electrical heater in the center of the fuel rod and coolant flow on the rod outer cladding. Both tasks were canceled due to difficulties with fuel pellet fabrication, delays in the preparation of the test apparatus, and changes in the Fissile Materials Disposition program budget.

  8. Discovery of Gallium, Germanium, Lutetium, and Hafnium Isotopes

    CERN Document Server

    Gross, J L

    2011-01-01

    Currently, twenty-eight gallium, thirty-one germanium, thirty-five lutetium, and thirty-six hafnium isotopes have been observed and the discovery of these isotopes is discussed here. For each isotope a brief synopsis of the first refereed publication, including the production and identification method, is presented.

  9. Standard specification for nuclear-grade hafnium oxide powder

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification defines the physical and chemical requirements for hafnium oxide powder intended for fabrication into shapes for use in a nuclear reactor core. 1.2 The material described herein shall be particulate in nature. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

  10. X-Ray Photoemission Study of the Oxidation of Hafnium

    Directory of Open Access Journals (Sweden)

    A. R. Chourasia

    2009-01-01

    Full Text Available About 20 Å of hafnium were deposited on silicon substrates using the electron beam evaporation technique. Two types of samples were investigated. In one type, the substrate was kept at the ambient temperature. After the deposition, the substrate temperature was increased to 100, 200, and 300∘C. In the other type, the substrate temperature was held fixed at some value during the deposition. For this type, the substrate temperatures used were 100, 200, 300, 400, 500, 550, and 600∘C. The samples were characterized in situ by the technique of X-ray photoelectron spectroscopy. No trace of elemental hafnium is observed in the deposited overlayer. Also, there is no evidence of any chemical reactivity between the overlayer and the silicon substrate over the temperature range used. The hafnium overlayer shows a mixture of the dioxide and the suboxide. The ratio of the suboxide to dioxide is observed to be more in the first type of samples. The spectral data indicate that hafnium has a strong affinity for oxygen. The overlayer gets completely oxidized to form HfO2 at substrate temperature around 300∘C for the first type of samples and at substrate temperature greater than 550∘C for the second type.

  11. Corrosion and tribocorrosion of hafnium in simulated body fluids.

    Science.gov (United States)

    Rituerto Sin, J; Neville, A; Emami, N

    2014-08-01

    Hafnium is a passive metal with good biocompatibility and osteogenesis, however, little is known about its resistance to wear and corrosion in biological environments. The corrosion and tribocorrosion behavior of hafnium and commercially pure (CP) titanium in simulated body fluids were investigated using electrochemical techniques. Cyclic polarization scans and open circuit potential measurements were performed in 0.9% NaCl solution and 25% bovine calf serum solution to assess the effect of organic species on the corrosion behavior of the metal. A pin-on-plate configuration tribometer and a three electrode electrochemical cell were integrated to investigate the tribocorrosion performance of the studied materials. The results showed that hafnium has good corrosion resistance. The corrosion density currents measured in its passive state were lower than those measured in the case of CP titanium; however, it showed a higher tendency to suffer from localized corrosion, which was more acute when imperfections were present on the surface. The electrochemical breakdown of the oxide layer was retarded in the presence of proteins. Tribocorrosion tests showed that hafnium has the ability to quickly repassivate after the oxide layer was damaged; however, it showed higher volumetric loss than CP titanium in equivalent wear-corrosion conditions. © 2013 Wiley Periodicals, Inc. J Biomed Mater Res Part B: Appl Biomater, 102B: 1157-1164, 2014. © 2013 Wiley Periodicals, Inc.

  12. Hafnium - an optical hydrogen sensor spanning six orders in pressure

    NARCIS (Netherlands)

    Boelsma, C.; Bannenberg, L.J.; Van Setten, M. J.; Steinke, N.J.; van Well, A.A.; Dam, B.

    2017-01-01

    Hydrogen detection is essential for its implementation as an energy vector. So far, palladium is considered to be the most effective hydrogen sensing material. Here we show that palladium-capped hafnium thin films show a highly reproducible change in optical transmission in response to a hydrogen

  13. Elastic and thermodynamic properties of zirconium-and hafnium ...

    Indian Academy of Sciences (India)

    ... Live Streaming. Home; Journals; Bulletin of Materials Science; Volume 41; Issue 1. Elastic and thermodynamic properties of zirconium- and hafnium-doped Rh 3 V intermetallic compounds: potential aerospace material. M MANJULA M SUNDARESWARI E VISWANATHAN. Volume 41 Issue 1 February 2018 Article ID 19 ...

  14. Elastic and thermodynamic properties of zirconium- and hafnium ...

    Indian Academy of Sciences (India)

    2018-02-02

    Feb 2, 2018 ... https://doi.org/10.1007/s12034-017-1537-3. Elastic and thermodynamic properties of zirconium- and hafnium-doped Rh3V intermetallic compounds: potential aerospace material. M MANJULA, M SUNDARESWARI. ∗ and E VISWANATHAN. Department of Physics, Sathyabama University, Chennai 600119, ...

  15. CLADDING DEGRADATION COMPONENT IN WASTE FORM DEGRADATION MODEL IN TSPA-SR

    Energy Technology Data Exchange (ETDEWEB)

    E. Siegmann; R.P. Rechard

    2001-01-19

    The U.S. Department of Energy (DOE) has prepared a total system performance assessment for a site recommendation (TSPA-SR), if suitable, on Yucca Mountain for disposal of radioactive waste. Discussed here is the Cladding Degradation Component of the Waste Form Degradation Model (WF Model), of the TSPA-SR. The Cladding Degradation Component determines the degradation rate of the Zircaloy cladding on commercial spent nuclear fuel (CSNF) and, thereby, the CSNF matrix exposed and radioisotopes available for dissolution in any water present. Since the 1950s, most CSNF has been clad with less than 1 mm (usually between 600 and 900 {micro}m) of Zircaloy, a zirconium alloy. Zircaloy cladding is not a designed engineered barrier of the Yucca Mountain disposal system, but rather is an existing characteristic of the CSNF that is important to determining the release rate of radioisotopes once the waste package (WP) has breached. Although studies of cladding degradation from fluoride [F] began at Lawrence Livermore National Laboratory as early as 1984, cladding as a characteristic of the waste was not considered in TSPAs, conducted in the early 1990s. However, enough information on cladding performance has accumulated in the literature such that cladding was considered in 1993 when examining the performance of DOE spent nuclear (DSNF) and most recently in TSPA for the viability assessment (TSPA-VA). The Nuclear Regulatory Commission (NRC) currently uses cladding data as the basis for extending the period of wet storage, for licensing dry storage facilities, and for licensing shipping casks for CNSF.

  16. Effect of hafnium-incorporation on the microstructure and dielectric properties of cobalt ferrite ceramics

    Science.gov (United States)

    Wells, Stephen Josiah

    The effect of hafnium ion (Hf4+) incorporation in cobalt ferrite (CFO) was studied. Samples of Hf substituted CFO ceramic (CoFe 2-xHfxO4), were synthesized in the laboratory with hafnium concentrations ranging from x=0.000 to x=0.200. X-ray diffraction scans show that the Hafnium CFO crystalizes in the inverse spinel phase. Inclusion of hafnium causes lattice expansion, increasing the lattice parameter from 8.374 A for pure CoFe2O4 to 8.391 A for the highest concentration of hafnium tested (x=0.020). The dielectric properties of CFO are greatly enhanced by inclusion of hafnium. The enhancement is due to the distortion on the lattice from the larger Hf-ions substituting the smaller Fe-ions. Frequency variation of the dielectric properties is well modeled by the modified Debye function, which takes into account multiple ions contributing to relaxation.

  17. Synthesis of Hafnium-Based Ceramic Materials for Ultra-High Temperature Aerospace Applications

    Science.gov (United States)

    Johnson, Sylvia; Feldman, Jay

    2004-01-01

    This project involved the synthesis of hafnium (Hf)-based ceramic powders and Hf-based precursor solutions that were suitable for preparation of Hf-based ceramics. The Hf-based ceramic materials of interest in this project were hafnium carbide (with nominal composition HE) and hafnium dioxide (HfO2). The materials were prepared at Georgia Institute of Technology and then supplied to research collaborators Dr. Sylvia Johnson and Dr. Jay Feldman) at NASA Ames Research Center.

  18. Irradiation and lithium presence influence on the crystallographic nature of zirconia in the framework of PWR zircaloy 4 fuel cladding corrosion study; Influence de l'irradiation et de la presence du lithium sur la nature cristallographique de la zircone dans le cadre de l'etude de la corrosion du zircaloy 4 en milieu reacteur a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Gibert, C

    1999-07-01

    The-increasing deterioration of the initially protective zirconia layer is one of the hypotheses which can explain the impairment with time of PWR fuel cladding corrosion. This deterioration could be worsened by irradiation or lithium presence in the oxidizing medium. The aim of this thesis was to underline the influence of those two parameters on zirconia crystallographic nature. We first studied the impact of ionic irradiation on pure, powdery, monoclinic zirconia and oxidation formed zirconia, mainly with X-ray diffraction and Raman microscopy. The high or low energy particles used (Kr{sup n+-}, Ar{sup n+}) respectively favored electronic or atomic defaults production. The crystallographic analyses showed that these irradiation have a significant effect on zirconia by inducing nucleation or growth of tetragonal phase. The extent depends on sample nature and particles energy. In all cases, phase transformation is correlated with crystalline parameters, grain size and especially micro-stress changes. The results are consistent with those obtained with 1 to 5 cycles PWR claddings. Therefore, the corrosion acceleration observed in reactor can partly be explained by the stress fields appearance under irradiation, which is particularly detrimental to zirconia layer cohesion. Last, we have underlined that the presence of considerable amounts of lithium in the oxidizing medium ((> 700 ppm) induces the disappearance of the tetragonal zirconia located at the metal/oxide interface and the appearance of a porosity of the dense under layer, which looses its protectiveness. (author)

  19. Coupon Surveillance For Corrosion Monitoring In Nuclear Fuel Basin

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Hydrothermal chemistry, structures, and luminescence studies of alkali hafnium fluorides.

    Science.gov (United States)

    Underwood, Christopher C; McMillen, Colin D; Chen, Hongyu; Anker, Jeffery N; Kolis, Joseph W

    2013-01-07

    This paper describes the hydrothermal chemistry of alkali hafnium fluorides, including the synthesis and structural characterization of five new alkali hafnium fluorides. Two ternary alkali hafnium fluorides are described: Li(2)HfF(6) in space group P31m with a = 4.9748(7) Å and c = 4.6449(9) Å and Na(5)Hf(2)F(13) in space group C2/m with a = 11.627(2) Å, b = 5.5159(11) Å, and c = 8.4317(17) Å. Three new alkali hafnium oxyfluorides are also described: two fluoroelpasolites, K(3)HfOF(5) and (NH(4))(3)HfOF(5), in space group Fm3m with a = 8.9766(10) and 9.4144(11) Å, respectively, and K(2)Hf(3)OF(12) in space group R3m with a = 7.6486(11) Å and c = 28.802(6) Å. Infrared (IR) spectra were obtained for the title solids to confirm the structure solutions. Comparison of these materials was made based on their structures and synthesis conditions. The formation of these species in hydrothermal fluids appears to be dependent upon both the concentration of the alkali fluoride mineralizer solution and the reaction temperature. Both X-ray and visible fluorescence studies were conducted on compounds synthesized in this study and showed that fluorescence was affected by a variety of factors, such as alkali metal size, the presence/absence of oxygen in the compound, and the coordination environment of Hf(4+).

  1. Pentamethylcyclopentadienyl Zirconium and Hafnium Polyhydride Complexes : Synthesis, Structure, and Reactivity

    NARCIS (Netherlands)

    Visser, Cindy; Hende, Johannes R. van den; Meetsma, Auke; Hessen, Bart; Teuben, Jan H.

    2001-01-01

    The half-sandwich zirconium and hafnium N,N-dimethylaminopropyl complexes Cp*M[(CH2)3NMe2]Cl2 (Cp* = η5-C5Me5, M = Zr, 1; Hf, 2) and Cp*M[(CH2)3NMe2]2Cl (M = Zr, 3; Hf, 4) were synthesized by mono- or dialkylation of Cp*MCl3 with the corresponding alkyllithium and Grignard reagents. Hydrogenolysis

  2. Reactions of zirconium and hafnium fluoride hydrates with hydrogen peroxide

    Energy Technology Data Exchange (ETDEWEB)

    Gerasimova, S.O.; Polishchuk, S.A.; Avkhutskii, L.M.; Kalennik, V.M. (AN SSSR, Vladivostok. Inst. Khimii)

    1981-01-01

    Zirconium peroxofluoride of ZrO/sub 2/F/sub 2/x2H/sub 2/O composition is prepared by interaction of zirconium tetrafluoride trihydrate with hydrogen peroxide at pH 2-3. Hafnium peroxofluoride compound is not formed under similar conditions. It can be caused by their structural peculiarities for the compounds are not isostructural IR, PMR and NMR spectra for Zr peroxofluoride are presented.

  3. Frictional Behavior of Fe-based Cladding Candidates for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ho; Kim, Hyung-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Byun, Thak Sang [Oak Ridge National Lab., Oak Ridge (United States)

    2014-10-15

    After the recent nuclear disaster at Fukushima Daiichi reactors, there is a growing consensus on the development of new fuel systems (i.e., accident-tolerant fuel, ATF) that have high safety margins under design-basis accident (DBA) and beyond design-basis accident (BDBA). A common objective of various developing candidates is to archive the outstanding corrosion-resistance under severe accidents such as DBA and DBDA conditions for decreasing hydrogen production and increasing coping time to respond to severe accidents. ATF could be defined as new fuel/cladding system with enhanced accident tolerant to loss of active cooling in the core for a considerably longer time period under severe accidents while maintaining or improving the fuel performance during normal operations. This means that, in normal operating conditions, new fuel systems should be applicable to current operating PWRs for suppressing various degradation mechanisms of current fuel assembly without excessive design changes. When considering that one of the major degradation mechanisms of PWR fuel assemblies is a grid-to-rod fretting (GTRF), it is necessary to examine the tribological behavior of various ATF candidates at initial development stage. In this study, friction and reciprocating wear behavior of two kinds of Fe-based ATF candidates were examined with a reciprocating wear tests at room temperature (RT) air and water. The objective is to examine the compatibilities of these Fe-based alloys against current Zr-based alloy properties, which is used as major structural materials of PWR fuel assemblies. The reciprocating wear behaviors of Fe-based accident-tolerant fuel cladding candidates against current Zr-based alloy has been studied using a reciprocating sliding wear tester in room temperature air and water. Frictional behavior and wear depth were used for evaluating the applicability and compatibilities of Fe-based candidates without significant design changes of PWR fuel assemblies

  4. Hafnium trifluoromethanesulfonate (hafnium triflate) as a highly efficient catalyst for chemoselective thioacetalization and transthioacetalization of carbonyl compounds.

    Science.gov (United States)

    Wu, Yan-Chao; Zhu, Jieping

    2008-12-05

    A range of carbonyl compounds including aliphatic and aromatic aldehydes and ketones were converted to the corresponding thioacetals in high yields in the presence of a catalytic amount of hafnium trifluoromethanesulfonate (0.1 mol %, room temperature). The mild conditions tolerated various sensitive functional and protecting groups and were racemization-free when applied to alpha-aminoaldehydes. Transacetalization and chemoselective thioacetalization of aromatic aldehydes in the presence of aliphatic aldehydes and ketones were also documented.

  5. Zirconium and hafnium in the southeastern Atlantic States

    Science.gov (United States)

    Mertie, J.B.

    1958-01-01

    The principal source of zirconium and hafnium is zircon, though a minor source is baddeleyite, mined only in Brazil. Zircon is an accessory mineral in igneous, metamorphic, and sedimentary rocks, but rarely occurs in hardrock in minable quantities. The principal sources of zircon are therefore alluvial deposits, which are mined in many countries of five continents. The principal commercial deposits in the United States are in Florida, though others exist elsewhere in the southeastern Coastal Plain. The evidence indicates that conditions for the accumulation of workable deposits of heavy minerals were more favorable during the interglacial stages of the Pleistocene epoch than during Recent time. Therefore detrital ores of large volume and high tenor are more likely to be found in the terrace deposits than along the present beaches. Other concentrations of heavy minerals, however, are possible at favored sites close to the Fall Line where the Tuscaloosa formation rests upon the crystalline rocks of the Piedmont province. A score of heavy and semiheavy minerals occur in the detrital deposits of Florida, but the principal salable minerals are ilmenite, leucoxene, rutile, and zircon, though monazite and staurolite are saved at some mining plants. Commercial deposits of heavy minerals are generally required to have a tenor of 4 percent, though ores with a lower tenor can be mined at a profit if the content of monazite is notably high. The percentages of zircon in the concentrates ranges from 10 to 16 percent, and in eastern Florida from 13 to 15 percent. Thus the tenor in zircon of the ore-bearing sands ranges from 0.4 to 0.6 percent. The content of hafnium in zircon is immaterial for many uses, but for some purposes very high or very low tenors in hafnium are required. Alluvial zircon cannot be separated into such varieties, which, if needed, must be obtained from sources in bedrock. It thus becomes necessary to determine the Hf : Zr ratios in zircon from many kinds of

  6. Hafnium carbamates and ureates: new class of precursors for low-temperature growth of HfO2 thin films.

    Science.gov (United States)

    Pothiraja, Ramasamy; Milanov, Andrian P; Barreca, Davide; Gasparotto, Alberto; Becker, Hans-Werner; Winter, Manuela; Fischer, Roland A; Devi, Anjana

    2009-04-21

    Novel volatile compounds of hafnium, namely tetrakis-N,O-dialkylcarbamato hafnium(iv) [Hf((i)PrNC(O)O(i)Pr)(4)] () and tetrakis-N,N,N'-trialkylureato hafnium(iv) [Hf((i)PrNC(O)N-(Me)Et)(4)] (), have been synthesized through the simple insertion reaction of isopropyl isocyanate into hafnium isopropoxide and hafnium ethylmethylamide, respectively; based on the promising thermal properties, compound has been evaluated as a precursor for metalorganic chemical vapor deposition (MOCVD) of HfO(2) thin films, which resulted in the growth of stoichiometric and crystalline layers with a uniform morphology at temperature as low as 250 degrees C.

  7. Evaluation of the fabricability of advanced iron aluminide-clad austenitic stainless steel tubing

    Energy Technology Data Exchange (ETDEWEB)

    Mohn, W.R.; Topolski, M.J. [Babcock and Wilcox Co., Alliance, OH (United States). Research and Development Div.

    1993-07-01

    Researchers at Babcock & Wilcox Alliance Research Center have investigated methods to produce bimetallic tubing consisting of iron aluminide-clad austenitic stainless steel for practical use in fossil fueled energy equipment. In the course of this work, the compatibility of iron aluminide with four candidate austenitic stainless steel substrates was first evaluated using diffusion couples. Based on these results, a combination of iron aluminide and 304 stainless steel was selected for further development. Two composite billets of this combination were then prepared and extruded in separate trails at 2200F and 2000F. Both extrusions yielded 2-inch OD clad tubes, each approximately 18 feet long. Results of the evaluation show that the tube extruded at 2000F had a sound, integrally bonded clad layer throughout its entire length. However, the tube extruded at 2200F exhibited regions of disbonding between the clad layer and the substrate. In supplement to this work, an assessment of the technical and economic merits of iron aluminide-clad austenitic stainless steel components in power generation systems was conducted by B&W Fossil Power Division. Future activities should include an investigation of lower extrusion processing temperatures to optimize the fabrication of high quality iron-aluminide clad tubing.

  8. Rigorous modeling of cladding modes in photonic crystal fibers

    DEFF Research Database (Denmark)

    Rindorf, Lars Henning; Bang, Ole

    We study the cladding modes of a photonic crystal fiber (PCF) with a finite size cladding using a finite element method. The cladding consists of seven rings of air holes with bulk silica outside.......We study the cladding modes of a photonic crystal fiber (PCF) with a finite size cladding using a finite element method. The cladding consists of seven rings of air holes with bulk silica outside....

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

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, Anthony M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pardini, Allan F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Denslow, Kayte M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crawford, Susan L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Larche, Michael R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    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.

  10. Capture of Tritium Released from Cladding in the Zirconium Recycle Process

    Energy Technology Data Exchange (ETDEWEB)

    Bruffey, Stephanie H [ORNL; Spencer, Barry B [ORNL; DelCul, Guillermo Daniel [ORNL

    2016-08-31

    This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

  11. Capture of Tritium Released from Cladding in the Zirconium Recycle Process

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Barry B [ORNL; Bruffey, Stephanie H [ORNL; DelCul, Guillermo Daniel [ORNL; Walker, Trenton Baird [ORNL

    2016-08-31

    Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

  12. Cladding Alloys for Fluoride Salt Compatibility

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Walker, Larry R [ORNL; Santella, Michael L [ORNL; Holcomb, David Eugene [ORNL

    2011-06-01

    This report provides an overview of several candidate technologies for cladding nickel-based corrosion protection layers onto high-temperature structural alloys. The report also provides a brief overview of the welding and weld performance issues associated with joining nickel-clad nickel-based alloys. From the available techniques, two cladding technologies were selected for initial evaluation. The first technique is a line-of-sight method that would be useful for cladding large structures such as vessel interiors or large piping. The line-of-sight method is a laser-based surface cladding technique in which a high-purity nickel powder mixed into a polymer binder is first sprayed onto the surface, baked, and then rapidly melted using a high-power laser. The second technique is a vapor phase technique based on the nickel-carbonyl process that is suitable for cladding inaccessible surfaces such as the interior surfaces of heat exchangers. An initial evaluation for performed on the quality of nickel claddings processed using the two selected cladding techniques.

  13. Thermal Analysis of a TREAT Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Papadias, Dionissios [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, Arthur E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-07-09

    The objective of this study was to explore options as to reduce peak cladding temperatures despite an increase in peak fuel temperatures. A 3D thermal-hydraulic model for a single TREAT fuel assembly was benchmarked to reproduce results obtained with previous thermal models developed for a TREAT HEU fuel assembly. In exercising this model, and variants thereof depending on the scope of analysis, various options were explored to reduce the peak cladding temperatures.

  14. Slow DNA transport through nanopores in hafnium oxide membranes.

    Science.gov (United States)

    Larkin, Joseph; Henley, Robert; Bell, David C; Cohen-Karni, Tzahi; Rosenstein, Jacob K; Wanunu, Meni

    2013-11-26

    We present a study of double- and single-stranded DNA transport through nanopores fabricated in ultrathin (2-7 nm thick) freestanding hafnium oxide (HfO2) membranes. The high chemical stability of ultrathin HfO2 enables long-lived experiments with 50 000 DNA translocations with no detectable pore expansion. Mean DNA velocities are slower than velocities through comparable silicon nitride pores, providing evidence that HfO2 nanopores have favorable physicochemical interactions with nucleic acids that can be leveraged to slow down DNA in a nanopore.

  15. Properties of neutron-rich hafnium high-spin isomers

    CERN Multimedia

    Tungate, G; Walker, P M; Neyens, G; Billowes, J; Flanagan, K; Koester, U H; Litvinov, Y

    It is proposed to study highly-excited multi-quasiparticle isomers in neutron-rich hafnium (Z=72) isotopes. Long half-lives have already been measured for such isomers in the storage ring at GSI, ensuring their accessibility with ISOL production. The present proposal focuses on:\\\\ (i) an on-line experiment to measure isomer properties in $^{183}$Hf and $^{184}$Hf, and\\\\ (ii) an off-line molecular breakup test using REXTRAP, to provide Hf$^{+}$ beams for future laser spectroscopy and greater sensitivity for the future study of more neutron-rich isotopes.

  16. Hafnium-Based Contrast Agents for X-ray Computed Tomography.

    Science.gov (United States)

    Berger, Markus; Bauser, Marcus; Frenzel, Thomas; Hilger, Christoph Stephan; Jost, Gregor; Lauria, Silvia; Morgenstern, Bernd; Neis, Christian; Pietsch, Hubertus; Sülzle, Detlev; Hegetschweiler, Kaspar

    2017-05-15

    Heavy-metal-based contrast agents (CAs) offer enhanced X-ray absorption for X-ray computed tomography (CT) compared to the currently used iodinated CAs. We report the discovery of new lanthanide and hafnium azainositol complexes and their optimization with respect to high water solubility and stability. Our efforts culminated in the synthesis of BAY-576, an uncharged hafnium complex with 3:2 stoichiometry and broken complex symmetry. The superior properties of this asymmetrically substituted hafnium CA were demonstrated by a CT angiography study in rabbits that revealed excellent signal contrast enhancement.

  17. Graphite Furnace Atomic Absorption Spectrometric Determination of Bismuth(III) after Coprecipitation with Hafnium Hydroxide

    OpenAIRE

    Ueda, Joichi; Takagi, Midori

    1990-01-01

    A method for the coprecipitation of bismuth(III) with hafnium hydroxide followed by a graphite-furnace atomic absorption spectrometric determination is described. Hafnium hydroxide coprecipitates quantitatively 0.05–3 μg of bismuth(III) from 50–400 cm3 of sample solution at pH 5.8–11.2. The presence of 2.5–50 mg of hafnium in 25 cm3 does not affect the atomic absorbance of bismuth(III). The calibration curve is linear for 0.05–3 μg of bismuth(III) in 25 cm3 and passes through the origin. Inte...

  18. Gallium-cladding compatibility testing plan: Phase 3 -- Test plan for centrally heated surrogate rodlet test. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Morris, R.N.; Baldwin, C.A.; Wilson, D.F.

    1998-07-01

    The Fissile Materials Disposition Program (FMDP) is investigating the use of weapons grade plutonium in mixed oxide (MOX) fuel for light-water reactors (LWR). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons derived fuel may differ from the previous commercial fuels because of small amounts of gallium impurities. A concern presently exists that the gallium may migrate out of the fuel, react with and weaken the clad, and thereby promote loss of fuel pin integrity. Phases 1 and 2 of the gallium task are presently underway to investigate the types of reactions that occur between gallium and clad materials. This is a Level-2 document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-Oxide Fuel Irradiation Test Project Plan. This Plan summarizes the projected Phase 3 Gallium-Cladding compatibility heating test and the follow-on post test examination (PTE). This work will be performed using centrally-heated surrogate pellets, to avoid unnecessary complexities and costs associated with working with plutonium and an irradiation environment. Two sets of rodlets containing pellets prepared by two different methods will be heated. Both sets will have an initial bulk gallium content of approximately 10 ppm. The major emphasis of the PTE task will be to examine the material interactions, particularly indications of gallium transport from the pellets to the clad.

  19. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  20. Fuel System Compatibility Issues for Prometheus-1

    Energy Technology Data Exchange (ETDEWEB)

    DC Noe; KB Gibbard; MH Krohn

    2006-01-20

    Compatibility issues for the Prometheus-1 fuel system have been reviewed based upon the selection of UO{sub 2} as the reference fuel material. In particular, the potential for limiting effects due to fuel- or fission product-component (cladding, liner, spring, etc) chemical interactions and clad-liner interactions have been evaluated. For UO{sub 2}-based fuels, fuel-component interactions are not expected to significantly limit performance. However, based upon the selection of component materials, there is a potential for degradation due to fission products. In particular, a chemical liner may be necessary for niobium, tantalum, zirconium, or silicon carbide-based systems. Multiple choices exist for the configuration of a chemical liner within the cladding; there is no clear solution that eliminates all concerns over the mechanical performance of a clad/liner system. A series of tests to evaluate the performance of candidate materials in contact with real and simulated fission products is outlined.

  1. Zirconium and hafnium Salalen complexes in isospecific polymerisation of propylene.

    Science.gov (United States)

    Press, Konstantin; Venditto, Vincenzo; Goldberg, Israel; Kol, Moshe

    2013-07-07

    The activity of dibenzylzirconium and dibenzylhafnium Salalen complexes in polymerisation of propylene with MAO as a cocatalyst is described. Three Salalen ligand precursors combining a bulky alkyl group (1-adamantyl) on the imine-side phenol and electron withdrawing halo groups of different sizes on the amine-side phenol were explored. All metal complexes were obtained as single diastereomers. An X-ray crystallographic structure of a hafnium complex of an additional ligand carrying the combination of tert-butyl and chloro substituted phenolates, 4-Hf, revealed a fac-mer wrapping of the Salalen ligand around the metal centre. All complexes led to active catalysts in propylene polymerisation and to isotactic polypropylene of high regioregularity. The zirconium complexes led to polypropylene having molecular weights of Mw = 132,000-200,000 and isotacticities of [mmmm] = 65.7-75.0%. The hafnium complexes led to polypropylene of higher molecular weights of Mw = 375,000-520,000 and higher stereoregularities of [mmmm] = 80.6-89.3%, the highest isotacticity obtained with 3-Hf.

  2. TEC – Thin Environmental Cladding

    Directory of Open Access Journals (Sweden)

    Alan Tomasi

    2015-05-01

    Full Text Available Permasteelisa Group developed with Fiberline Composites a new curtain wall system (Thin Environmental Cladding or TEC, making use of pultruded GFRP (Glass Fiber Reinforced Polymer material instead of traditional aluminum. Main advantages using GFRP instead of aluminum are the increased thermal performance and the limited environmental impact. Selling point of the selected GFRP resin is the light transmission, which results in pultruded profiles that allow the visible light to pass through them, creating great aesthetical effects. However, GFRP components present also weaknesses, such as high acoustic transmittance (due to the reduced weight and anisotropy of the material, low stiffness if compared with aluminum (resulting in higher facade deflection and sensible fire behavior (as combustible material. This paper will describe the design of the TEC-facade, highlighting the functional role of glass within the facade concept with regards to its acoustic, structural, aesthetics and fire behavior.

  3. Metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, Nina

    This work concerns planar optical waveguide sensors for biosensing applications, with the focus on deep-probe sensing for micron-scale biological objects like bacteria and whole cells. In the last two decades planar metal-clad waveguides have been brieflyintroduced in the literature applied...... for various biosensing applications, however a thorough study of the sensor configurations has not been presented, but is the main subject of this thesis. Optical sensors are generally well suited for bio-sensing asthey show high sensitivity and give an immediate response for minute changes in the refractive...... index of a sample, due to the high sensitivity of optical bio-sensors detection of non-labeled biological objects can be performed. The majority of opticalsensors presented in the literature and commercially available optical sensors are based on evanescent wave sensing, however most of these sensors...

  4. Enrichment/isolation of phosphorylated peptides on hafnium oxide prior to mass spectrometric analysis.

    Science.gov (United States)

    Rivera, José G; Choi, Yong Seok; Vujcic, Stefan; Wood, Troy D; Colón, Luis A

    2009-01-01

    Hafnium oxide (hafnia) exhibits unique enrichment properties towards phosphorylated peptides that are complementary to those of titanium oxide (titania) and zirconium oxide (zirconia) for use with mass spectrometric analysis in the field of proteomics.

  5. Friction Surface Cladding of AA1050 on AA2024-T351; influence of clad layer thickness and tool rotation rate

    NARCIS (Netherlands)

    Liu, Shaojie; Bor, Teunis Cornelis; Geijselaers, Hubertus J.M.; Akkerman, Remko

    2015-01-01

    Friction Surfacing Cladding (FSC) is a recently developed solid state process to deposit thin metallic clad layers on a substrate. The process employs a rotating tool with a central opening to supply clad material and support the distribution and bonding of the clad material to the substrate. The

  6. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın; Kazimi, Mujid S.

    2013-10-01

    The study evaluates the possible use of graphite foam as the bonding material between U-Pu-Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U-15Pu-6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600-660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

  7. Reduction of Liquid Clad Formation Due to Solid State Diffusion in Clad Brazing Sheet

    Science.gov (United States)

    Benoit, Michael J.; Whitney, Mark A.; Wells, Mary A.; Winkler, Sooky

    2016-12-01

    Warm forming operations have shown promise in expanding automotive heat exchanger designs by increasing forming limits of clad brazing sheet. The impact of isothermal holds below the clad melting temperature on subsequent brazeability has not previously been studied in detail. The effect of these holds on brazeability, as measured by the clad thickness loss due to solid state diffusion of Si out of the clad layer prior to clad melting, was assessed through parallel DSC and optical microscopy measurements, as well as through the use of a previously developed model. EPMA measurements were also performed to support the other measures. Overall, the same trends were predicted by DSC, microscopy, and the theoretical model; however, the DSC predictions were unable to accurately predict remaining clad thickness prior to melting, even after correcting the data for clad-core interactions. Microscopy measurements showed very good agreement with the model predictions, although there were slight discrepancies at short hold times due to the inability of the model to account for clad loss during heating to the brazing temperature. Further microscopy measurements showed that when the heating rate is set below a critical value, there is a reduction in the clad thickness from the as-received condition.

  8. Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J., E-mail: brad.merrill@inl.gov; Bragg-Sitton, Shannon M., E-mail: shannon.bragg-sitton@inl.gov; Humrickhouse, Paul W., E-mail: paul.humrickhouse@inl.gov

    2017-04-15

    Highlights: • Accident tolerant fuels (ATF) systems are currently under development for LWRs. • Many performance analysis tools are specifically developed for UO{sub 2}–Zr alloy fuel. • Modifications were made to the MELCOR code for candidate ATF cladding. • Preliminary analysis results for SiC and FeCrAl cladding concepts are presented. - Abstract: A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO{sub 2}–Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This paper describes modifications

  9. Reflood experiments in rod bundles with flow blockages due to clad ballooning

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S.K.; Kim, J.; Kim, K.; Kim, B.J.; Park, J.K.; Youn, Y.J.; Choi, H.S.; Song, C.H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-07-15

    Clad ballooning and the resulting partial flow blockage are one of the major thermal-hydraulic concerns associated with the coolability of partially blocked cores during a loss-of-coolant accident (LOCA). Several in-pile tests have shown that fuel relocation causes a local power accumulation and a high thermal coupling between the clad and fuel debris in the ballooned regions. However, previous experiments in the 1980s did not take into account the fuel relocation phenomena and resulting local power increase in the ballooned regions. The present paper presents the results of systematic investigations on the coolability of rod bundles with flow blockages. The experiments were mainly performed in 5 x 5 rod bundles, 2 x 2 rod bundles and other test facilities. The experiments include a reflood heat transfer, single-phase convective heat transfer, flow redistributions phenomena, and droplet break-up behavior. The effects of the fuel relocation and resulting local power increase were investigated using a 5 x 5 rod bundle. The fuel relocation phenomena increase the peak cladding temperature.

  10. Precipitation behaviour and recrystallisation resistance in aluminum alloys with additions of hafnium, scandium and zirconium

    OpenAIRE

    Hallem, Håkon

    2005-01-01

    The overall objective of this work has been to develop aluminium alloys, which after hot and cold deformation are able to withstand high temperatures without recrystallising. This has been done by investigating aluminium alloys with various additions of hafnium, scandium and zirconium, with a main focus on Hf and to which extent it may partly substitute or replace Zr and/or Sc as a dispersoid forming elements in these alloys. What is the effect of hafnium, alone and in combination with Zr...

  11. Thin Films of Reduced Hafnium Oxide with Excess Carbon for High-Temperature Oxidation Protection

    Science.gov (United States)

    2010-02-01

    contamination; thus the higher oxygen content found by XPS is partly due to organic impurities (and, possibly, water ) that are mostly concentrated in the...International Service Award, 2007. 25 REFERENCES ’C. B. Bargeron, R. C. Benson, and A. N. Jette , "High-Temperature Diffusion of Oxygen in Oxidizing Hafnium...A. N. Jette , and T. E. Phillips, "Oxidation of Hafnium Carbide in the Temperature Range 1400 ° to 2060 °C," Journal of the American Ceramic Society

  12. On the stabilization of niobium(V) solutions by zirconium(IV) and hafnium(IV)

    DEFF Research Database (Denmark)

    Sørensen, E.; Bjerre, A.B.

    1992-01-01

    Niobium cannot be separated from zirconium or hafnium when these elements occur together in solution with common anions such as chloride and sulphate. This is ascribed to the co-polymerization of niobium(V) and the hydrolysed ionic species of zirconium(IV) and hafnium(IV) to form colloidal partic...... particles. In hydrochloric acid the particles are positively charged, whereas in sulphate solution the Zr- and Hf-sulphate complexes confer a negative charge. The two cases are considered separately....

  13. RIA simulation tests using driver tube for ATF cladding

    Energy Technology Data Exchange (ETDEWEB)

    Cinbiz, Mahmut N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, N. R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lowden, R. R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, K. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    Pellet-cladding mechanical interaction (PCMI) is a potential failure mechanism for accident-tolerant fuel (ATF) cladding candidates during a reactivity-initiated accident (RIA). This report summarizes Fiscal Year (FY) 2017 research activities that were undertaken to evaluate the PCMI-like hoop-strain-driven mechanical response of ATF cladding candidates. To achieve various RIA-like conditions, a modified-burst test (MBT) device was developed to produce different mechanical pulses. The calibration of the MBT instrument was accomplished by performing mechanical tests on unirradiated Generation-I iron-chromium-aluminum (FeCrAl) alloy samples. Shakedown tests were also conducted in both FY 2016 and FY 2017 using unirradiated hydrided ZIRLO™ tube samples. This milestone report focuses on testing of ATF materials, but the benchmark tests with hydrided ZIRLO™ tube samples are documented in a recent journal article.a For the calibration and benchmark tests, the hoop strain was monitored using strain gauges attached to the sample surface in the hoop direction. A novel digital image correlation (DIC) system composed of a single high-speed camera and an array of six mirrors was developed for the MBT instrument to better resolve the failure behavior of samples and to provide useful data for validation of high-fidelity modeling and simulation tools. The DIC system enable a 360° view of a sample’s outer surface. This feature was added to the instrument to determine the precise failure location on a sample’s surface for strain predictions. The DIC system was tested on several silicon carbide fiber/silicon carbide matrix (SiC/SiC) composite tube samples at various pressurization rates of the driver tube (which correspond to the strain rates for the samples). The hoop strains for various loading conditions were determined for the SiC/SiC composite tube samples. Future work is planned to enhance understanding of the failure behavior of the ATF cladding candidates of age

  14. Optimization of metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, N.; Horvath, R.; Pedersen, H.C.

    2005-01-01

    The present paper deals with the optimization of metal-clad waveguides for sensor applications to achieve high sensitivity for adlayer and refractive index measurements. By using the Fresnel reflection coefficients both the angular shift and the width of the resonances in the sensorgrams are taken...... into account. Our optimization shows that it is possible for metal-clad waveguides to achieve a sensitivity improvement of 600% compared to surface-plasmon-resonance sensors....

  15. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  16. Glass-clad single crystalline fiber lasers

    Science.gov (United States)

    Lai, C. C.; Hsu, K. Y.; Huang, C. W.; Jheng, D. Y.; Wang, S. C.; Lin, S. L.; Yang, M. H.; Lee, Y. W.; Huang, D. W.; Huang, S. L.

    2012-06-01

    Yttrium aluminium garnet (YAG) has been widely used as a solid-state laser host because of its superior optical, thermal, mechanical properties, as well as its plurality in hosting active ions with a wide range of ionic radii. Drawing YAG into single crystalline fiber has the potential to further scale up the attainable power level with high mode quality. The recent advancement on the codrawing laser-heated pedestal growth (CDLHPG) technique can produce glass-clad YAG crystalline fibers for laser applications. The drawing speed can reach 10 cm/min for mass production. The CDLHPG technique has shown advantages on transition-metal ion doped YAG and short-fluorescent-lifetime ion doped YAG host. Compared to silica fiber lasers, the crystalline core offers high emission cross section for transition metal ions because of the unique local matrix. The challenges on the development of glass-clad YAG fibers, including core crystallinity, diameter uniformity, dopant segregation, residual strain, post-growth thermal treatment, and the thermal expansion coefficient mismatch between the crystalline core and glass clad are discussed. Chromium, ytterbium, and neodymium ions doped YAG fiber lasers have been successfully achieved with high efficiency and low threshold power. Power scaling with a clad-pump/side-coupling scheme using single clad or double clad YAG fibers is also discussed.

  17. Cryogenic mechanical loss measurements of heat-treated hafnium dioxide

    Science.gov (United States)

    Abernathy, M. R.; Reid, S.; Chalkley, E.; Bassiri, R.; Martin, I. W.; Evans, K.; Fejer, M. M.; Gretarsson, A.; Harry, G. M.; Hough, J.; MacLaren, I.; Markosyan, A.; Murray, P.; Nawrodt, R.; Penn, S.; Route, R.; Rowan, S.; Seidel, P.

    2011-10-01

    Low mechanical loss, high index-of-refraction thin-film coating materials are of particular interest to the gravitational wave detection community, where reduced mirror coating thermal noise in gravitational wave detectors is desirable. Current studies are focused on understanding the loss of amorphous metal oxides such as SiO2, Ta2O5 and HfO2. Here, we report recent measurements of the temperature dependence of the mechanical loss of ion-beam sputtered hafnium dioxide (HfO2) coatings that have undergone heat treatment. The results indicate that, even when partially crystallized, these coatings have lower loss than amorphous Ta2O5 films below ~100 K and that their loss exhibits some features which are heat-treatment dependent in the temperature range of ~100-200 K, with higher heat treatment yielding lower mechanical loss. The potential for using silica doping of hafnia coatings to prevent crystallization is discussed.

  18. Amorphous Hafnium-Indium-Zinc Oxide Semiconductor Thin Film Transistors

    Directory of Open Access Journals (Sweden)

    Sheng-Po Chang

    2012-01-01

    Full Text Available We reported on the performance and electrical properties of co-sputtering-processed amorphous hafnium-indium-zinc oxide (α-HfIZO thin film transistors (TFTs. Co-sputtering-processed α-HfIZO thin films have shown an amorphous phase in nature. We could modulate the In, Hf, and Zn components by changing the co-sputtering power. Additionally, the chemical composition of α-HfIZO had a significant effect on reliability, hysteresis, field-effect mobility (μFE, carrier concentration, and subthreshold swing (S of the device. Our results indicated that we could successfully and easily fabricate α-HfIZO TFTs with excellent performance by the co-sputtering process. Co-sputtering-processed α-HfIZO TFTs were fabricated with an on/off current ratio of ~106, higher mobility, and a subthreshold slope as steep as 0.55 V/dec.

  19. Cyclopentadienyl complexes of hafnium and zirconium containing nitrate ligands

    Energy Technology Data Exchange (ETDEWEB)

    Minacheva, M.Kh.; Brajnina, Eh.M.; Klemenkova, Z.S.; Lokshin, B.V.; Nikolaeva, T.D.; Zhdanov, S.I.; Petrovskij, P.V. (AN SSSR, Moscow. Inst. Ehlementoorganicheskikh Soedinenij)

    1983-06-01

    New types of monocyclopentadienyl nitrate complexes of zirconium and hafnium CpHf(DBM)(NO/sub 3/)/sub 2/ and CpHfCl/sub 2/(NO/sub 3/)x4H/sub 2/O (DBM = dibenzoylmethane residue) are synthesized. CpMCl/sub 2/(NO/sub 3/) dichlorides are formed during the reaction of CpM(chel)/sub 2/Cl and HNO/sub 3/ as a result of the interaction of the extracted HCl with the CpM(chel)/sub 2/(NO/sub 3/) exchange product. A supposition is made about the non-ionic character of the metal-nitrate bonds and the bidentate character of the nitrate ligands in Cp/sub 2/M(NO/sub 3/)Cl on the base of studying the electric conductivity, IR- and Raman spectra.

  20. TEC – Thin Environmental Cladding

    Directory of Open Access Journals (Sweden)

    Alan Tomasi

    2014-06-01

    Full Text Available Corresponding author: Alan Tomasi, Group R&D Project Manager, Permasteelisa S.p.A., viale E. Mattei 21/23 | 31029 Vittorio Veneto, Treviso, Italy. Tel.: +39 0438 505207; E-mail: a.tomasi@permasteelisagroup.com; www.permasteelisagroup.com Permasteelisa Group developed with Fiberline Composites a new curtain wall system (Thin Environmental Cladding or TEC, making use of pultruded GFRP (Glass Fiber Reinforced Polymer material instead of traditional aluminum. Main advantages using GFRP instead of aluminum are the increased thermal performance and the limited environmental impact. Selling point of the selected GFRP resin is the light transmission, which results in pultruded profiles that allow the visible light to pass through them, creating great aesthetical effects. However, GFRP components present also weaknesses, such as high acoustic transmittance (due to the reduced weight and anisotropy of the material, low stiffness if compared with aluminum (resulting in higher facade deflection and sensible fire behavior (as combustible material. This paper will describe the design of the TEC-facade, highlighting the functional role of glass within the facade concept with regards to its acoustic, structural, aesthetics and fire behavior.

  1. Industry Application Emergency Core Cooling System Cladding Acceptance Criteria Early Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngblood, Robert W. [FPoliSolutions LLC, Murrysville, PA (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Frepoli, Cesare [FPoliSolutions LLC, Murrysville, PA (United States); Yurko, Joseph P. [FPoliSolutions LLC, Murrysville, PA (United States); Swindlehurst, Gregg [GS Nuclear Consulting, Charlotte, NC (United States); Zoino, Angelo [Univ. of Rome Tor Vergata (Italy)

    2015-09-01

    The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA)/emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burnup on cladding performance as well as to address other technical issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in April 2016. The impact of the final 50.46c rule on the industry may involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or re-analyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 4-6 years following the rule effective date. As motivated by the new rule, the need to use advanced cladding designs may be a result. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently, there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin. The proposed rule would apply to a light water reactor and to all cladding types.

  2. Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Office of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high

  3. Nuclear fuel element with axially aligned fuel pellets and fuel microspheres therein

    Science.gov (United States)

    Sease, J.D.; Harrington, F.E.

    1973-12-11

    Elongated single- and multi-region fuel elements are prepared by replacing within a cladding container a coarse fraction of fuel material which includes plutonium and uranium in the appropriate regions of the fuel element and then infiltrating with vibration a fine-sized fraction of uranium-containing microspheres throughout all interstices in the coarse material in a single loading. The fine, rigid material defines a thin annular layer between the coarse fraction and the cladding to reduce adverse mechanical and chemical interactions. (Official Gazette)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. SiC/SiC Cladding Materials Properties Handbook

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Mary A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Singh, Gyanender P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    When a new class of material is considered for a nuclear core structure, the in-pile performance is usually assessed based on multi-physics modeling in coordination with experiments. This report aims to provide data for the mechanical and physical properties and environmental resistance of silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites for use in modeling for their application as accidenttolerant fuel cladding for light water reactors (LWRs). The properties are specific for tube geometry, although many properties can be predicted from planar specimen data. This report presents various properties, including mechanical properties, thermal properties, chemical stability under normal and offnormal operation conditions, hermeticity, and irradiation resistance. Table S.1 summarizes those properties mainly for nuclear-grade SiC/SiC composites fabricated via chemical vapor infiltration (CVI). While most of the important properties are available, this work found that data for the in-pile hydrothermal corrosion resistance of SiC materials and for thermal properties of tube materials are lacking for evaluation of SiC-based cladding for LWR applications.

  6. Cladding and Structural Materials for Advanced Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Microstructure of laser cladded martensitic stainless steel

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2006-08-01

    Full Text Available -steel substrate. A 4.4 kW Rofin DY044 diode pumped Nd:YAG laser coupled to a Kuka KR60L30 articulated arm robot and Precitec YW50 welding head with 300 mm focal length was used. Powder cladding was performed with Praxair Fe211-1 (420), Praxair Fe211-5 (400... dilution, low heat input, less distortion, increased mechanical and corrosion properties excellent repeatability and control of process parameters. Solidification of laser cladded martensitic stainless steel is primarily austenitic. Microstructures...

  8. CLADS analysis deliverables. Volume II. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Holland, L.L.; Frede, W.G.; Schleuter, M.E.; Grant, S.E.; Glass, H.H.; Atkinson, K.C.

    1985-08-01

    The functional specification for the Laboratory Technical Information System is contained in Volumes I and II of the CLADS Analysis Deliverables. This specification is the result of applying Structured Analysis and Information Analysis to the Materials Evaluation Laboratory during the analysis phase of this project. Volume I includes 22 data flow diagrams (DFDs), a complete data dictionary containing data elements, data flows, and dialog definitions. Definitions also are included for 77 automated stores or files. These deliverables comprise the user's functional system specification and will be used as input to subsequent project phases, including software design. Volume II of the CLADS Analysis Deliverables covers Data Flow and Element Definitions.

  9. Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants

    Directory of Open Access Journals (Sweden)

    Rebak Raul B.

    2017-01-01

    Full Text Available After a tsunami caused plant black out at Fukushima, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2-zirconium alloy system. This accident tolerant fuel alternative should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General electric, Oak ridge national laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl cladding such as APMT or C26M. Extensive testing and evaluation is being conducted to determine the suitability of FeCrAl under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher neutron absorption and tritium release effects can be minimized by design changes. The implementation of FeCrAl cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.

  10. Development of breached pin performance analysis code SAFFRON (System of Analyzing Failed Fuel under Reactor Operation by Numerical method)

    Energy Technology Data Exchange (ETDEWEB)

    Ukai, Shigeharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1995-03-01

    On the assumption of fuel pin failure, the breached pin performance analysis code SAFFRON was developed to evaluate the fuel pin behavior in relation to the delayed neutron signal response during operational mode beyond the cladding failure. Following characteristic behavior in breached fuel pin is modeled in 3-dimensional finite element method : pellet swelling by fuel-sodium reaction, fuel temperature change, and resultant cladding breach extension and delayed neutron precursors release into coolant. Particularly, practical algorithm of numerical procedure in finite element method was originally developed in order to solve the 3-dimensional non-linear contact problem between the swollen pellet due to fuel-sodium reaction and breached cladding. (author).

  11. State of hydroxogroups in zirconium- and hafnium hydroxysulfate complexes

    Energy Technology Data Exchange (ETDEWEB)

    Bochkarev, Eh.P.; Kolpachkova, N.M.; Muravlev, Yu.B.; Nekhamkin, L.G.; Ovsyannikova, N.V.; Sokolova, E.L. (Gosudarstvennyj Nauchno-Issledovatel' skij i Proektnyj Inst. Redkometallicheskoj Promyshlennosti, Moscow (USSR))

    1984-01-01

    Hydroxogroup state in zirconium sesquisulfate and hafnium monosulfate is studied. On the basis of analyzing the data of PMR and IR-spectroscopy of hydrolized sulfate Zr and Hf compounds are found. Compounds studied have Zr=(OH)/sub 2/=Zr dimers or (Hf(OH)sub(2))sub(n) chains as elements of the structure in which atoms of binary oxygen bridge form hydroxyl groups with hydrogen atoms. Complexes include molecules of crystallization water toughly coordinated by Zr or Hf atoms. The splitting of absorption bands observed in IR-spectrum is connected with the fact that during dissociation of hydroxogroups of the Me-OH-Me grouping, new bridge groupings with Me-O-Me oxogroups and bidentate water molecules Me-OH/sub 2/-Me appear. It is established that in compounds which are considered not oxo but hydroxocompounds the phenomenon of protonolytic dissociation of a part of hydroxogroups takes place. Exchange reactions occurring at a high rate and dissociation of hydroxogroups are typical for most hydrolized Zr and Hf compounds. The presence of hydrolized Zr- and Hf sulfates and the composition relatively to movable protons and OH-groups presupposes the ability of compounds of this class to participation in cation- and anion-exchange reactions and proves amorphous properties of Zr and Hf hydroxocompounds.

  12. Reinvestigation of high pressure polymorphism in hafnium metal

    Energy Technology Data Exchange (ETDEWEB)

    Pandey, K. K., E-mail: kkpandey@barc.gov.in; Sharma, Surinder M. [High Pressure and Synchrotron Radiation Physics Division, Bhabha Atomic Research Centre, Mumbai-400 085 (India); Gyanchandani, Jyoti; Dey, G. K. [Materials Science Division, Bhabha Atomic Research Centre, Mumbai-400 085 (India); Somayazulu, M. [Geophysical Laboratory, Carnegie Institution of Washington, Washington, D.C. 20015 (United States); Sikka, S. K. [Indian National Science Academy, New Delhi-110 002 (India)

    2014-06-21

    There has been a recent controversy about the high pressure polymorphism of Hafnium (Hf). Unlike, the earlier known α→ω structural transition at 38 ± 8 GPa, at ambient temperature, Hrubiak et al. [J. Appl. Phys. 111, 112612 (2012)] did not observe it till 51 GPa. They observed this transition only at elevated temperatures. We have reinvestigated the room temperature phase diagram of Hf, employing x-ray diffraction (XRD) and DFT based first principles calculations. Experimental investigations have been carried out on several pure and impure Hf samples and also with different pressure transmitting media. Besides demonstrating the significant role of impurity levels on the high pressure phase diagram of Hf, our studies re-establish room temperature α→ω transition at high pressures, even in quasi-hydrostatic environment. We observed this transition in pure Hf with equilibrium transition pressure P{sub o} = 44.5 GPa; however, with large hysteresis. The structural sequence, transition pressures, the lattice parameters, the c/a ratio and its variation with compression for the α and ω phases as predicted by our ab-initio scalar relativistic (SR) calculations are found to be in good agreement with our experimental results of pure Hf.

  13. Oxidation Effect in Octahedral Hafnium Disulfide Thin Film.

    Science.gov (United States)

    Chae, Sang Hoon; Jin, Youngjo; Kim, Tae Soo; Chung, Dong Seob; Na, Hyunyeong; Nam, Honggi; Kim, Hyun; Perello, David J; Jeong, Hye Yun; Ly, Thuc Hue; Lee, Young Hee

    2016-01-26

    Atomically smooth van der Waals materials are structurally stable in a monolayer and a few layers but are susceptible to oxygen-rich environments. In particular, recently emerging materials such as black phosphorus and perovskite have revealed stronger environmental sensitivity than other two-dimensional layered materials, often obscuring the interesting intrinsic electronic and optical properties. Unleashing the true potential of these materials requires oxidation-free sample preparation that protects thin flakes from air exposure. Here, we fabricated few-layer hafnium disulfide (HfS2) field effect transistors (FETs) using an integrated vacuum cluster system and study their electronic properties and stability under ambient conditions. By performing all the device fabrication and characterization procedure under an oxygen- and moisture-free environment, we found that few-layer AA-stacking HfS2-FETs display excellent field effect responses (Ion/Ioff ≈ 10(7)) with reduced hysteresis compared to the FETs prepared under ambient conditions. Oxidation of HfS2 occurs uniformly over the entire area, increasing the film thickness by 250% at a prolonged oxidation time of >120 h, while defects on the surface are the preferential initial oxidation sites. We further demonstrated that the stability of the device in air is significantly improved by passivating FETs with BN in a vacuum cluster.

  14. Hafnium oxide nanoparticles: toward an in vitro predictive biological effect?

    Science.gov (United States)

    Marill, Julie; Anesary, Naeemunnisa Mohamed; Zhang, Ping; Vivet, Sonia; Borghi, Elsa; Levy, Laurent; Pottier, Agnes

    2014-06-30

    Hafnium oxide, NBTXR3 nanoparticles were designed for high dose energy deposition within cancer cells when exposed to ionizing radiation. The purpose of this study was to assess the possibility of predicting in vitro the biological effect of NBTXR3 nanoparticles when exposed to ionizing radiation. Cellular uptake of NBTXR3 nanoparticles was assessed in a panel of human cancer cell lines (radioresistant and radiosensitive) by transmission electron microscopy. The radioenhancement of NBTXR3 nanoparticles was measured by the clonogenic survival assay. NBTXR3 nanoparticles were taken up by cells in a concentration dependent manner, forming clusters in the cytoplasm. Differential nanoparticle uptake was observed between epithelial and mesenchymal or glioblastoma cell lines. The dose enhancement factor increased with increase NBTXR3 nanoparticle concentration and radiation dose. Beyond a minimum number of clusters per cell, the radioenhancement of NBTXR3 nanoparticles could be estimated from the radiation dose delivered and the radiosensitivity of the cancer cell lines. Our preliminary results suggest a predictable in vitro biological effect of NBTXR3 nanoparticles exposed to ionizing radiation.

  15. Crack resistance curve determination of zircaloy-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Bertsch, J.; Alam, A.; Zubler, R

    2009-03-15

    Fracture mechanics properties of fuel claddings are of relevance with respect to fuel rod integrity. The integrity of a fuel rod, in turn, is important for the fuel performance, for the safe handling of fuel rods, for the prevention of leakages and subsequent dissemination of fuel, for the avoidance of unnecessary dose rates, and for safe operation. Different factors can strongly deteriorate the mechanical fuel rod properties: irradiation damage, thermo-mechanical impact, corrosion or hydrogen uptake. To investigate the mechanical properties of fuel rod claddings which are used in Swiss nuclear power plants, PSI has initiated a program for mechanical testing. A major issue was the interaction between specific loading devices and the tested cladding tube, e.g. in the form of bending or friction. Particular for Zircaloy is the hexagonal closed packed structure of the zirconium crystallographic lattice. This structure implies plastic deformation mechanisms with specific, preferred orientations. Further, the manufacturing procedure of Zircaloy claddings induces a specific texture which plays a salient role with respect to the embrittlement by irradiation or integration of hydrogen in the form of hydrides. Both, the induced microstructure as well as the plastic deformation behaviour play a role for the mechanical properties. At PSI, in a first step inactive thin walled Zircaloy tubes and, for comparison reasons, plates were tested. The validity of the mechanical testing of the non standard tube and plate geometries had to be verified. The used Zircaloy-4 cladding tube sections and small plates of the same wall thickness have been notched, fatigue pre-cracked and tensile tested to evaluate the fracture toughness properties at room temperature, 300 {sup o}C and 350 {sup o}C. The crack propagation has been determined optically. The test results of the plates have been further used to validate FEM calculations. For each sample a complete crack resistance (J-R) curve could

  16. Cladding Effects on Structural Integrity of Nuclear Components

    Energy Technology Data Exchange (ETDEWEB)

    Sattari-Far, Iradi; Andersson, Magnus [lnspecta Technology AB, Stockholm (Sweden)

    2006-06-15

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the

  17. Thick tool steel coatings with laser cladding

    NARCIS (Netherlands)

    Ocelik, V.; de Oliveira, U.; De Hosson, J. Th. M.; DeHosson, JTM; Brebbia, CA; Nishida, SI

    2007-01-01

    This paper concentrates on thick and crack-free laser clad coatings (up to 3 mm). The coating material is a chromium-molybdenum-tungsten-vanadium alloyed high-speed steel that shows high wear resistance, high compressive strength, good toughness, very good dimensional stability on heat treatment and

  18. Study and Behaviour of Prefabricated Composite Cladding

    Science.gov (United States)

    Sai Avinash, P.; Thiagarajan, N.; Santhi, A. S.

    2017-07-01

    The incessant population rise entailed for an expeditious construction at competitive prices that steered the customary path to the light weight structural components. This lead to construction of structural components using ferrocement. The load bearing structural cladding, sizing 3200x900x100 mm, is chosen for the study, which, is analyzed using the software ABAQUS 6.14 in accordance with the IS:875-87 Part1, IS:875-87 Part2, ACI 549R-97, ACI 318R-08 and NZS:3101-06 Part1 standards. The Ferrocement claddings (FCs) are fabricated to a scaled dimension of 400x115x38 mm. The light weight-high strength phenomena are corroborated by incorporating Glass Fibre Reinforced Polymer Laminates (GFRPL) of thickness 6mm, engineered with the aid of hand layup (wet layup) technique wielding epoxy resin, followed by curing under room temperature. The epoxy resin is employed for fastening ferrocement cladding with the Glass fiber reinforced polymer laminate, with the contemporary methodology. The compressive load carrying capacity of the amalgamated assembly, both in presence and absence of Glass Fibre Reinforced polymer laminates (GFRPL) on either side of Ferrocement cladding, has been experimented.

  19. Functionally graded materials with laser cladding

    NARCIS (Netherlands)

    de Hosson, J.T.M.; Pei, Y.T.; Brebbia, CA

    2001-01-01

    Al-40 w/o Si functionally graded materials (FGMs) were produced by a onestep laser cladding process on cast Al-alloy substrate as a possible solution for interfacial problems often present in laser coatings. The microstructure of the FGMs consists of a large amount of silicon primary particles

  20. Plasmonic waveguides cladded by hyperbolic metamaterials

    DEFF Research Database (Denmark)

    Ishii, Satoshi; Shalaginov, Mikhail Y.; Babicheva, Viktoriia E.

    2014-01-01

    Strongly anisotropic media with hyperbolic dispersion can be used for claddings of plasmonic waveguides (PWs). In order to analyze the fundamental properties of such waveguides, we analytically study 1D waveguides arranged from a hyperbolic metamaterial (HMM) in a HMM-Insulator-HMM (HIH) structure...

  1. CLAD/BCLAD Examinations. Test Specifications.

    Science.gov (United States)

    California Commission on Teacher Credentialing, Sacramento.

    This draft handbook provides descriptions and summaries of the six tests that California elementary and secondary teachers must pass to earn the Crosscultural, Language, and Academic Development (CLAD) or Bilingual, Crosscultural, Language, and Academic Development (BCLAD) certificates. The six tests include: (1) "Language Structure and…

  2. Dimensional, microstructural and compositional stability of metal fuels

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, A.A.; Dayananda, M.A.

    1993-03-15

    The projects undertaken were to address two areas of concern for metal-fueled fast reactors: metallurgical compatibility of fuel and its fission products with the stainless steel cladding, and effects of porosity development in the fuel on fuel/cladding interactions and on sodium penetration in fuel. The following studies are reported on extensively in appendices: hot isostatic pressing of U-10Zr by coupled boundary diffusion/power law creep cavitation, liquid Na intrusion into porous U-10Zr fuel alloy by differential capillarity, interdiffusion between U-Zr fuel and selected Fe-Ni-Cr alloys, interdiffusion between U-Zr fuel vs selected cladding steels, and interdiffusion of Ce in Fe-base alloys with Ni or Cr.

  3. Modelling Accident Tolerant Fuel Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Gamble, Kyle Allan Lawrence [Idaho National Laboratory

    2016-05-01

    The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. The United States Department of Energy (DOE) through its Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is a three-year project to perform research on two accident tolerant concepts. The final outcome of the ATF HIP will be an in-depth report to the DOE Advanced Fuels Campaign (AFC) giving a recommendation on whether either of the two concepts should be included in their lead test assembly scheduled for placement into a commercial reactor in 2022. The two ATF concepts under investigation in the HIP are uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (Idaho National Laboratory, Los Alamos National Laboratory, and Argonne National Laboratory), a comprehensive multiscale approach to modeling is being used that includes atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. Model development and fuel performance analysis are critical since a full suite of experimental studies will not be complete before AFC must prioritize concepts for focused development. In this paper, we present simulations of the two proposed accident tolerance fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. Sensitivity analyses are completed using Sandia National Laboratories’ Dakota software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). We also outline the multiscale modelling approach being employed. Considerable additional work is required prior to preparing the recommendation report for the Advanced

  4. BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

    Energy Technology Data Exchange (ETDEWEB)

    Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sweet, Ryan T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling. As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations

  5. Evolution of transmission spectra of double cladding fiber during etching

    Science.gov (United States)

    Ivanov, Oleg V.; Tian, Fei; Du, Henry

    2017-11-01

    We investigate the evolution of optical transmission through a double cladding fiber-optic structure during etching. The structure is formed by a section of SM630 fiber with inner depressed cladding between standard SMF-28 fibers. Its transmission spectrum exhibits two resonance dips at wavelengths where two cladding modes have almost equal propagation constants. We measure transmission spectra with decreasing thickness of the cladding and show that the resonance dips shift to shorter wavelengths, while new dips of lower order modes appear from long wavelength side. We calculate propagation constants of cladding modes and resonance wavelengths, which we compare with the experiment.

  6. Research on laser cladding control system based on fuzzy PID

    Science.gov (United States)

    Zhang, Chuanwei; Yu, Zhengyang

    2017-12-01

    Laser cladding technology has a high demand for control system, and the domestic laser cladding control system mostly uses the traditional PID control algorithm. Therefore, the laser cladding control system has a lot of room for improvement. This feature is suitable for laser cladding technology, Based on fuzzy PID three closed-loop control system, and compared with the conventional PID; At the same time, the laser cladding experiment and friction and wear experiment were carried out under the premise of ensuring the reasonable control system. Experiments show that compared with the conventional PID algorithm in fuzzy the PID algorithm under the surface of the cladding layer is more smooth, the surface roughness increases, and the wear resistance of the cladding layer is also enhanced.

  7. Hafnium(IV) chloride complexes with chelating β-ketiminate ligands: Synthesis, spectroscopic characterization and volatility study.

    Science.gov (United States)

    Patil, Siddappa A; Medina, Phillip A; Antic, Aleks; Ziller, Joseph W; Vohs, Jason K; Fahlman, Bradley D

    2015-09-05

    The synthesis and characterization of four new β-ketiminate hafnium(IV) chloride complexes dichloro-bis[4-(phenylamido)pent-3-en-2-one]-hafnium (4a), dichloro-bis[4-(4-methylphenylamido)pent-3-en-2-one]-hafnium (4b), dichloro-bis[4-(4-methoxyphenylamido)pent-3-en-2-one]-hafnium (4c), and dichloro-bis[4-(4-chlorophenylamido)pent-3-en-2-one]-hafnium (4d) are reported. All the complexes (4a-d) were characterized by spectroscopic methods ((1)H NMR, (13)C NMR, IR), and elemental analysis while the compound 4c was further examined by single-crystal X-ray diffraction, revealing that the complex is monomer with the hafnium center in octahedral coordination environment and oxygens of the chelating N-O ligands are trans to each other and the chloride ligands are in a cis arrangement. Volatile trends are established for four new β-ketiminate hafnium(IV) chloride complexes (4a-d). Sublimation enthalpies (ΔHsub) were calculated from thermogravimetric analysis (TGA) data, which show that, the dependence of ΔHsub on the molecular weight (4a-c) and inductive effects from chlorine (4d). Copyright © 2015 Elsevier B.V. All rights reserved.

  8. Neutronic fuel element fabrication

    Science.gov (United States)

    Korton, George

    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

  9. Cytotoxicity and physicochemical properties of hafnium oxide nanoparticles.

    Science.gov (United States)

    Field, James A; Luna-Velasco, Antonia; Boitano, Scott A; Shadman, Farhang; Ratner, Buddy D; Barnes, Chris; Sierra-Alvarez, Reyes

    2011-09-01

    Nano-sized hafnium oxide (HfO(2)) particles are being considered for applications within the semiconductor industry. However, little is known about their cytotoxicity. The objective of this work was to assess several HfO(2) nanoparticles (NPs) samples for their acute cytotoxicity. Dynamic light scattering analysis of the samples indicated that the average particle size of the HfO(2) in aqueous dispersions was in the submicron range with a fraction of particles having nano-dimensions. The media used in the toxicity assays decreased or increased the average particle size of HfO(2) NPs due to dispersion or agglomeration. Static time-of-flight secondary ion mass spectrometry (ToF-SIMS) revealed numerous surface contaminants on the NPs. Only one HfO(2) sample caused moderate cytotoxicity to human cell lines. The inhibitory sample caused a 50% response in the Live/Dead assay with HaCaT skin cells at 2200 mg L(-1); and a 50% response in the mitochondrial toxicity test at 300 mg L(-1). A microbial inhibition assay based on methanogenic activity also revealed that another HFO(2) sample caused moderate inhibition. The difference in toxicity between samples could not be attributed to size. Instead the difference in toxicity was likely due to differences in the contaminants of the HfO(2). The ToF-SIMS analysis indicated unique signatures of Br and P in the sample toxic to human cell lines suggesting a distinct synthesis was used for that sample which may have been accompanied by inhibitory impurities. The results taken as a whole indicate that HfO(2) itself is relatively non-toxic. Copyright © 2011 Elsevier Ltd. All rights reserved.

  10. An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

    Science.gov (United States)

    Gamble, K. A.; Barani, T.; Pizzocri, D.; Hales, J. D.; Terrani, K. A.; Pastore, G.

    2017-08-01

    Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterion is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Further experiments are required to confirm these observations.

  11. ENTIRELY AQUEOUS SOLUTION-GEL ROUTE FOR THE PREPARATION OF ZIRCONIUM CARBIDE, HAFNIUM CARBIDE AND THEIR TERNARY CARBIDE POWDERS

    Directory of Open Access Journals (Sweden)

    Zhang Changrui

    2016-07-01

    Full Text Available An entirely aqueous solution-gel route has been developed for the synthesis of zirconium carbide, hafnium carbide and their ternary carbide powders. Zirconium oxychloride (ZrOCl₂.8H₂O, malic acid (MA and ethylene glycol (EG were dissolved in water to form the aqueous zirconium carbide precursor. Afterwards, this aqueous precursor was gelled and transformed into zirconium carbide at a relatively low temperature (1200 °C for achieving an intimate mixing of the intermediate products. Hafnium and the ternary carbide powders were also synthesized via the same aqueous route. All the zirconium, hafnium and ternary carbide powders exhibited a particle size of ∼100 nm.

  12. Laser-Based Characterization of Nuclear Fuel Plates

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. Measurement of L(III) Subshell Absorption Jump Parameters of Hafnium.

    Science.gov (United States)

    Cengiz, E; Saritas, N; Dogan, M; Koksal, O K; Karabulut, K; Apaydin, G; Tirasoglu, E

    2015-12-01

    The L(III) subshell absorption jump ratio and jump factor of hafnium have been measured using two different ways which are X-ray attenuation method and Energy Dispersive X-ray Fluorescence technique. The results obtained both ways have been compared with theoretical values. They are in good agreement with each other.

  14. Self-diffusion in the hexagonal structure of Zirconium and Hafnium: computer simulation studies

    Directory of Open Access Journals (Sweden)

    Diego Hernán Ruiz

    2005-12-01

    Full Text Available Self-diffusion by vacancy mechanism is studied in two metals of hexagonal close packed structure, namely Hafnium and Zirconium. Computer simulation techniques are used together with many-body potentials of the embedded atom type. Defect properties are calculated at 0 K by molecular static while molecular dynamic is used to explore a wide temperature range.

  15. Synthesis, properties, and structure of potassium titanyl phosphate single crystals doped with hafnium

    Energy Technology Data Exchange (ETDEWEB)

    Orlova, E. I.; Kharitonova, E. P. [Moscow State University, Faculty of Physics (Russian Federation); Novikova, N. E., E-mail: natnov@ns.crys.ras.ru; Verin, I. A.; Alekseeva, O. A.; Sorokina, N. I. [Russian Academy of Sciences, Shubnikov Institute of Crystallography (Russian Federation); Voronkova, V. I. [Moscow State University, Faculty of Physics (Russian Federation)

    2010-05-15

    Single crystals of potassium titanyl phosphate doped with hafnium are grown by spontaneous flux crystallization. Their physical properties are studied, and the structure of three KTi{sub 1-x}Hf{sub x}OPO{sub 4} crystals (x = 0.01, 0.03, and 0.12) is determined. In the crystals studied, hafnium mostly occupies the second titanium position. The doping of KTP crystals with hafnium results in an elongation of K-O bonds in the potassium polyhedra and, as a consequence, in a considerable (by approximately 180 deg. C) decrease in the temperature of ferroelectric phase transition. The magnitude of anomalous permittivity substantially decreases. The electrical conduction in the specimens studied decreases by approximately half an order of magnitude in the low-temperature region but remains almost unchanged in the high-temperature region. Even at minor concentrations, the presence of a hafnium additive in the specimens considerably (by 35%) enhances the intensity of the second harmonic generation of laser radiation.

  16. Modification of zirconium and hafnium alkoxides : the effect of molecular structure on derived materials

    NARCIS (Netherlands)

    Spijksma, G.I.

    2006-01-01

    This thesis deals with the influence of modifying ligands on the structure and stability of zirconium and hafnium precursors. The applicability of the obtained modified alkoxides has been evaluated for MOCVD and sol-gel. Furthermore, the influence of the introduction of heteroligands on the sol-gel

  17. Development of Cone Wedge Ring Expansion Test to Evaluate Mechanical Properties of Clad Tubing Structure

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-10-01

    To determine the hoop tensile properties of irradiated fuel cladding in a hot cell, a cone wedge ring expansion test method was developed. A four-piece wedge insert was designed with tapered angles matched to the cone shape of a loading piston. The ring specimen was expanded in the radial direction by the lateral expansion of the wedges under the downward movement of the piston. The advantages of the proposed method are that implementation of the test setup in a hot cell is simple and easy, and that it enables a direct strain measurement of the test specimen from the piston’s vertical displacement soon after the wedge-clad contact resistance is initiated.

  18. CLADS Analysis Deliverables. Volume I. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Holland, L.L.; Frede, W.G.; Schleuter, M.E.; Grant, S.E.; Glass, H.H.; Atkinson, K.C.

    1985-08-01

    The functional specification for the Laboratory Technical Information System is contained in Volumes I and II of the CLADS Analysis Deliverables. This specification is the result of applying Structured Analysis and Information Analysis to the Materials Evaluation Laboratory during the analysis phase of this project. Volume I includes 22 data flow diagrams (DFDs), a complete data dictionary containing data elements, data flows, and dialog definitions. Definitions also are included for 77 automated stores or files. These deliverables comprise the user's functional system specification and will be used as input to subsequent project phases, including software design. Volume I of the CLADS Analysis Deliverables contains Physical diagrams, Mini Specs, Automated Files, Manual Stores, and Dialog Definitions.

  19. Investigation of semiconductor clad optical waveguides

    Science.gov (United States)

    Batchman, T. E.; Carson, R. F.

    1985-01-01

    A variety of techniques have been proposed for fabricating integrated optical devices using semiconductors, lithium niobate, and glasses as waveguides and substrates. The use of glass waveguides and their interaction with thin semiconductor cladding layers was studied. Though the interactions of these multilayer waveguide structures have been analyzed here using glass, they may be applicable to other types of materials as well. The primary reason for using glass is that it provides a simple, inexpensive way to construct waveguides and devices.

  20. Calculation of the linear heat generation rates which violate the thermomechanical limit of plastic deformation of the fuel cladding in function of the burn up of a BWR fuel rod type; Calculo de las razones de generacion de calor lineal que violen el limite termomecanico de deformacion plastica de la camisa en funcion del quemado de una barra combustible tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Hernandez L, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2003-07-01

    The linear heat generation rates (LHGR) for a BWR type generic fuel rod, as function of the burnup that violate the thermomechanical limit of circumferential plastic deformation of the can (canning) in nominal operation in stationary state of the fuel rod are calculated. The evaluation of the LHGR in function of the burnt of the fuel, is carried out under the condition that the deformation values of the circumferential plastic deformation of the can exceeds in 0.1 the thermomechanical value operation limit of 1%. The results of the calculations are compared with the generation rates of linear operation heat in function of the burnt for this fuel rod type. The calculations are carried out with the FEMAXI-V and RODBURN codes. The results show that for exhibitions or burnt between 0 and 16,000 M Wd/tU a minimum margin of 160.8 W/cm exists among LHGR (439.6 W/cm) operation peak for the given fuel and maximum LHGR of the fuel (calculated) to reach 1.1% of circumferential plastic deformation of the can, for the peak factor of power of 1.40. For burnt of 20,000 MWd/tU and 60,000 MWd/tU exist a margin of 150.3 and 298.6 W/cm, respectively. (Author)

  1. French investigations of high burnup effect on LOCA thermomecanical behavior. Part two. Oxidation and quenching experiments under simulated LOCA conditions with high burnup clad material

    Energy Technology Data Exchange (ETDEWEB)

    GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)] [and others

    1997-01-01

    In the frame of the high burnup fuel studies to support a possible extension of the current discharge burnup limit, experimental programs have been undertaken, jointly by EDF and IPSN in order to study the thermal-shock behavior of high burnup fuel claddings under typical LOCA conditions. The TAGUS program used unirradiated cladding samples, bare or bearing a pre-corrosion state simulating the end-of-life state of high burnup fuel claddings: the TAGCIR program used actually irradiated cladding samples taken from high burnup rods irradiated over 5 cycles in a commercial EDF PWR and having reached a rod burnup close to 60 GWd/tU. The thermal-shock failure tests consisted in oxidizing the cladding samples under steam flow, on both inner and outer faces or on the outer face alone, and subjecting them to a final water quench. The heating was provided by an inductive furnace the power of which being regulated through monitoring of the sample surface temperature with use of a single-wave optical pyrometer. Analysis of the irradiated tests (TAGCIR series) evidenced an increased oxidation rate as compared to similar tests on unirradiated samples. Results of the quenching tests series on unirradiated and irradiated samples are plotted under the usual presentation of failure maps relative to the oxidation parameters ECR (equivalent cladding reacted) or e{sub {beta}} (thickness of the remaining beta phase layer) as a function of the oxidation temperature. Comparison of the failure limits for irradiated specimens to those for unirradiated specimens indicates a lower brittleness under two side oxidation and possibly the opposite under one-side oxidation. The tentative analysis of the oxidation and quenching tests results on irradiated samples reveals the important role played by the hydrogen charged during in-reactor corrosion on the oxidation kinetics and the failure bearing capability of the cladding under LOCA transient conditions.

  2. Research Progress on Laser Cladding Amorphous Coatings on Metallic Substrates

    Directory of Open Access Journals (Sweden)

    CHEN Ming-hui

    2017-01-01

    Full Text Available The microstructure and property of amorphous alloy as well as the limitations of the traditional manufacturing methods for the bulk amorphous alloy were briefly introduced in this paper.Combined with characteristics of the laser cladding technique,the research status of the laser cladding Fe-based,Zr-based,Ni-based,Cu-based and Al-based amorphous coatings on the metal substrates were mainly summarized.The effects of factors such as laser processing parameter,micro-alloying element type and content and reinforcing phase on the laser cladding amorphous coatings were also involved.Finally,the main problems and the future research directions of the composition design and control of the laser-cladded amorphous coating,the design and optimization of the laser cladding process,and the basic theory of the laser cladding amorphous coatings were also put forward finally.

  3. Propagation characteristics of a partially metal-clad optical guide: metal-clad optical strip line.

    Science.gov (United States)

    Yamamoto, Y; Kamiya, T; Yanai, H

    1975-02-01

    Experimental investigations of the partially metal-clad optical guide were performed. The optical guide consisted of a silicon substrate, an SiO(2) low index dielectric layer, an Al(2)O(3) high index dielectric layer, and a partial cladding layer of Al. The far-field patterns of the output beam from the second prism coupler were consistent with the calculated transverse field distributions using the effective index-of-refraction method. The experimental results on the coupling angles, confinement at the curved section, and attenuation constants are also reported.

  4. Enhanced Accident Tolerant LWR Fuels: Metrics Development

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton; Lori Braase; Rose Montgomery; Chris Stanek; Robert Montgomery; Lance Snead; Larry Ott; Mike Billone

    2013-09-01

    The Department of Energy (DOE) Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) is conducting research and development on enhanced Accident Tolerant Fuels (ATF) for light water reactors (LWRs). This mission emphasizes the development of novel fuel and cladding concepts to replace the current zirconium alloy-uranium dioxide (UO2) fuel system. The overall mission of the ATF research is to develop advanced fuels/cladding with improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation. The initial effort will focus on implementation in operating reactors or reactors with design certifications. To initiate the development of quantitative metrics for ATR, a LWR Enhanced Accident Tolerant Fuels Metrics Development Workshop was held in October 2012 in Germantown, MD. This paper summarizes the outcome of that workshop and the current status of metrics development for LWR ATF.

  5. Improving Rail Wear and RCF Performance using Laser Cladding

    OpenAIRE

    Lewis, S.R.; Fretwell-Smith, S.; Goodwin, P.S.; Smith, L.; Lewis, R.; Aslam, M.; Fletcher, D.I.; Murray, K.; Lambert, R.

    2016-01-01

    Laser cladding has been considered as a method for improving the wear and RCF performance of standard grade rail. This paper presents results of small scale tests carried out to assess the wear and RCF performance of rail which had been laser clad. Using the laser cladding process premium metals can be deposited on to the working surface of standard rail with the aim of enhancing the wear and RCF life of the rail. Various laser clad samples were tested using a twin-disc method. The candidate ...

  6. Comparison of ring compression testing to three point bend testing for unirradiated ZIRLO cladding

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2015-04-01

    Safe shipment and storage of nuclear reactor discharged fuel requires an understanding of how the fuel may perform under the various conditions that can be encountered. One specific focus of concern is performance during a shipment drop accident. Tests at Savannah River National Laboratory (SRNL) are being performed to characterize the properties of fuel clad relative to a mechanical accident condition such as a container drop. Unirradiated ZIRLO tubing samples have been charged with a range of hydride levels to simulate actual fuel rod levels. Samples of the hydrogen charged tubes were exposed to a radial hydride growth treatment (RHGT) consisting of heating to 400°C, applying initial hoop stresses of 90 to 170 MPa with controlled cooling and producing hydride precipitates. Initial samples have been tested using both a) ring compression test (RCT) which is shown to be sensitive to radial hydride and b) three-point bend tests which are less sensitive to radial hydride effects. Hydrides are generated in Zirconium based fuel cladding as a result of coolant (water) oxidation of the clad, hydrogen release, and a portion of the released (nascent) hydrogen absorbed into the clad and eventually exceeding the hydrogen solubility limit. The orientation of the hydrides relative to the subsequent normal and accident strains has a significant impact on the failure susceptability. In this study the impacts of stress, temperature and hydrogen levels are evaluated in reference to the propensity for hydride reorientation from the circumferential to the radial orientation. In addition the effects of radial hydrides on the Quasi Ductile Brittle Transition Temperature (DBTT) were measured. The results suggest that a) the severity of the radial hydride impact is related to the hydrogen level-peak temperature combination (for example at a peak drying temperature of 400°C; 800 PPM hydrogen has less of an impact/ less radial hydride fraction than 200 PPM hydrogen for the same thermal

  7. Immobilization mechanisms of deoxyribonucleic acid (DNA) to hafnium dioxide (HfO2) surfaces for biosensing applications.

    Science.gov (United States)

    Fahrenkopf, Nicholas M; Rice, P Zachary; Bergkvist, Magnus; Deskins, N Aaron; Cady, Nathaniel C

    2012-10-24

    Immobilization of biomolecular probes to the sensing substrate is a critical step for biosensor fabrication. In this work we investigated the phosphate-dependent, oriented immobilization of DNA to hafnium dioxide surfaces for biosensing applications. Phosphate-dependent immobilization was confirmed on a wide range of hafnium oxide surfaces; however, a second interaction mode was observed on monoclinic hafnium dioxide. On the basis of previous materials studies on these films, DNA immobilization studies, and density functional theory (DFT) modeling, we propose that this secondary interaction is between the exposed nucleobases of single stranded DNA and the surface. The lattice spacing of monoclinic hafnium dioxide matches the base-to-base pitch of DNA. Monoclinic hafnium dioxide is advantageous for nanoelectronic applications, yet because of this secondary DNA immobilization mechanism, it could impede DNA hybridization or cause nonspecific surface intereactions. Nonetheless, DNA immobilization on polycrystalline and amorphous hafnium dioxide is predominately mediated by the terminal phosphate in an oriented manner which is desirable for biosensing applications.

  8. Thermal behavior analysis of PWR fuel during RIA at various fuel burnups using modified theatre code

    Directory of Open Access Journals (Sweden)

    Nawaz Amjad

    2016-01-01

    Full Text Available The fuel irradiation and burnup causes geometrical and dimensional changes in the fuel rod which affects its thermal resistance and ultimately affects the fuel rod behavior during steady-state and transient conditions. The consistent analysis of fuel rod thermal performance is essential for precise evaluation of reactor safety in operational transients and accidents. In this work, analysis of PWR fuel rod thermal performance is carried out under steady-state and transient conditions at different fuel burnups. The analysis is performed by using thermal hydraulic code, THEATRe. The code is modified by adding burnup dependent fuel rod behavior models. The original code uses as-fabricated fuel rod dimensions during steady-state and transient conditions which can be modified to perform more consistent reactor safety analysis. AP1000 reactor is considered as a reference reactor for this analysis. The effect of burnup on steady-state fuel rod parameters has been investigated. For transient analysis, hypothetical reactivity initiated accident was simulated by considering a triangular power pulse of variable pulse height (relative to the full power reactor operating conditions and pulse width at different fuel burnups which corresponds to fresh fuel, low and medium burnup fuels. The effect of power pulse height, pulse width and fuel burnup on fuel rod temperatures has been investigated. The results of reactivity initiated accident analysis show that the fuel failure mechanisms are different for fresh fuel and fuel at different burnup levels. The fuel failure in fresh fuel is expected due to fuel melting as fuel temperature increases with increase in pulse energy (pulse height. However, at relatively higher burnups, the fuel failure is expected due to cladding failure caused by strong pellet clad mechanical interaction, where, the contact pressure increases beyond the cladding yield strength.

  9. Rheological evaluation of pretreated cladding removal waste

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, D.; Chan, M.K.C.; Lokken, R.O.

    1986-01-01

    Cladding removal waste (CRW) contains concentrations of transuranic (TRU) elements in the 80 to 350 nCi/g range. This waste will require pretreatment before it can be disposed of as glass or grout at Hanford. The CRW will be pretreated with a rare earth strike and solids removal by centrifugation to segregate the TRU fraction from the non-TRU fraction of the waste. The centrifuge centrate will be neutralized with sodium hydroxide. This neutralized cladding removal waste (NCRW) is expected to be suitable for grouting. The TRU solids removed by centrifugation will be vitrified. The goal of the Rheological Evaluation of Pretreated Cladding Removal Waste Program was to evaluate those rheological and transport properties critical to assuring successful handling of the NCRW and TRU solids streams and to demonstrate transfers in a semi-prototypic pumping environment. This goal was achieved by a combination of laboratory and pilot-scale evaluations. The results obtained during these evaluations were correlated with classical rheological models and scaled-up to predict the performance that is likely to occur in the full-scale system. The Program used simulated NCRW and TRU solid slurries. Rockwell Hanford Operations (Rockwell) provided 150 gallons of simulated CRW and 5 gallons of simulated TRU solid slurry. The simulated CRW was neutralized by Pacific Northwest Laboratory (PNL). The physical and rheological properties of the NCRW and TRU solid slurries were evaluated in the laboratory. The properties displayed by NCRW allowed it to be classified as a pseudoplastic or yield-pseudoplastic non-Newtonian fluid. The TRU solids slurry contained very few solids. This slurry exhibited the properties associated with a pseudoplastic non-Newtonian fluid.

  10. Review of CTF s Fuel Rod Modeling Using FRAPCON-4.0 s Centerline Temperature Predictions

    Energy Technology Data Exchange (ETDEWEB)

    Toptan, Aysenur [North Carolina State University (NCSU), Raleigh; Salko, Robert K [ORNL; Avramova, Maria [North Carolina State University (NCSU), Raleigh

    2017-01-01

    Coolant Boiling in Rod Arrays Two Fluid (COBRA-TF), or CTF1 [1], is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF s fuel rod modeling is originally developed for VIPRE code [2]. Its methodology is based on GAPCON [3] and FRAP [4] fuel performance codes, and material properties are included from MATPRO handbook [5]. This work focuses on review of CTF s fuel rod modeling to address shortcomings in CTF s temperature predictions. CTF is compared to FRAPCON which is U.S. NRC s steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes in fuel rod variables and temperatures including the eects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite dierence or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 [6] is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF s dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF s models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF s dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF s dynamic gap conductance model. Improving the CTF s dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.

  11. Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

    Energy Technology Data Exchange (ETDEWEB)

    Ernst, Frank [Case Western Reserve Univ., Cleveland, OH (United States)

    2016-10-13

    We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, and fatigue life), corrosion resistance, resistance to stress–corrosion cracking (hydrogen-induced embrittlement), and – potentially – radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non–treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.

  12. History of Resistance Welding Oxide Dispersion Strengthened Cladding and other High Temperature Materials at Center for Advanced Energy Studies

    Energy Technology Data Exchange (ETDEWEB)

    Larry Zirker; Nathan Jerred; Dr. Indrajit Charit; James Cole

    2012-03-01

    Research proposal 08-1079, 'A Comparative Study of Welded ODS Cladding Materials for AFCI/GNEP,' was funded in 2008 under an Advanced Fuel Cycle Initiative (AFCI) Research and Development Funding Opportunity, number DE-PS07-08ID14906. Th proposal sought to conduct research on joining oxide dispersion strengthen (ODS) tubing material to a solid end plug. This document summarizes the scientific and technical progress achieved during the project, which ran from 2008 to 2011.

  13. Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

    2009-08-01

    A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

  14. FPIN2 analysis of metal fueled pins

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, T H

    1985-09-01

    The FPIN2 code is used to calculate the thermal and mechanical behavior of a single fast reactor fuel pin experiencing a wide range of transients. This report describes the modifications and additions made to the FPIN2 code in order to handle metal fueled pins. The changes discussed in this status report have initially focused upon EBR-II driver fuel in order to calculate the temperature histories of EBR-II fuel and to understand the detailed behavior of metal fuel during recent TREAT experiments. Equations describing material behavior of metal fuel including thermal properties, elastic properties, secondary creep and fission gas swelling are presented and were easily incorporated into the FPIN2 code framework. The addition of a fission gas plenum model and a model to close the fuel-clad radial gap at 100% fuel melting required substantial modification to the code. Additional changes were made to handle heat transfer with a sodium bond, eutectic release of cladding axial restraint and eutectic penetration of the cladding. Three example calculations are presented including a study of the effect of fuel creep on axial elongation and temperature contour maps for EBR-II driver fuel. Some results from an FPIN2 analysis of TREAT experiment M2 are discussed.

  15. Switching Kinetics in Nanoscale Hafnium Oxide Based Ferroelectric Field-Effect Transistors.

    Science.gov (United States)

    Mulaosmanovic, Halid; Ocker, Johannes; Müller, Stefan; Schroeder, Uwe; Müller, Johannes; Polakowski, Patrick; Flachowsky, Stefan; van Bentum, Ralf; Mikolajick, Thomas; Slesazeck, Stefan

    2017-02-01

    The recent discovery of ferroelectricity in thin hafnium oxide films has led to a resurgence of interest in ferroelectric memory devices. Although both experimental and theoretical studies on this new ferroelectric system have been undertaken, much remains to be unveiled regarding its domain landscape and switching kinetics. Here we demonstrate that the switching of single domains can be directly observed in ultrascaled ferroelectric field effect transistors. Using models of ferroelectric domain nucleation we explain the time, field and temperature dependence of polarization reversal. A simple stochastic model is proposed as well, relating nucleation processes to the observed statistical switching behavior. Our results suggest novel opportunities for hafnium oxide based ferroelectrics in nonvolatile memory devices.

  16. Effect of hafnium and titanium coated implants on several blood biochemical markers after osteosynthesis in rabbits.

    Science.gov (United States)

    Yousef, Ashraf; Akhtyamov, Ildar; Shakirova, Faina; Zubairova, Lyaili; Gatina, Elmira; Aliev, Capital Ie Cyrilliclchin

    2014-01-01

    An experimental study comparing the dynamics of several biochemical markers before and after osteosynthesis, utilizing implants coated with titanium and hafnium nitrides and non-coated implants on rabbits' bones. The Study has been conducted on 30 rabbits of both sexes, at the age of 6-7 months, weighing 2526.5±74.4 gm. Animals underwent open osteotomy of the tibia in the middle third of the diaphysis followed by the intramedullary nailing. The level of alkaline phosphatase, calcium, phosphorus, total protein, glucose, ALT and AST were monitored for 60 days. the use of implants coated with titanium and hafnium nitrides, which have high strength, thermal and chemical stability, was not accompanied by the development of additional negative reactive changes compared to non-coated implants. Nanotechnology used in manufacturing bioinert coatings for implants for osteosynthesis, has made the post-operative period less complicated as reflected by less expressed changing in the markers of bone metabolism and hepatotoxicity.

  17. Synthesis and characterization of hafnium oxide films for thermo and photoluminescence applications.

    Science.gov (United States)

    Mendoza, J Guzmán; Frutis, M A Aguilar; Flores, G Alarcón; Hipólito, M García; Maciel Cerda, A; Azorín Nieto, J; Montalvo, T Rivera; Falcony, C

    2010-01-01

    Hafnium oxide (HfO(2)) films were deposited by the ultrasonic spray pyrolysis process. The films were synthesized from hafnium chloride as raw material in deionized water as solvent and were deposited on corning glass substrates at temperatures from 300 to 600 degrees C. For substrate temperatures lower than 400 degrees C the deposited films were amorphous, while for substrate temperatures higher than 450 degrees C, the monoclinic phase of HfO(2) appeared. Scanning electron microscopy showed that the film's surface resulted rough with semi-spherical promontories. The films showed a chemical composition close to HfO(2), with an Hf/O ratio of about 0.5. UV radiation was used in order to achieve the thermoluminescent characterization of the films; the 240 nm wavelength induced the best response. In addition, preliminary photoluminescence spectra, as a function of the deposition temperatures, are shown. Copyright 2009 Elsevier Ltd. All rights reserved.

  18. Hafnium-an optical hydrogen sensor spanning six orders in pressure.

    Science.gov (United States)

    Boelsma, C; Bannenberg, L J; van Setten, M J; Steinke, N-J; van Well, A A; Dam, B

    2017-06-05

    Hydrogen detection is essential for its implementation as an energy vector. So far, palladium is considered to be the most effective hydrogen sensing material. Here we show that palladium-capped hafnium thin films show a highly reproducible change in optical transmission in response to a hydrogen exposure ranging over six orders of magnitude in pressure. The optical signal is hysteresis-free within this range, which includes a transition between two structural phases. A temperature change results in a uniform shift of the optical signal. This, to our knowledge unique, feature facilitates the sensor calibration and suggests a constant hydrogenation enthalpy. In addition, it suggests an anomalously steep increase of the entropy with the hydrogen/metal ratio that cannot be explained on the basis of a classical solid solution model. The optical behaviour as a function of its hydrogen content makes hafnium well-suited for use as a hydrogen detection material.

  19. Highly flexible resistive switching memory based on amorphous-nanocrystalline hafnium oxide films.

    Science.gov (United States)

    Shang, Jie; Xue, Wuhong; Ji, Zhenghui; Liu, Gang; Niu, Xuhong; Yi, Xiaohui; Pan, Liang; Zhan, Qingfeng; Xu, Xiao-Hong; Li, Run-Wei

    2017-06-01

    Flexible and transparent resistive switching memories are highly desired for the construction of portable and even wearable electronics. Upon optimization of the microstructure wherein an amorphous-nanocrystalline hafnium oxide thin film is fabricated, an all-oxide based transparent RRAM device with stable resistive switching behavior that can withstand a mechanical tensile stress of up to 2.12% is obtained. It is demonstrated that the superior electrical, thermal and mechanical performance of the ITO/HfOx/ITO device can be ascribed to the formation of pseudo-straight metallic hafnium conductive filaments in the switching layer, and is only limited by the choice of electrode materials. When the ITO bottom electrode is replaced with platinum metal, the mechanical failure threshold of the device can be further extended.

  20. Metal clad aramid fibers for aerospace wire and cable

    Science.gov (United States)

    Tokarsky, Edward W.; Dunham, Michael G.; Hunt, James E.; Santoleri, E. David; Allen, David B.

    1995-01-01

    High strength light weight metal clad aramid fibers can provide significant weight savings when used to replace conventional metal wire in aerospace cable. An overview of metal clad aramid fiber materials and information on performance and use in braided electrical shielding and signal conductors is provided.

  1. Cladding of Advanced Al Alloys Employing Friction Stir Welding

    NARCIS (Netherlands)

    van der Stelt, A.A.; Bor, Teunis Cornelis; Geijselaers, Hubertus J.M.; Akkerman, Remko; van den Boogaard, Antonius H.

    2013-01-01

    In this paper an advanced solid state cladding process, based on Friction Stir Welding, is presented. The Friction Surface Cladding (FSC) technology enables the deposition of a solid-state coating using filler material on a substrate with good metallurgical bonding. A relatively soft AA1050 filler

  2. Long-range plasmonic waveguides with hyperbolic cladding

    DEFF Research Database (Denmark)

    Babicheva, Viktoriia E.; Shalaginov, Mikhail Y.; Ishii, Satoshi

    2015-01-01

    waveguides. We show that the proposed structures support long-range surface plasmon modes, which exist when the permittivity of the core matches the transverse effective permittivity component of the metamaterial cladding. In this regime, the surface plasmon polaritons of each cladding layer are strongly...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-10-01

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

  5. International Summer School on Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures.

  6. OECD - HRP Summer School on Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures.

  7. On-line separation of refractory hafnium and tantalum isotopes at the ISOCELE separator

    CERN Document Server

    Liang, C F; Obert, J; Paris, P; Putaux, J C

    1981-01-01

    By chemical evaporation technique, neutron deficient hafnium nuclei have been on-line separated at the ISOCELE facility, from the isobar rare-earth elements, in the metal-fluoride HfF/sub 3//sup +/ ion form. Half-lives of /sup 162-165/Hf have been measured. Similarly, tantalum has been selectively separated on the TaF/sub 4//sup +/ form. (4 refs) .

  8. Hafnium nitride buffer layers for growth of GaN on silicon

    Science.gov (United States)

    Armitage, Robert D.; Weber, Eicke R.

    2005-08-16

    Gallium nitride is grown by plasma-assisted molecular-beam epitaxy on (111) and (001) silicon substrates using hafnium nitride buffer layers. Wurtzite GaN epitaxial layers are obtained on both the (111) and (001) HfN/Si surfaces, with crack-free thickness up to 1.2 {character pullout}m. However, growth on the (001) surface results in nearly stress-free films, suggesting that much thicker crack-free layers could be obtained.

  9. The control of weathering processes on riverine and seawater hafnium isotope ratios

    OpenAIRE

    Bayon, Germain; Vigier, Nathalie; Burton, Kevin W.; Brenot, Agnès; Carignan, Jean; Etoubleau, Joel; Chu, Nan-chin

    2006-01-01

    Hafnium Hf-176/Hf-177 isotope ratio variations in marine records are thought to reflect changes in continental weathering through time, but the behavior of Hf in rivers, and during weathering, is not well understood. Here, we present Hf-176/Hf-177 data for rivers, bedrock, soils, and leaching experiments for the Moselle basin, Vosges, France. These data strongly suggest that the Hf-176/Hf-177 composition of river waters is controlled by preferential dissolution of accessory phases (i.e., apat...

  10. Precipitation of zirconium and hafnium basic sulphates from chloride and nitrate solutions

    Energy Technology Data Exchange (ETDEWEB)

    Nekhamkin, L.G.; Tsylov, Yu.A.; Reznik, A.M.; Kondrashova, I.A.; Bogushevskaya, S.V. (Gosudarstvennyj Nauchno-Issledovatel' skij i Proektnyj Inst. Redkometallicheskoj Promyshlennosti, Moscow (USSR); Moskovskij Inst. Tonkoj Khimicheskoj Tekhnologii (USSR))

    1984-01-01

    Solubility of main zirconium sulfates (MZS) and that of hafnium (MHS) in systems with nitric and hydrochloric acids is studied. It was interesting to compare the strength of zirconium and hafnium hydrocomplexes in the process of precipitation of main sulfates, as well as the ability of some ligands to form complexes with these metals. Experiments are performed in the range of comparatively low concentrations of acids that are most valuable from the practical point of view. Chemical analysis of equilibrium phases is performed by gravimetric and volumetric methods. To determine substantial composition of sediments thermographic and X-ray graphic analysis are used. It is shown that formation of hydroxosulfate complexes from hydroxochloride of hydroxobromide complexes is difficult as compared with their formation from hydroxonitrate or hydroxoperchlorate. As regards the strength of bond with zirconium and hafnium in hydrolized complexes, acidoligands must be placed in the Clo/sub 4//sup -/

  11. Hafnium metallocene compounds used as cathode interfacial layers for enhanced electron transfer in organic solar cells.

    Science.gov (United States)

    Park, Keunhee; Oh, Seungsik; Jung, Donggeun; Chae, Heeyeop; Kim, Hyoungsub; Boo, Jin-Hyo

    2012-01-09

    We have used hafnium metallocene compounds as cathode interfacial layers for organic solar cells [OSCs]. A metallocene compound consists of a transition metal and two cyclopentadienyl ligands coordinated in a sandwich structure. For the fabrication of the OSCs, poly[3,4-ethylenedioxythiophene]:poly(styrene sulfonate), poly(3-hexylthiophene-2,5-diyl) + 66-phenyl C61 butyric acid methyl ester, bis-(ethylcyclopentadienyl)hafnium(IV) dichloride, and aluminum were deposited as a hole transport layer, an active layer, a cathode interfacial layer, and a cathode, respectively. The hafnium metallocene compound cathode interfacial layer improved the performance of OSCs compared to that of OSCs without the interfacial layer. The current density-voltage characteristics of OSCs with an interfacial layer thickness of 0.7 nm and of those without an interfacial layer showed power conversion efficiency [PCE] values of 2.96% and 2.34%, respectively, under an illumination condition of 100 mW/cm2 (AM 1.5). It is thought that a cathode interfacial layer of an appropriate thickness enhances the electron transfer between the active layer and the cathode, and thus increases the PCE of the OSCs.

  12. CO assisted N2 functionalization activated by a dinuclear hafnium complex: a DFT mechanistic exploration.

    Science.gov (United States)

    Ma, Xuelu; Zhang, Xin; Zhang, Wenchao; Lei, Ming

    2013-01-21

    In this paper, the reaction mechanisms of CO assisted N(2) cleavage and functionalization activated by a dinuclear hafnium complex are studied using a density function theory (DFT) method. Several key intermediates (Ia, Ib, Ic and Id) with axial/equatorial N=C=O coordination structures are found to be of importance along reaction pathways of CO assisted N(2) functionalization, which could provide a profound theoretical insight into the C-N bond formation and N-N bond cleavage. There are two different attack directions to insert the first CO molecule into the Hf-N bonds of the dinuclear hafnium complex, which lead to C-N bond formation. The calculated results imply that CO insertion into the Hf(1)-N(3) bond (Path A1) reacts more easily than that into the Hf(2)-N(3) bond (Path A3). But for the insertion of the second CO insertion to give 2A, there are two possibilities (Path A1 and Path A2) according to this insertion being after/before N-N bond cleavage. Two pathways (Path A1 and Path A2) are proved to be possible to form final dinitrogen functionalized products (oxamidide 2A, 2B and 2C) in this study, which explain the formation of different oxamidide isomers in CO assisted N(2) functionalization activated by a dinuclear hafnium complex.

  13. X-Ray diffraction study of KTiOPO{sub 4} single crystals doped with hafnium

    Energy Technology Data Exchange (ETDEWEB)

    Novikova, N. E., E-mail: natnov@ns.crys.ras.ru; Verin, I. A.; Sorokina, N. I.; Alekseeva, O. A. [Russian Academy of Sciences, Shubnikov Institute of Crystallography (Russian Federation); Orlova, E. I.; Voronkova, V. I. [Moscow State University, Faculty of Physics (Russian Federation)

    2011-05-15

    Single crystals of KTi{sub 1-x}Hf{sub x}OPO{sub 4} (x = 0.015(2), 0.035(1), and 0.128(1) are reinvestigated by precision X-ray diffraction at room temperature. It is found that the implantation of hafnium atoms in the crystal structure of KTiOPO{sub 4} does not lead to significant changes in the framework and affects only the positions of the potassium atoms in the channel. Our studies reveal the displacements of the potassium atoms from their main and additional positions in the structure of pure KTP in all three structures studied. The largest displacements from the K1 Prime and K1 Double-Prime additional positions are observed in the structure with x = 0.035. At this hafnium concentration, the occupancy of the main positions of potassium atoms decreases and the occupancy of the additional positions increases in relation to those in KTP. This redistribution of potassium atoms enhances the nonuniformity of distribution of the electron density in the vicinity of their positions, which is probably responsible for the increase in the nonlinear susceptibility of KTP crystals that contain 3.5% hafnium in relation to crystals of pure KTP.

  14. Activation cross sections of proton induced nuclear reactions on natural hafnium

    Energy Technology Data Exchange (ETDEWEB)

    Takacs, S., E-mail: s.takacs@atomki.hu [Institute of Nuclear Research of the Hungarian Academy of Sciences (ATOMKI), 4026 Debrecen (Hungary); Tarkanyi, F. [Institute of Nuclear Research of the Hungarian Academy of Sciences (ATOMKI), 4026 Debrecen (Hungary); Hermanne, A.; Adam Rebeles, R. [Cyclotron Laboratory, Vrije Universiteit Brussel (VUB), Laarbeeklaan 103, 1090 Brussels (Belgium)

    2011-12-01

    Highlights: {yields} Cross sections of proton induced reactions on natural hafnium. {yields} Production of Ta, Hf and Lu isotopes. {yields} Comparison of experimental cross sections with results of TENDL-2010 calculations. {yields} Thick target yields. {yields} Production of {sup 177g}Lu for medical use is not feasible. - Abstract: In a systematic study of light charged particle induced nuclear reactions we investigated the excitation functions of proton induced reactions on natural hafnium targets. Experimental excitation functions of proton induced reactions up to 36 MeV on high purity natural hafnium were determined using the stacked foil activation technique. High resolution off-line gamma-ray spectrometry was applied to assess the activity of each foil. From the measured activity independent and/or cumulative elemental or isotopic cross section data for production of Ta, Hf and Lu radioisotopes were determined. The experimental data were compared to the data published earlier by other authors and to results of TALYS theoretical code taken from TENDL-2010 database. Thick target yields of the investigated radionuclides were calculated from the excitation function that was deduced as an analytical fit to our experimental cross section data points.

  15. Dimensional, microstructural and compositional stability of metal fuels. Final performance report

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, A.A.; Dayananda, M.A.

    1993-03-15

    The projects undertaken were to address two areas of concern for metal-fueled fast reactors: metallurgical compatibility of fuel and its fission products with the stainless steel cladding, and effects of porosity development in the fuel on fuel/cladding interactions and on sodium penetration in fuel. The following studies are reported on extensively in appendices: hot isostatic pressing of U-10Zr by coupled boundary diffusion/power law creep cavitation, liquid Na intrusion into porous U-10Zr fuel alloy by differential capillarity, interdiffusion between U-Zr fuel and selected Fe-Ni-Cr alloys, interdiffusion between U-Zr fuel vs selected cladding steels, and interdiffusion of Ce in Fe-base alloys with Ni or Cr.

  16. The post irradiation examinations of twenty-years stored spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sasahara, A.; Matsumura, T. [Central Research Institute of Electric Power Industry, Tokyo (Japan)

    2000-07-01

    The post irradiation examinations (PIE) on the twenty years stored spent fuels were carried out to evaluate fuel integrity during storage. The spent BWR-MOX fuel rods and PWR-UO{sub 2} fuel rod irradiated in commercial LWR were used. The burnup of the BWR-MOX fuels (five fuel rods) are about 20 GWd/t and that of the PWR-UO{sub 2} fuel is 58 GWd/t. PIE items in this study are, a) visual inspection of the cladding surface, b) puncture test or gas analysis in capsule, c) ceramographic examination to observe oxide layer thickness on outside/inside cladding and pellet microstructure such as grain size, d) electron probe microanalysis (EPMA) on pellet, e) hydrogen content in cladding. The preliminary result shows that twenty years stored fuel rods were not different from that before storage. (authors)

  17. Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code

    Directory of Open Access Journals (Sweden)

    Giovedi Claudia

    2016-01-01

    Full Text Available Early pressurized water reactors were originally designed to operate using stainless steel as cladding material, but during their lifetime this material was replaced by zirconium-based alloys. However, after the Fukushima Daiichi accident, the problems related to the zirconium-based alloys due to the hydrogen production and explosion under severe accident brought the importance to assess different materials. In this sense, initiatives as ATF (Accident Tolerant Fuel program are considering different material as fuel cladding and, one candidate is iron-based alloy. In order to assess the fuel performance of fuel rods manufactured using iron-based alloy as cladding material, it was necessary to select a specific stainless steel (type 348 and modify properly conventional fuel performance codes developed in the last decades. Then, 348 stainless steel mechanical and physics properties were introduced in the TRANSURANUS code. The aim of this paper is to present the obtained results concerning the verification of the modified TRANSURANUS code version against data collected from the open literature, related to reactors which operated using stainless steel as cladding. Considering that some data were not available, some assumptions had to be made. Important differences related to the conventional fuel rods were taken into account. Obtained results regarding the cladding behavior are in agreement with available information. This constitutes an evidence of the modified TRANSURANUS code capabilities to perform fuel rod investigation of fuel rods manufactured using 348 stainless steel as cladding material.

  18. Correlations between nuclear data and integral slab experiments: the case of hafnium; Correlations entre donnees nucleaires et experiences integrales a plaques: le cas du hafnium

    Energy Technology Data Exchange (ETDEWEB)

    Palau, J.M

    1999-07-01

    The aim of this work was to evaluate how much integral slab experiments can both reduce discrepancies between experimental results and calculations, and improve the knowledge of hafnium isotopes neutronic parameters by an adapted sensitivity and uncertainty method. A statistical approach, based on the generalized least squares method and perturbation theory, has been incorporated into our calculation system in order to deduce microscopic cross-section adjustments from observed integral measurements on this particular 'mock-up' reactor.In this study it has been established that the correlations between integral parameters and hafnium capture cross-sections enable specific variations in the region of resolved resonances at the level of multigroup and punctual cross-sections recommended data (JEF-2.2 evaluation) to be highlighted. The use of determinist methods together with Monte Carlo- type simulations enabled a depth analysis of the modelling approximations to be carried out. Furthermore, the sensitivity coefficient validation technique employed leads to a reliable assessment of the quality of the new basic nuclear data. In this instance, the adjustments proposed for certain isotope {sup 177}Hf resonance parameters reduce, after error propagation, by 3 to 5 per cent the difference between experimental results and calculations related to this absorbent's efficiency. Beyond this particular application, the qualification methodology integrated in our calculation system should enable other basic sizing parameters to be treated (chemical / geometric data or other unexplored nuclear data) to make technological requirements less stringent. (author)

  19. Bragg grating induced cladding mode coupling due to asymmetrical index modulation in depressed cladding fibers

    DEFF Research Database (Denmark)

    Berendt, Martin Ole; Grüne-Nielsen, Lars; Soccolich, C.F.

    1998-01-01

    UV-written Bragg gratings find wide spread use as wavelength selective components. In reflection high extinction ratios are routinely obtained. However, coupling to cladding modes gives excess loss on the short wavelength side of the main reflection. Different fiber-designs have been proposed...... to reduce this problem. None of these designs seems to give complete solutions. In particular, the otherwise promising depressed cladding design gives a pronounced coupling to one LP01 mode, this has been referred to as a Ghost grating. To find the modes of the fiber we have established a numerical mode......-solver based on the staircase-approximation method. The Bragg grating causes coupling between the fundamental LP01 mode and higher order LP1p modes that satisfy phase-matching. The coupling strength is determined by the overlap integral of the LP01, the LP1p mode, and the UV-induced index perturbation. For LP0...

  20. Low-voltage bendable pentacene thin-film transistor with stainless steel substrate and polystyrene-coated hafnium silicate dielectric.

    Science.gov (United States)

    Yun, Dong-Jin; Lee, Seunghyup; Yong, Kijung; Rhee, Shi-Woo

    2012-04-01

    The hafnium silicate and aluminum oxide high-k dielectrics were deposited on stainless steel substrate using atomic layer deposition process and octadecyltrichlorosilane (OTS) and polystyrene (PS) were treated improve crystallinity of pentacene grown on them. Besides, the effects of the pentacene deposition condition on the morphologies, crystallinities and electrical properties of pentacene were characterized. Therefore, the surface treatment condition on dielectric and pentacene deposition conditions were optimized. The pentacene grown on polystyrene coated high-k dielectric at low deposition rate and temperature (0.2-0.3 Å/s and R.T.) showed the largest grain size (0.8-1.0 μm) and highest crystallinity among pentacenes deposited various deposition conditions, and the pentacene TFT with polystyrene coated high-k dielectric showed excellent device-performance. To decrease threshold voltage of pentacene TFT, the polystyrene-thickness on high-k dielectric was controlled using different concentration of polystyrene solution. As the polystyrene-thickness on hafnium silicate decreases, the dielectric constant of polystyrene/hafnium silicate increases, while the crystallinity of pentacene grown on polystyrene/hafnium silicate did not change. Using low-thickness polystyrene coated hafnium silicate dielectric, the high-performance and low voltage operating (1 × 10(4)) and complementary inverter (DC gains, ~20) could be fabricated.

  1. Investigations on dry sliding of laser cladded aluminum bronze

    Directory of Open Access Journals (Sweden)

    Freiße Hannes

    2016-01-01

    Full Text Available The aim of this study was to investigate the tribological behaviour of laser cladded aluminum bronze tool surfaces for dry metal forming. In a first part of this work a process window for cladding aluminum bronze on steel substrate was investigated to ensure a low dilution. Therefore, the cladding speed, the powder feed rate, the laser power and the distance between the process head and the substrate were varied. The target of the second part was to investigate the influence of different process parameters on the tribological behaviour of the cladded tracks. The laser claddings were carried out on both aluminum bronze and cold work tool steel as substrate materials. Two different particle sizes of the cladding powder material were used. The cladding speed was varied and a post-processing laser remelting treatment was applied. It is shown that the tribological behaviour of the surface in a dry oscillating ball-on-plate test is highly dependent on the substrate material. In the third part a deep drawing tool was additively manufactured by direct laser deposition. Furthermore, the tool was applied to form circular cups with and without lubrication.

  2. Bending of pipes with inconel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Nachpitz, Leonardo; Menezes, Carlos Eduardo B.; Vieira, Carlos R. Tavares [Primus Processamento de Tubos S.A. (PROTUBO), Macae, RJ (Brazil)

    2009-07-01

    The high-frequency induction bending process, using API pipes coated with Inconel 625 reconciled to a mechanical transformation for a higher degree of resistance, was developed through a careful specification and control of the manufacturing parameters and inherent heat treatments. The effects of this technology were investigated by a qualification process consisting of a sequence of tests and acceptance criteria typically required by the offshore industry, and through the obtained results was proved the effectiveness of this entire manufacturing process, without causing interference in the properties and the quality of the inconel cladding, adding a gain of resistance to the base material, guaranteed by the requirements of the API 5L Standard. (author)

  3. Weld overlay cladding with iron aluminides

    Energy Technology Data Exchange (ETDEWEB)

    Goodwin, G.M. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    The hot and cold cracking tendencies of some early iron aluminide alloy compositions have limited their use in applications where good weldability is required. Using hot crack testing techniques invented at ORNL, and experimental determinations of preheat and postweld heat treatment needed to avoid cold cracking, we have developed iron aluminide filler metal compositions which can be successfully used to weld overlay clad various substrate materials, including 9Cr-1Mo steel, 2-1/4Cr-1Mo steel, and 300-series austenitic stainless steels. Dilution must be carefully controlled to avoid crack-sensitive deposit compositions. The technique used to produce the current filler metal compositions is aspiration-casting, i.e. drawing the liquid from the melt into glass rods. Future development efforts will involve fabrication of composite wires of similar compositions to permit mechanized gas tungsten arc (GTA) and/or gas metal arc (GMA) welding.

  4. Metal-Element Compounds of Titanium, Zirconium, and Hafnium as Pyrotechnic Fuels

    Science.gov (United States)

    2015-05-04

    Atlantic Equipment Engineers ( AEE ), Alfa Aesar, and American Elements. These were characterized by X-ray diffraction (XRD), X-ray fluorescence (XRF... AEE phase pure - mostly 1-7 minimally processed TiC Alfa Aesar phase pure (TiC0 93) Fe, Cr, V fines < 1 intermediate 2-8 minimally processed...TiN AEE phase pure Fe fines < 1 intermediate 2-8 milled TiSi2 AEE trace TiSi, trace SiO2 Fe, Cr, Al fines < 2 int. 5-20, coarse 50-100 milled

  5. Characterization of spent fuel approved testing material--ATM-104

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.

  6. Deep-probe metal-clad waveguide biosensors

    DEFF Research Database (Denmark)

    Skivesen, Nina; Horvath, Robert; Thinggaard, S.

    2007-01-01

    Two types of metal-clad waveguide biosensors, so-called dip-type and peak-type, are analyzed and tested. Their performances are benchmarked against the well-known surface-plasmon resonance biosensor, showing improved probe characteristics for adlayer thicknesses above 150-200 nm. The dip-type metal......-clad waveguide sensor is shown to be the best all-round alternative to the surface-plasmon resonance biosensor. Both metal-clad waveguides are tested experimentally for cell detection, showing a detection linut of 8-9 cells/mm(2). (c) 2006 Elsevier B.V. All rights reserved....

  7. FY17 Progress in Modeling of Lanthanide Transport in Metallic Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Matthews, Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-19

    A critical review of fuel-clad-chemical interactions along with modelling requirements is published. The mechanism of lanthanide transport is studied experimentally (NEUP collaboration) and using simulations and initial results are published in Refs.

  8. Effect of Partial Cladding Pattern of Aluminum 7075 T651 on Corrosion and Mechanical Properties

    Directory of Open Access Journals (Sweden)

    E. Rendell

    2017-01-01

    Full Text Available The corrosion resistance of aluminum 7075 T651 in full clad (Alclad, partial clad, and bare (unclad forms was compared after 300 hours of corrosion exposure in an acidic salt spray cabinet test at 36°C. After corrosion exposure, severe to moderate exfoliation corrosion was observed on the unprotected medium sized test panel, light general corrosion was observed on the partially clad panel, and patches of corrosion not penetrating the clad layer were observed on the fully clad panel. After corrosion tests, the tensile strength of partially clad, fully clad, and unprotected panels decreased by 3.4%, 4.0%, and 5.3%, respectively.

  9. Risk-Informed Margin Management (RIMM) Industry Applications IA1 - Integrated Cladding ECCS/LOCA Performance Analysis - Problem Statement

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngblood, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Frepoli, Cesare [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yurko, Joseph P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swindlehurst, Gregg [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the LOCA/ECCS acceptance criteria to include the effects of higher burnup on cladding performance as well as to address some other issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in the summer of 2016. The impact of the final 50.46c rule on the industry will involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or reanalyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 5-10 years following the rule effective date. The need to use advanced cladding designs is expected. A loss of operational margin will result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin.

  10. Advanced Fuels Campaign FY 2015 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Braase, Lori Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States); Carmack, William Jonathan [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-29

    The mission of the Advanced Fuels Campaign (AFC) is to perform research, development, and demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This report is a compilation of technical accomplishment summaries for FY-15. Emphasis is on advanced accident-tolerant LWR fuel systems, advanced transmutation fuels technologies, and capability development.

  11. Strong influence of polymer architecture on the microstructural evolution of hafnium-alkoxide-modified silazanes upon ceramization.

    Science.gov (United States)

    Papendorf, Benjamin; Nonnenmacher, Katharina; Ionescu, Emanuel; Kleebe, Hans-Joachim; Riedel, Ralf

    2011-04-04

    The present study focuses on the synthesis and ceramization of novel hafnium-alkoxide-modified silazanes as well as on their microstructure evolution at high temperatures. The synthesis of hafnia-modified polymer-derived SiCN ceramic nanocomposites is performed via chemical modification of a polysilazane and of a cyclotrisilazane, followed by cross-linking and pyrolysis in argon atmosphere. Spectroscopic investigation (i.e., NMR, FTIR, and Raman) shows that the hafnium alkoxide reacts with the N-H groups of the cyclotrisilazane; in the case of polysilazane, reactions of N-H as well as Si-H groups with the alkoxide are observed. Consequently, scanning and transmission electron microscopy studies reveal that the ceramic nanocomposites obtained from cyclotrisilazane and polysilazane exhibited markedly different microstructures, which is a result of the different reaction pathways of the hafnium alkoxide with cyclotrisilazane and with polysilazane. Furthermore, the two prepared ceramic nanocomposites are unexpectedly found to exhibit extremely different high-temperature behavior with respect to decomposition and crystallization; this essential difference is found to be related to the different distribution of hafnium throughout the ceramic network in the two samples. Thus, the homogeneous distribution of hafnium observed in the polysilazane-derived ceramic leads to an enhanced thermal stability with respect to decomposition, whereas the local enrichment of hafnium within the matrix of the cyclotrisilazane-based sample induces a pronounced decomposition upon annealing at high temperatures. The results indicate that the chemistry and architecture of the precursor has a crucial effect on the microstructure of the resulting ceramic material and consequently on its high-temperature behavior. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  12. Study on the formation of self-assembled monolayers on sol-gel processed hafnium oxide as dielectric layers.

    Science.gov (United States)

    Ting, Guy G; Acton, Orb; Ma, Hong; Ka, Jae Won; Jen, Alex K-Y

    2009-02-17

    High dielectric constant (k) metal oxides such as hafnium oxide (HfO2) have gained significant interest due to their applications in microelectronics. In order to study and control the surface properties of hafnium oxide, self-assembled monolayers (SAMs) of four different long aliphatic molecules with binding groups of phosphonic acid, carboxylic acid, and catechol were formed and characterized. Surface modification was performed to improve the interface between metal oxide and top deposited materials as well as to create suitable dielectric properties, that is, leakage current and capacitance densities, which are important in organic thin film transistors. Attenuated total reflectance Fourier transform infrared (ATR-FTIR) spectroscopy, contact angle goniometry, atomic force microscopy (AFM), and simple metal-HfO2-SAM-metal devices were used to characterize the surfaces before and after SAM modification on sol-gel processed hafnium oxide. The alkylphosphonic acid provided the best monolayer formation on sol-gel processed hafnium oxide to generate a well-packed, ultrathin dielectric exhibiting a low leakage current density of 2x10(-8) A/cm2 at an applied voltage of -2.0 V and high capacitance density of 0.55 microF/cm2 at 10 kHz. Dialkylcatechol showed similar characteristics and the potential for using the catechol SAMs to modify HfO2 surfaces. In addition, the integration of this alkylphosphonic acid SAM/hafnium oxide hybrid dielectric into pentacene-based thin film transistors yields low-voltage operation within 1.5 V and improved performance over bare hafnium oxide.

  13. The application of electrorefining for recovery and purification of fuel discharged from the Integral Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burris, L.; Steunenberg, R.K.; Miller, W.E.

    1986-01-01

    An electrorefining process employing a molten salt electrolyte and a molten cadmium anode is proposed for the separation of uranium and plutonium from fission products and cladding material in discharged IFR driver fuel. The use of a liquid cadmium anode, which is the unique feature of the process, permits selective dissolution of the fuel from the cladding and prevents electrolytic corrosion of the steel container and contamination of the product by noble metal fission products.

  14. High Temperature Resistance Claddings for Nuclear Thermal Rockets Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This program will develop a series of nano-/micro-composite coated nuclear reactor facing components using MesoCoat's CermaCladTM process. This proposed SBIR program...

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

    Energy Technology Data Exchange (ETDEWEB)

    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 ß-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.

  16. Chalcogenide optical microwires cladded with fluorine-based CYTOP.

    Science.gov (United States)

    Li, Lizhu; Abdukerim, Nurmemet; Rochette, Martin

    2016-08-22

    We demonstrate optical transmission results of highly nonlinear As2Se3 optical microwires cladded with fluorine-based CYTOP, and compare them with microwires cladded with typical hydrogen-based polymers. In the linear optics regime, the CYTOP-cladded microwire transmits light in the spectral range from 1.3 µm up to >2.5 µm without trace of absorption peaks such as those observed using hydrogen-based polymer claddings. The microwire is also pumped in the nonlinear optics regime, showing multiple-orders of four-wave mixing and supercontinuum generation spanning from 1.0 µm to >4.3 µm. We conclude that with such a broadband transparency and high nonlinearity, the As2Se3-CYTOP microwire is an appealing solution for nonlinear optical processing in the mid-infrared.

  17. A Multi-Scale Modeling of Laser Cladding Process (Preprint)

    National Research Council Canada - National Science Library

    Cao, J; Choi, J

    2006-01-01

    Laser cladding is an additive manufacturing process that a laser generates a melt-pool on the substrate material while a second material, as a powder or a wire form, is injected into that melt-pool...

  18. Separate-effect tests on zirconium cladding degradation in air ingress situations

    Energy Technology Data Exchange (ETDEWEB)

    Duriez, C. [Institut de Radioprotection et de Surete Nucleaire, IRSN, Direction de Prevention des Accidents Majeurs, Centre de Cadarache, 13115 St Paul Lez Durance (France)], E-mail: christian.duriez@irsn.fr; Steinbrueck, M. [Forschungszentrum Karlsruhe, FZK, Institut fuer Materialforschung, Postfach 3640, 76021 Karlsruhe (Germany); Ohai, D.; Meleg, T. [Institute for Nuclear Research, INR, Nuclear Material and Corrosion Department, Pitesti, 115400 Mioveni Arges (Romania); Birchley, J.; Haste, T. [Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)

    2009-02-15

    In the event of air ingress during a reactor or spent fuel pond low probability accident, the fuel rods will be exposed to air-containing atmospheres at high temperatures. In comparison with steam, the presence of air is expected to result in a more rapid escalation of the accident. A state-of-the-art review performed before SARNET started showed that the existing data on zirconium alloy oxidation in air were scarce. Moreover, the exact role of zirconium nitride on the cladding degradation process was poorly understood. Regarding the cladding behaviour in air + steam or nitrogen-enriched atmospheres (encountered in oxygen-starved conditions), almost no data were available. New experimental programmes comprising small-scale tests have therefore been launched at FZK, IRSN (MOZART programme in the frame of the International Source Term Program-ISTP) and INR. Zircaloy-4 cladding in PWR (FZK, IRSN) and in CANDU (INR) geometry are investigated. On-line kinetic data are obtained on centimetre size tube segments, by thermogravimetry (FZK, IRSN and INR) or by mass spectrometry (FZK). Plugged tubes 15 cm long (FZK) are also investigated. The samples are air-oxidised either in the 'as-received' state, or after pre-oxidation in steam. 'Analytical' tests at constant temperature and gas composition provide basic kinetic data, while more prototypical temperature transients and sequential gas compositions are also investigated. The temperature domains extend from 600 deg. C up to 1500 deg. C. Systematic post-test metallographic inspections are performed. The paper gives a synthesis of the results obtained, comparing them in terms of kinetics and oxide scale structure and composition. A comparative analysis is performed with results of the QUENCH-10 (Q-10) bundle test, which included an air ingress phase. It is shown how the data contribute to a better understanding of the cladding degradation process, especially regarding the role of nitrogen. For modelling of

  19. R&D Plan for RISMC Industry Application #1: ECCS/LOCA Cladding Acceptance Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron Simon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tu, Lei [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-04-01

    The Nuclear Regulatory Commission (NRC) is finalizing a rulemaking change that would revise the requirements in 10 CFR 50.46. In the proposed new rulemaking, designated as 10 CFR 50.46c, the NRC proposes a fuel performance-based equivalent cladding reacted (ECR) criterion as a function of cladding hydrogen content before the accident (pre-transient) in order to include the effects of higher burnup on cladding performance as well as to address other technical issues. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee costs as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. The Idaho National Laboratory (INL) has initiated a project, as part of the DOE Light Water Reactor Sustainability Program (LWRS), to develop analytical capabilities to support the industry in the transition to the new rule. This project is called the Industry Application 1 (IA1) within the Risk-Informed Safety Margin Characterization (RISMC) Pathway of LWRS. The general idea behind the initiative is the development of an Integrated Evaluation Model (IEM). The motivation is to develop a multiphysics framework to analyze how uncertainties are propagated across the stream of physical disciplines and data involved, as well as how risks are evaluated in a LOCA safety analysis as regulated under 10 CFR 50.46c. This IEM is called LOTUS which stands for LOCA Toolkit for US, and it represents the LWRS Program’s response to the proposed new rule making. The focus of this report is to complete an R&D plan to describe the demonstration of the LOCA/ECCS RISMC Industry Application # 1 using the advanced RISMC Toolkit and methodologies. This report includes the description and development plan for a RISMC LOCA tool that fully couples advanced MOOSE tools already in development in order to characterize and optimize

  20. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

  1. Explosion Clad for Upstream Oil and Gas Equipment

    Science.gov (United States)

    Banker, John G.; Massarello, Jack; Pauly, Stephane

    2011-01-01

    Today's upstream oil and gas facilities frequently involve the combination of high pressures, high temperatures, and highly corrosive environments, requiring equipment that is thick wall, corrosion resistant, and cost effective. When significant concentrations of CO2 and/or H2S and/or chlorides are present, corrosion resistant alloys (CRA) can become the material of choice for separator equipment, piping, related components, and line pipe. They can provide reliable resistance to both corrosion and hydrogen embrittlement. For these applications, the more commonly used CRA's are 316L, 317L and duplex stainless steels, alloy 825 and alloy 625, dependent upon the application and the severity of the environment. Titanium is also an exceptional choice from the technical perspective, but is less commonly used except for heat exchangers. Explosion clad offers significant savings by providing a relatively thin corrosion resistant alloy on the surface metallurgically bonded to a thick, lower cost, steel substrate for the pressure containment. Developed and industrialized in the 1960's the explosion cladding technology can be used for cladding the more commonly used nickel based and stainless steel CRA's as well as titanium. It has many years of proven experience as a reliable and highly robust clad manufacturing process. The unique cold welding characteristics of explosion cladding reduce problems of alloy sensitization and dissimilar metal incompatibility. Explosion clad materials have been used extensively in both upstream and downstream oil, gas and petrochemical facilities for well over 40 years. The explosion clad equipment has demonstrated excellent resistance to corrosion, embrittlement and disbonding. Factors critical to insure reliable clad manufacture and equipment design and fabrication are addressed.

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

    Science.gov (United States)

    Forsberg, Charles W.

    1992-01-01

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

  3. Hafnium at subduction zones: isotopic budget of input and output fluxes; L'hafnium dans les zones de subduction: bilan isotopique des flux entrant et sortant

    Energy Technology Data Exchange (ETDEWEB)

    Marini, J.Ch

    2004-05-15

    Subduction zones are the primary regions of mass exchanges between continental crust and mantle of Earth through sediment subduction toward the earth's mantle and by supply of mantellic magmas to volcanic arcs. We analyze these mass exchanges using Hafnium and Neodymium isotopes. At the Izu-Mariana subduction zone, subducting sediments have Hf and Nd isotopes equivalent to Pacific seawater. Altered oceanic crust has Hf and Nd isotopic compositions equivalent to the isotopic budget of unaltered Pacific oceanic crust. At Luzon and Java subduction zones, arc lavas present Hf isotopic ratios highly radiogenic in comparison to their Nd isotopic ratios. Such compositions of the Luzon and Java arc lavas are controlled by a contamination of their sources by the subducted oceanic sediments. (author)

  4. Fuel rod pressure in nuclear power reactors: Statistical evaluation of the fuel rod internal pressure in LWRs with application to lift-off probability

    Energy Technology Data Exchange (ETDEWEB)

    Jelinek, Tomas

    2001-02-01

    In this thesis, a methodology for quantifying the risk of exceeding the Lift-off limit in nuclear light water power reactors is outlined. Due to fission gas release, the pressure in the gap between the fuel pellets and the cladding increases with burnup of the fuel. An increase in the fuel-clad gap due to clad creep would be expected to result in positive feedback, in the form of higher fuel temperatures, leading to more fission gas release, higher rod pressure, etc, until the cladding breaks. An increase in the fuel-clad gap that leads to this positive feedback is a phenomenon called Lift-off and is a limitation that must be considered in the fuel core management. Lift-off is a consequence of very high internal fuel rod pressure. The internal fuel rod pressure is therefore used as a Lift-off indicator. The internal fuel rod pressure is closely connected to the fission gas release into the fuel rod plenum and is thus used to increase the database. It is concluded that the dominating error source in the prediction of the pressure in Boiling Water Reactors (BWR), is the power history. There is a bias in the fuel pressure prediction that is dependent on the fuel rod position in the fuel assembly for BWRs. A methodology to quantify the risk of the fuel rod internal pressure exceeding a certain limit is developed; the risk is dependent of the pressure prediction and the fuel rod position. The methodology is based on statistical treatment of the discrepancies between predicted and measured fuel rod internal pressures. Finally, a methodology to estimate the Lift-off probability of the whole core is outlined.

  5. Inter-Diffusion of Copper and Hafnium as Studied by X-Ray Photoelectron Spectroscopy

    Science.gov (United States)

    Pearson, Justin Seth

    The purpose of this study is to investigate the interdiffusion of copper and hafnium. Thin films (thicknesses ranging from 100 nm to 150 nm) of hafnium were deposited on a silicon substrate. About 80 nm of copper was then deposited on such samples. High purity samples have been used in this investigation. The deposition of the elements was done by the e-beam technique. The interfaces thus formed were annealed for a fixed time (30 minutes) at temperatures of 100, 200, and 300°C. The samples were characterized in situ by the x-ray photoelectron spectroscopy technique. To carry out the depth profiling of these samples a controlled amount of the over layer was removed and the spectral data were acquired. The argon ion sputtering technique was used to sputter the layers away. Spectral data in the copper 2p and hafnium 4f regions were investigated. The atomic concentration of the constituents as a function of depth across the interface was determined by analyzing the areas under the curves. The depth profiling data thus obtained was analyzed by the Matano-Boltzmann's procedure. For this analysis the Matano plane was determined based on the criteria of equal area on each side of the interface. The Fick's Law second law was used to calculate the interdiffuison coefficient for each of these interfaces. The interdiffusion coefficient as a function of temperature was determined from these analyses. From these coefficients the activation energy and the pre-exponential factor was determined by using the Arrhenius plot. The activation energy was found to be 0.128 eV/atom and the pre-exponential factor was 3.33E-14 cm2/s. The results from this investigation will be useful in the application of Cu/Hf interface in design and fabrication of semiconductor devices.

  6. FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

    1981-01-01

    FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

  7. Study of oxide and α-Zr(O) growth kinetics from high temperature steam oxidation of Zircaloy-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sawarn, Tapan K., E-mail: sawarn@barc.gov.in; Banerjee, Suparna, E-mail: sup@barc.gov.in; Samanta, Akanksha, E-mail: akanksha@barc.gov.in; Rath, B.N., E-mail: bibhur@barc.gov.in; Kumar, Sunil, E-mail: sunilkmr@barc.gov.in

    2015-12-15

    Oxidation kinetics of Zircaloy-4 cladding of fuel pins of Indian pressurized heavy water reactors (IPHWRs) under a simulated loss of coolant accident (LOCA) condition was investigated. The kinetic rate constants for the oxide and oxygen stabilized α-Zr phase growth were established from the isothermal metal-steam reaction at high temperatures (900–1200 °C) with soaking periods in the range of 60–900 s. Oxide and α-Zr(O) layer thickness were measured to derive the respective growth rates. The observed rates obeyed a parabolic law and Arrhenius expressions of rate constants were established. Percentage equivalent clad reacted (%ECR) was calculated using Baker-Just equation. Hydrogen estimation was carried out on the oxidized samples using inert gas fusion technique. The hydrogen pick up was found to be in the range 10–30 ppm. The measured values of oxide and α-Zr(O) layer thickness were compared with the results obtained using OXYCON, an indigenously developed model. The model predicts the oxide growth reasonably well but under predicts the α-Zr(O) growth significantly at thickness values higher than 80 μm. - Highlights: • Steam oxidation kinetics of IPHWR fuel cladding material, Zircaloy-4 in the temperature range 900–1200 °C has been studied. • The growth kinetics of the oxide and α-Zr(O) were established from the microstructural analysis. • An indigenously developed model, OXYCON has been validated against the experimental data. • The hydrogen pick up in the cladding during oxidation was observed to be in the range 10–30 ppm.

  8. Zirconium(IV) and Hafnium(IV)-Catalyzed Highly Enantioselective Epoxidation of Homoallylic and Bishomoallylic Alcohols

    Science.gov (United States)

    Li, Zhi; Yamamoto, Hisashi

    2010-01-01

    In this report, zirconium(IV) and hafnium(IV)-bishydroxamic acid complexes were utilized in the highly enantioselective epoxidation of homoallylic alcohols and bishomoallylic alcohols, which used to be quite difficult substrates for other types of asymmetric epoxidation reactions. The performance of the catalyst was improved by adding polar additive and molecular sieves. For homoallylic alcohols, the reaction could provide epoxy alcohols in up to 81% yield and up to 98% ee, while for bishomoallylic alcohols, up to 75% yield and 99% ee of epoxy alcohols rather than cyclize compounds could be obtained in most cases. PMID:20481541

  9. Zirconium(IV)- and hafnium(IV)-catalyzed highly enantioselective epoxidation of homoallylic and bishomoallylic alcohols.

    Science.gov (United States)

    Li, Zhi; Yamamoto, Hisashi

    2010-06-16

    In this report, zirconium(IV)- and hafnium(IV)-bishydroxamic acid complexes were utilized in the highly enantioselective epoxidation of homoallylic alcohols and bishomoallylic alcohols, which used to be quite difficult substrates for other types of asymmetric epoxidation reactions. The performance of the catalyst was improved by adding polar additive and molecular sieves. For homoallylic alcohols, the reaction could provide epoxy alcohols in up to 83% yield and up to 98% ee, while, for bishomoallylic alcohols, up to 79% yield and 99% ee of epoxy alcohols rather than cyclized tetrahydrofuran compounds could be obtained in most cases.

  10. Effect of Hafnium Impurities on the Magnetoresistance of {YBa}2{Cu}3{O}_{7-δ }

    Science.gov (United States)

    Savich, S. V.; Samoylov, A. V.; Kamchatnaya, S. N.; Goulatis, I. L.; Vovk, R. V.; Chroneos, A.; Solovjov, A. L.; Omelchenko, L. V.

    2017-02-01

    In the present study, we investigate the influence of the hafnium (Hf) impurities on the magnetoresistance of {YBa}2{Cu}3{O}_{7-δ } ceramic samples in the temperature interval of the transition to the superconducting state in constant magnetic field up to 12 T. The cause of the appearance of low- temperature "tails" (paracoherent transitions) on the resistive transitions, corresponding to different phase regimes of the vortex matter state is discussed. At temperatures higher than the critical temperature ( T > T_c), the temperature dependence of the excess paraconductivity can be described within the Aslamazov-Larkin theoretical model of the fluctuation conductivity for layered superconductors.

  11. Zirconium alloys for fuel element structures

    Energy Technology Data Exchange (ETDEWEB)

    Bart, G.; Bertsch, J

    2005-07-01

    Today more than 400 light water power reactors (LWRs) operate worldwide providing approximately 17% of the world's electricity demand. One important component for their successful operation is the fuel tube, made out of a zirconium alloy. A huge number of out-of-pile and in-pile experiments have been performed to improve step by step the fuel for higher burn-up and to reduce the failure rates of fuel pins close to zero. The influencing parameters for excellent or poor cladding behaviour are numerous and sometimes counteract each other. The process of cladding corrosion is slow, difficult to follow, the mechanistic understanding at best incomplete. A vast amount of literature documents the abundant tests and comes up with hypotheses and models for the materials behaviour. PSI has supported for the past 20 years the development of high burn-up fuel cladding by microstructural research studies and service work in post-irradiation examination of test pins. This article reviews the development of the cladding tubes, focussing on the chemical and materials science aspects. (author)

  12. Cladding Attachment Over Thick Exterior Insulating Sheathing

    Energy Technology Data Exchange (ETDEWEB)

    Baker, P. [Building Science Corporation, Somerville, MA (United States); Eng, P. [Building Science Corporation, Somerville, MA (United States); Lepage, R. [Building Science Corporation, Somerville, MA (United States)

    2014-01-01

    The addition of insulation to the exterior of buildings is an effective means of increasing the thermal resistance of both wood framed walls as well as mass masonry wall assemblies. For thick layers of exterior insulation (levels greater than 1.5 inches), the use of wood furring strips attached through the insulation back to the structure has been used by many contractors and designers as a means to provide a convenient cladding attachment location (Straube and Smegal 2009, Pettit 2009, Joyce 2009, Ueno 2010). The research presented in this report is intended to help develop a better understanding of the system mechanics involved and the potential for environmental exposure induced movement between the furring strip and the framing. BSC sought to address the following research questions: 1. What are the relative roles of the mechanisms and the magnitudes of the force that influence the vertical displacement resistance of the system? 2. Can the capacity at a specified deflection be reliably calculated using mechanics based equations? 3. What are the impacts of environmental exposure on the vertical displacement of furring strips attached directly through insulation back to a wood structure?

  13. Cladding Attachment Over Thick Exterior Insulating Sheathing

    Energy Technology Data Exchange (ETDEWEB)

    Baker, P. [Building Science Corporation, Somerville, MA (United States); Eng, P. [Building Science Corporation, Somerville, MA (United States); Lepage, R. [Building Science Corporation, Somerville, MA (United States)

    2014-01-01

    The addition of insulation to the exterior of buildings is an effective means of increasing the thermal resistance of both wood framed walls as well as mass masonry wall assemblies. For thick layers of exterior insulation (levels greater than 1.5 inches), the use of wood furring strips attached through the insulation back to the structure has been used by many contractors and designers as a means to provide a convenient cladding attachment location (Straube and Smegal 2009, Pettit 2009, Joyce 2009, Ueno 2010). The research presented in this report is intended to help develop a better understanding of the system mechanics involved and the potential for environmental exposure induced movement between the furring strip and the framing. BSC sought to address the following research questions: 1.What are the relative roles of the mechanisms and the magnitudes of the force that influence the vertical displacement resistance of the system? 2.Can the capacity at a specified deflection be reliably calculated using mechanics based equations? 3.What are the impacts of environmental exposure on the vertical displacement of furring strips attached directly through insulation back to a wood structure?

  14. Weld overlay cladding with iron aluminides

    Energy Technology Data Exchange (ETDEWEB)

    Goodwin, G.M. [Oak Ridge National Lab., TN (United States)

    1997-12-01

    The author has established a range of compositions for these alloys within which hot cracking resistance is very good, and within which cold cracking can be avoided in many instances by careful control of welding conditions, particularly preheat and postweld heat treatment. For example, crack-free butt welds have been produced for the first time in 12-mm thick wrought Fe{sub 3}Al plate. Cold cracking, however, still remains an issue in many cases. The author has developed a commercial source for composite weld filler metals spanning a wide range of achievable aluminum levels, and are pursuing the application of these filler metals in a variety of industrial environments. Welding techniques have been developed for both the gas tungsten arc and gas metal arc processes, and preliminary work has been done to utilize the wire arc process for coating of boiler tubes. Clad specimens have been prepared for environmental testing in-house, and a number of components have been modified and placed in service in operating kraft recovery boilers. In collaboration with a commercial producer of spiral weld overlay tubing, the author is attempting to utilize the new filler metals for this novel application.

  15. Fission gas induced deformation model for FRAP-T6 and NSRR irradiated fuel test simulations

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Sasajima, Hideo; Fuketa, Toyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hosoyamada, Ryuji; Mori, Yukihide

    1996-11-01

    Pulse irradiation tests of irradiated fuels under simulated reactivity initiated accidents (RIAs) have been carried out at the Nuclear Safety Research Reactor (NSRR). Larger cladding diameter increase was observed in the irradiated fuel tests than in the previous fresh fuel tests. A fission gas induced cladding deformation model was developed and installed in a fuel behavior analysis code, FRAP-T6. The irradiated fuel tests were analyzed with the model in combination with modified material properties and fuel cracking models. In Test JM-4, where the cladding temperature rose to higher temperatures and grain boundary separation by the pulse irradiation was significant, the fission gas model described the cladding deformation reasonably well. The fuel had relatively flat radial power distribution and the grain boundary gas from the whole radius was calculated to contribute to the deformation. On the other hand, the power density in the irradiated LWR fuel rods in the pulse irradiation tests was remarkably higher at the fuel periphery than the center. A fuel thermal expansion model, GAPCON, which took account of the effect of fuel cracking by the temperature profile, was found to reproduce well the LWR fuel behavior with the fission gas deformation model. This report present details of the models and their NSRR test simulations. (author)

  16. Application of the Zr/Hf ratio in the determination of hafnium in geochemical samples by high-resolution inductively coupled plasma mass spectrometry.

    Science.gov (United States)

    Liu, Ya Xuan; Li, Qing Xia; Ma, Na; Sun, Xiao Ling; Bai, Jin Feng; Zhang, Qin

    2014-12-02

    Hafnium content and its change are of significance in geochemistry and cosmochemistry; however, the determination of hafnium has always been problematic in analytical chemistry. In this paper, a new idea is proposed for the determination of hafnium in geochemical samples, including rocks, soils, and stream sediments. Through the comparison of two conventional open-type acid digestion methods (HF-HNO3-HClO4 and HF-HNO3-H2SO4), it was found that although neither of these methods could fully digest the zirconium and hafnium in a sample, the zirconium and hafnium digestion behaviors in one sample were consistent in the 60 experimental geochemical reference materials with different properties, so the experimentally determined Zr/Hf ratio in solution could be used to calculate the hafnium content in a sample. In addition, possible mass spectral interferences during the determination of zirconium and hafnium by high resolution inductively coupled plasma mass spectrometry (HR-ICPMS) were studied, and it was found that the mass spectral interferences of the selected isotopes (90)Zr and (178)Hf could be neglected. The mass spectral behaviors of (90)Zr and (178)Hf were also very consistent during the determination by HR-ICPMS. Since the hafnium content was calculated using the ratio value, all of the errors (including the errors in weighing process, the accidental errors during operation and the instrument fluctuation in the determination) of the Zr/Hf ratio could be effectively reduced or even eliminated. The relative standard deviation of the actual samples was lower than 3.2%, and the detection limit of the method (considering the dilution effect and matrix effect during measurement of the Zr/Hf ratio and zirconium content) was 0.04 μg/g. The proposed method could satisfy the requirement for the determination of hafnium in geochemical samples.

  17. Oxygen segregation in pre-hydrided Zircaloy-4 cladding during a simulated LOCA transient

    Directory of Open Access Journals (Sweden)

    Torres Elodie

    2017-01-01

    Full Text Available Oxygen and hydrogen distributions are key elements influencing the residual ductility of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA. During the high temperature oxidation, a complex partitioning of the alloying elements is observed. A finite-difference code for solving the oxygen diffusion equations has been developed by Institut de Radioprotection et de Sûreté Nucléaire to predict the oxygen profile within the samples. The comparison between the calculations and the experimental results in the mixed α+β region shows that the oxygen diffusion is not accurately predicted by the existing modeling. This work aims at determining the key parameters controlling the average oxygen profile within the sample in the two-phase regions at 1200 °C. High temperature steam oxidation tests interrupted by water quench were performed using pre-hydrided Zircaloy-4 samples. Experimental oxygen distribution was measured by Electron Probe Micro-Analysis (EPMA. The phase distributions within the cladding thickness, was measured using image analysis to determine the radial profile of α(O phase fraction. It is further demonstrated and experimentally checked that the α-phase fraction in these regions follows a diffusion-like radial profile. A new phase fraction modeling is then proposed in the cladding metallic part during steam oxidation. The modeling results are compared to a large set of experiments including the influence of exposure duration and hydrogen content. Another key outcome from this modeling is that oxygen average profile is straightforward derived from the proposed modeling.

  18. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  19. Microstructure and optical properties of Pr3+-doped hafnium silicate films

    Science.gov (United States)

    2013-01-01

    In this study, we report on the evolution of the microstructure and photoluminescence properties of Pr3+-doped hafnium silicate thin films as a function of annealing temperature (TA). The composition and microstructure of the films were characterized by means of Rutherford backscattering spectrometry, spectroscopic ellipsometry, Fourier transform infrared absorption, and X-ray diffraction, while the emission properties have been studied by means of photoluminescence (PL) and PL excitation (PLE) spectroscopies. It was observed that a post-annealing treatment favors the phase separation in hafnium silicate matrix being more evident at 950°C. The HfO2 phase demonstrates a pronounced crystallization in tetragonal phase upon 950°C annealing. Pr3+ emission appeared at TA = 950°C, and the highest efficiency of Pr3+ ion emission was detected upon a thermal treatment at 1,000°C. Analysis of the PLE spectra reveals an efficient energy transfer from matrix defects towards Pr3+ ions. It is considered that oxygen vacancies act as effective Pr3+ sensitizer. Finally, a PL study of undoped HfO2 and HfSiOx matrices is performed to evidence the energy transfer. PMID:23336520

  20. Linear Trimeric Hafnium Clusters in Hf0.86(1I3

    Directory of Open Access Journals (Sweden)

    Jan Beekhuizen

    2011-05-01

    Full Text Available The reduction of hafnium tetraiodide, HfI4, with aluminum at 600 °C or 850 °C in the presence of a NaI flux resulted in black single crystals of Hf0.86(1I3. This composition corresponds well to the upper end of the non-stoichiometry range 0.89 ≤ x ≤ 1.00 previously reported for HfxI3. The crystal structure (a = 1250.3(2, c = 1999.6(3 pm, R-3m, Z = 18 is made up of hexagonal closest packed layers of iodide ions. One third of the octahedral holes would be filled as in TiI3 or ZrI3 if it were Hf1.00I3. In Hf0.86(1I3, one out of six octahedral holes along [001] are, however, only occupied by 16%. In contrast to TiI3-I and ZrI3, one striking structural feature is in the formation of linear hafnium trimers with identical Hf―Hf distances of 318.3(2 pm rather than the formation of dimers. These may be associated with Hf―Hf bonding although only 2.64 electrons are available for one Hf5.16I18 column.

  1. Effect of hafnium addition on solidifi cation structure of cast Ti-46Al alloys

    Directory of Open Access Journals (Sweden)

    Su Yanqing

    2008-11-01

    Full Text Available To investigate the effect of hafnium addition on the solidifi cation structure, Ti-46Al alloys with nominal compositions of Ti-46Al-xHf (x = 0, 3, 5, 7 (at.% were arc-melted into small ingots in an argon atmosphere. The characteristics of the macrostructures and microstructures were studied using a linear intercept method, OM, SEM (BSE, XRD and TEM. The results showed that the ingots with Hf have near lamellar microstructure in columnar and dendrite morphology. The hafnium concentration has a strong effect on the columnar spacing refi nement. Increasing Hf from 0 to 7 (at.%, the columnar spacing can be reduced from ~ 1000 to ~ 400 μm. Constitute phases of the ingots are α2, a small amount of B2 and c. Most of the B2 phases, richer in Hf and leaner in Al and Ti, exist on the node of the dendrite core in block shape and a little across the lamellar colonies in stick shape. The c phases exist on the boundaries of lamellar colonies in small cellular shape. There also exists a segregation of Hf on the columnar and dendrite core. Particularly, both the α- and β-phase form from the melt as prior phases. The possible phase sequencing during solidifi cation and solid-state transformations with Hf is given in this paper.

  2. Thermal behaviour of hafnium diethylenetriaminepentaacetate studied using the perturbed angular correlation technique

    Energy Technology Data Exchange (ETDEWEB)

    Chain, Cecilia Y. [Universidad Nacional de La Plata (Argentina). Dept. de Fisica; Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET), La Plata (Argentina). IFLP-CCT; Consejo Nacional de Investigaciones Cientificas y Tecnicas, Buenos Aires (Argentina); Rivas, Patricia [Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET), La Plata (Argentina). IFLP-CCT; Consejo Nacional de Investigaciones Cientificas y Tecnicas, Buenos Aires (Argentina); Universidad Nacional de La Plata (Argentina). Facultad de Ciencias Agrarias y Forestales; Pasquevich, Alberto F. [Universidad Nacional de La Plata (Argentina). Dept. de Fisica; Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET), La Plata (Argentina). IFLP-CCT; Comision de Investigaciones Cientificas de la Provincia de Buenos Aires (CIC-PBA) (Argentina)

    2014-07-01

    Polyaminecarboxilic ligands like diethylenetriaminepentaacetic acid form stable complexes with many heavy metal ions, excelling as cation chelants especially in the field of radiopharmacy. The aim of this work is to characterize, by using the Time Differential Perturbed Angular Correlations (TDPAC) technique, the hyperfine interactions at hafnium sites in hafnium diethylenetriaminepentaacetate and to investigate their evolution as temperature increases. TDPAC results for KHfDTPA.3H{sub 2}O obtained by chemical synthesis yield a well defined and highly asymmetric interaction of quadrupole frequency ω{sub Q} = 141 Mrad/s, which is consistent with the existence of a unique site for the metal in the crystal lattice. The thermal behaviour of the chelate is investigated by means of differential scanning calorimetry and thermogravimetrical analyses revealing that an endothermic dehydration of KHfDTPA.3H{sub 2}O takes place in one step between 80 C and 180 C. The anhydrous KHfDTPA thus arising is characterized by a fully asymmetric and well defined interaction of quadrupole frequency ω{sub Q} = 168 Mrad/s. (orig.)

  3. Molecular structure, vibrational, HOMO-LUMO, MEP and NBO analysis of hafnium selenite

    Science.gov (United States)

    Yankova, Rumyana; Genieva, Svetlana; Dimitrova, Ginka

    2017-08-01

    In hydrothermal condition hafnium selenite with estimated chemical composition Hf(SeO3)2·n(H2O) was obtained and characterized by powder X-Ray diffraction, IR spectroscopy and thermogravimetrical analysis. The composition of the obtained crystalline phase was established as dihydrate of tetraaqua complex of the hafnium selenite [Hf(SeO3)2(H2O)4]·2H2O. The results of the thermogravimetrical analysis are shown that the two hydrated water molecules are released in the temperature interval 80-110°C, while the four coordinated water molecules - at 210-300°C. By DFT method, with Becke's three parameter exchange-functional combined with gradient-corrected correlation functional of Lee, Yang and Parr and 6-31G(d), 6-311 + G(d,p) basis sets and LANL2DZ for Hf atom were calculated the molecular structure, vibrational frequencies and thermodynamic properties of the structure. The UV-Vis spectra and electronic properties are presented. The energy and oscillator strength calculated by time-dependent density functional theory corresponds well with the experimental ones. Molecular electrostatic potential (MEP) was performed. Mulliken population analysis on atomic charges was also calculated. The stability and intramolecular interactions are interpreted by NBO analysis.

  4. Compositional analysis of polycrystalline hafnium oxide thin films by heavy-ion elastic recoil detection analysis

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, F.L. [Departamento de Electronica y Tecnologia de Computadoras, Universidad Politecnica de Cartagena, Campus Universitario Muralla del Mar, E-30202 Cartagena (Spain)]. E-mail: Felix.Martinez@upct.es; Toledano, M. [Departamento de Fisica Aplicada III, Universidad Complutense de Madrid, E-28025 Madrid (Spain); San Andres, E. [Departamento de Fisica Aplicada III, Universidad Complutense de Madrid, E-28025 Madrid (Spain); Martil, I. [Departamento de Fisica Aplicada III, Universidad Complutense de Madrid, E-28025 Madrid (Spain); Gonzalez-Diaz, G. [Departamento de Fisica Aplicada III, Universidad Complutense de Madrid, E-28025 Madrid (Spain); Bohne, W. [Hahn-Meitner-Institut Berlin, Abteilung SF-4, D-14109 Berlin (Germany); Roehrich, J. [Hahn-Meitner-Institut Berlin, Abteilung SF-4, D-14109 Berlin (Germany); Strub, E. [Hahn-Meitner-Institut Berlin, Abteilung SF-4, D-14109 Berlin (Germany)

    2006-10-25

    The composition of polycrystalline hafnium oxide thin films has been measured by heavy-ion elastic recoil detection analysis (HI-ERDA). The films were deposited by high-pressure reactive sputtering (HPRS) on silicon wafers using an oxygen plasma at pressures between 0.8 and 1.6 mbar and during deposition times between 0.5 and 3.0 h. Hydrogen was found to be the main impurity and its concentration increased with deposition pressure. The composition was always slightly oxygen-rich, which is attributed to the oxygen plasma. Additionally, an interfacial silicon oxide thin layer was detected and taken into account. The thickness of the hafnium oxide film was found to increase linearly with deposition time and to decrease exponentially with deposition pressure, whereas the thickness of the silicon oxide interfacial layer has a minimum as a function of pressure at around 1.2 mbar and increases slightly as a function of time. The measurements confirmed that this interfacial layer is formed mainly during the early stages of the deposition process.

  5. High density fuels using dispersion and monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br, E-mail: alfredo@ctmsp.mar.mil.br, E-mail: rafael.orm@gmail.com, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Departamento de Engenharia Naval e Oceânica

    2017-07-01

    Fuel plates used in high-performance research reactors need to be converted to low-enrichment uranium fuel; the fuel option based on a monolithic formulation requires alloys to contain 6 - 10 wt% Mo. In this case, the fuel plates are composed of the metallic alloy U-10Mo surrounded by a thin zirconium layer encapsulated in aluminum cladding. This study reviewed the physical properties of monolithic forms. The constraints produced during the manufacturing process were analyzed and compared to those of dispersed fuel. The bonding process used for dispersion fuels differs from the techniques applied to foil bonding used for pure alloys. The quality of monolithic plates depends on the fabrication method, which usually involves hot isostatic pressing and the thermal annealing effect of residual stress, which degrades the uranium cubic phase. The preservation of the metastable phase has considerable influence on fuel performance. The physical properties of the foil fuel under irradiation are superior to those of aluminum-dispersed fuels. The fuel meat, using zirconium as the diffusion barrier, prevents the interaction layer from becoming excessively thick. The problem with dispersed fuel is breakaway swelling with a medium fission rate. It has been observed that the fuel dispersed in aluminum was minimized in monolithic forms. The pure alloys exhibited a suitable response from a rate at least twice as much as the fission rate of dispersions. The foils can support fissile material concentration combined with a reduced swelling rate. (author)

  6. Hafnium(IV) complexation with oxalate at variable temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Friend, Mitchell T.; Wall, Nathalie A. [Washington State Univ., Pullmanm, WA (United States). Dept. of Chemistry

    2017-08-01

    Appropriate management of fission products in the reprocessing of spent nuclear fuel (SNF) is crucial in developing advanced reprocessing schemes. The addition of aqueous phase complexing agents can prevent the co-extraction of these fission products. A solvent extraction technique was used to study the complexation of Hf(IV) - an analog to fission product Zr(IV) - with oxalate at 15, 25, and 35 C in 1 M HClO{sub 4} utilizing a {sup 175+181}Hf radiotracer. The mechanism of the solvent extraction system of 10{sup -5} M Hf(IV) in 1 M HClO{sub 4} to thenoyltrifluoroacetone (TTA) in toluene demonstrated a 4{sup th}-power dependence in both TTA and H{sup +}, with Hf(TTA){sub 4} the only extractable species. The equilibrium constant for the extraction of Hf(TTA){sub 4} was determined to be log K{sub ex}=7.67±0.07 (25±1 C, 1 M HClO{sub 4}). The addition of oxalate to the aqueous phase decreased the distribution ratio, indicating aqueous Hf(IV)-oxalate complex formation. Polynomial fits to the distribution data identified the formation of Hf(ox){sup 2+} and Hf(ox){sub 2(aq)} and their stability constants were measured at 15, 25, and 35 C in 1 M HClO{sub 4}. van't Hoff analysis was used to calculate Δ{sub r}G, Δ{sub r}H, and Δ{sub r}S for these species. Stability constants were observed to increase at higher temperature, an indication that Hf(IV)-oxalate complexation is endothermic and driven by entropy.

  7. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  8. Fuel safety research 2001

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-11-01

    The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

  9. Cladding Alloys for Fluoride Salt Compatibility Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Santella, Michael L [ORNL; Holcomb, David Eugene [ORNL

    2011-05-01

    This interim report provides an overview of several candidate technologies for cladding nickel-based corrosion protection layers onto high-temperature structural alloys. The report also provides a brief overview of the welding and weld performance issues associated with joining nickel-clad nickel-based alloys. From the available techniques, two cladding technologies were selected for initial evaluation. The first technique is a line-of-sight method that would be useful for coating large structures such as vessel interiors or large piping. The line-of-sight method is a laser-based surface cladding technique in which a high-purity nickel powder mixed into a polymer binder is first sprayed onto the surface, baked, and then rapidly melted using a high power laser. The second technique is a vapor phase technique based on the nickel-carbonyl process that is suitable for coating inaccessible surfaces such as the interior surfaces of heat exchangers. The final project report will feature an experimental evaluation of the performance of the two selected cladding techniques.

  10. Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

    Science.gov (United States)

    Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

    2016-11-01

    Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

  11. Spent fuel and fuel pool component integrity. Annual report, FY 1980

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.; Bailey, W.J.; Bradley, E.R.; Bruemmer, S.M.; Langstaff, D.C.

    1981-09-01

    During program FY 1980 staff members of the Spent Fuel and Fuel Pool Component Integrity Program at Pacific Northwest Laboratory (PNL) completed the following major tasks: represented DOE on the international Behavior of Fuel Assemblies in Storage (BEFAST) Committee; the program manager, A.B. Johnson, Jr., participated in an International Survey of Water Reactor Spent Fuel Storage Experience, which was conducted jointly by the International Atomic Energy Agency (Vienna) and the Nuclear Energy Agency (Paris); provided written testimony and cross statement for the Proposed Rulemaking on Storage and Disposal of Nuclear Waste; acquired and began examination of the world's oldest pool-stored Zircaloy-clad fuel from the Shippingport reactor, stored approx. 21 years in deionized water; acquired and began examination of stainless-clad spent fuel from the Connecticut Yankee Reactor (PWR); negotiated for specimens from components stored in spent fuel pools at fuel storage facilities from the Savannah River Plant, Aiken, South Carolina, Zion (PWR) spent fuel pool, Zion, Illinois, and La Crosse (BWR) spent fuel pool, La Crosse, Wisconsin; planned for examinations in FY 81 of specimens from the three spent fuel pools; investigated a low-temperature stress corrosion cracking mechanism that developed in piping at a few PWR spent fuel pools. This report summarizes the results of these activities and investigations. Details are provided in the presentationsand publications generated under this program and summarized in Appendix A.

  12. IN-PILE PERFORMANCE OF HANA CLADDING TESTED IN HALDEN REACTOR

    Directory of Open Access Journals (Sweden)

    HYUN-GIL KIM

    2014-06-01

    Full Text Available An in-pile performance test of HANA claddings was conducted at up to 67 GWD/MTU in the Halden research reactor in Norway over a 6.5 year period. Four types of HANA claddings (HANA-3, HANA-4, HANA-5, and HANA-6 and a reference Zircaloy-4 cladding were used for the in-pile test. The evaluation parameters of the HANA claddings were the corrosion behavior, dimensional changes, hydrogen uptake, and tensile strength after the claddings were tested under the simulated operation conditions of a Korean commercial reactor. The oxide thickness ranged from 15 to 37 mm at a high flux region in the test rods, and all HANA claddings showed corrosion resistance superior to the Zircaloy-4 cladding. The creep-down rate of all HANA claddings was lower than that of the Zircaloy-4 cladding. In addition, the hydrogen content of the HANA claddings ranged from 54 to 96 wppm at the high heat flux region of the test rods, whereas the hydrogen content of the Zircaloy-4 cladding was 119 wppm. The tensile strength of the HANA and Zircaloy-4 claddings was similarly increased when compared to the un-irradiated claddings owing to the radiation-induced hardening.

  13. Characterization of spent fuel approved testing material---ATM-105

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

  14. Experimental and numerical investigation on cladding of corrosion-erosion resistant materials by a high power direct diode laser

    Science.gov (United States)

    Farahmand, Parisa

    In oil and gas industry, soil particles, crude oil, natural gas, particle-laden liquids, and seawater can carry various highly aggressive elements, which accelerate the material degradation of component surfaces by combination of slurry erosion, corrosion, and wear mechanisms. This material degradation results into the loss of mechanical properties such as strength, ductility, and impact strength; leading to detachment, delamination, cracking, and ultimately premature failure of components. Since the failure of high valued equipment needs considerable cost and time to be repaired or replaced, minimizing the tribological failure of equipment under aggressive environment has been gaining increased interest. It is widely recognized that effective management of degradation mechanisms will contribute towards the optimization of maintenance, monitoring, and inspection costs. The hardfacing techniques have been widely used to enhance the resistance of surfaces against degradation mechanisms. Applying a surface coating improves wear and corrosion resistance and ensures reliability and long-term performance of coated parts. A protective layer or barrier on the components avoids the direct mechanical and chemical contacts of tool surfaces with process media and will reduce the material loss and ultimately its failure. Laser cladding as an advanced hardfacing technique has been widely used for industrial applications in order to develop a protective coating with desired material properties. During the laser cladding, coating material is fused into the base material by means of a laser beam in order to rebuild a damaged part's surface or to enhance its surface function. In the hardfacing techniques such as atmospheric plasma spraying (APS), high velocity oxygen-fuel (HVOF), and laser cladding, mixing of coating materials with underneath surface has to be minimized in order to utilize the properties of the coating material most effectively. In this regard, laser cladding offers

  15. Advanced Fuels Campaign FY 2010 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Flux Density through Guides with Microstructured Twisted Clad DB Medium

    Directory of Open Access Journals (Sweden)

    M. A. Baqir

    2014-01-01

    Full Text Available The paper deals with the study of flux density through a newly proposed twisted clad guide containing DB medium. The inner core and the outer clad sections are usual dielectrics, and the introduced twisted windings at the core-clad interface are treated under DB boundary conditions. The pitch angle of twist is supposed to greatly contribute towards the control over the dispersion characteristics of the guide. The eigenvalue equation for the guiding structure is deduced, and the analytical investigations are made to explore the propagation patterns of flux densities corresponding to the sustained low-order hybrid modes under the situation of varying pitch angles. The emphasis has been put on the effects due to the DB twisted pitch on the propagation of energy flux density through the guide.

  17. Low-Stress Silicon Cladding for Surface Finishing Large UVOIR Mirrors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this Phase I research, ZeCoat Corporation demonstrated a low-stress silicon cladding process for surface finishing large UVOIR mirrors. A polishable cladding is...

  18. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    Science.gov (United States)

    Lemmens, K.; González-Robles, E.; Kienzler, B.; Curti, E.; Serrano-Purroy, D.; Sureda, R.; Martínez-Torrents, A.; Roth, O.; Slonszki, E.; Mennecart, T.; Günther-Leopold, I.; Hózer, Z.

    2017-02-01

    The instant release of fission products from high burn-up UO2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45-63 GWd/tHM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride - bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways.

  19. Preliminary Modeling of Accident Tolerant Fuel Concepts under Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle A.; Hales, Jason D.

    2016-12-01

    The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. Thus, the United States Department of Energy through its NEAMS (Nuclear Energy Advanced Modeling and Simulation) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is funded for a three-year period. The purpose of the HIP is to perform research into two potential accident tolerant concepts and provide an in-depth report to the Advanced Fuels Campaign (AFC) describing the behavior of the concepts, both of which are being considered for inclusion in a lead test assembly scheduled for placement into a commercial reactor in 2022. The initial focus of the HIP is on uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (INL, LANL, and ANL) a comprehensive mulitscale approach to modeling is being used including atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. In this paper, we present simulations of two proposed accident tolerant fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. The simulations investigate the fuel performance response of the proposed ATF systems under Loss of Coolant and Station Blackout conditions using the BISON code. Sensitivity analyses are completed using Sandia National Laboratories’ DAKOTA software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). Early results indicate that each concept has significant advantages as well as areas of concern. Further work is required prior to formulating the proposition report for the Advanced Fuels Campaign.

  20. Uranium nitride fuel fabrication for SP-100 reactors

    Science.gov (United States)

    Mason, Richard E.; Chidester, Kenneth M.; Hoth, Carl W.; Matthews, Bruce R.

    Fuel pins of uranium mononitride clad in Nb-1 percent Zr were fabricated for irradiation tests in EBR-II. Laboratory scale process parameters to synthesize UN powders and fabricate UN pellets were developed. Uranium mononitride was prepared by converting UO2 to UN. Fuel pellets were prepared by communition of UN briquettes, uniaxial pressing, and high temperature sintering. Techniques for machining, cleaning, and welding Nb-1 percent Zr cladding components were developed. End caps were electron beam welded to the tubing. Helium back-fill holes were sealed with a laser weld.

  1. Integrated Computational Modeling of Water Side Corrosion in Zirconium Metal Clad Under Nominal LWR Operating Conditions

    Science.gov (United States)

    Aryanfar, Asghar; Thomas, John; Van der Ven, Anton; Xu, Donghua; Youssef, Mostafa; Yang, Jing; Yildiz, Bilge; Marian, Jaime

    2016-11-01

    A mesoscopic chemical reaction kinetics model to predict the formation of zirconium oxide and hydride accumulation light-water reactor (LWR) fuel clad is presented. The model is designed to include thermodynamic information from ab initio electronic structure methods as well as parametric information in terms of diffusion coefficients, thermal conductivities and reaction constants. In contrast to approaches where the experimentally observed time exponents are captured by the models by design, our approach is designed to be predictive and to provide an improved understanding of the corrosion process. We calculate the time evolution of the oxide/metal interface and evaluate the order of the chemical reactions that are conducive to a t 1/3 dependence. We also show calculations of hydrogen cluster accumulation as a function of temperature and depth using spatially dependent cluster dynamics. Strategies to further cohesively integrate the different elements of the model are provided.

  2. Effect of bundle size on cladding deformation in LOCA simulation tests. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, R.H.; Crowley, J.L.; Longest, A.W.

    1982-01-01

    Two LOCA simulation tests were conducted to investigate the effects of temperature uniformity and radial restraint boundary conditions on Zircaloy cladding deformation. In one of the tests (B-5), boundary conditions typical of a large array were imposed on an inner 4 x 4 square array by two concentric rings of interacting guard fuel pin simulators. In the other test (B-3), the boundary conditions were imposed on a 4 x 4 square array by a non-interacting heated shroud. Test parameters conducive to large deformation were selected in order to favor rod-to-rod interactions. The tests showed that rod-to-rod interactions play an important role in the deformation process.

  3. In-situ crack repair by laser cladding

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2010-09-01

    Full Text Available . Successful sealing of simulated cracks was achieved under extreme conditions. Not only was water squirting out occasionally, but water was continually running down the peened crack to the weld pool. Laser cladding was shown to be a robust process under... to the simulated crack prior to the cladding of the first layer to avoid excessive water squirting out. Adequate laser power is required to ensure proper fusion into the base material due to water squirting out of the crack adjacent to the weld pool. As a...

  4. Steady-state fuel behavior modeling of nitride fuels in FRAPCON-EP

    Science.gov (United States)

    Feng, Bo; Karahan, Aydın; Kazimi, Mujid S.

    2012-08-01

    Fuel material properties and mechanistic fission gas models in FRAPCON-EP were updated to model the steady-state behavior of high-porosity nitride fuel operating at temperatures below half of the melting point. The fuel thermal conductivity and fuel thermal expansion models were updated with correlations for UN and (U,Pu)N fuels. Hot-pressing of the as-fabricated porosity was modeled as a function of the hydrostatic pressure and creep rate. The solid fission product swelling was assumed to increase linearly with burnup. Fission gas swelling constitutive models were updated to appropriately capture the intragranular gas bubble evolution in nitride fuel. Intergranular gas swelling was neglected due to the assumed high porosity of the fuel. The fission gas release behavior was modeled by fitting the fission gas diffusion coefficient in UN to FRAPCON's default fission gas release model. This fitted gas diffusion coefficient reflects the effects of porosity, burnup, operating temperature, fission rate, and bubble sink strength. Fission gas release and fuel swelling benchmarks against irradiation data were performed. The updated code was applied to UN fuel in typical PWR geometry and operating conditions, with an extended cycle length of 24 months. The results show that swelling of the nitride fuel up to 60 MWd/kg burnup did not lead to excessive straining of the cladding. Furthermore, this study showed that a porous (>15% porosity) nitride fuel pellet could achieve a much higher margin to failure from the cladding collapse and grid-to-rod fretting.

  5. Fuel flexible fuel injector

    Science.gov (United States)

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  6. White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Louthan, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); PNNL, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-05-29

    This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history, residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed to

  7. 78 FR 9676 - Clad Steel Plate From Japan: Continuation of Antidumping Duty Order

    Science.gov (United States)

    2013-02-11

    ... them from products metalized in other manners (e.g., by normal electroplating). The various cladding... welding (e.g., electrocladding), in which the cladding metal (nickel, chromium, etc.) is applied to the...). Stainless clad steel plate is manufactured to American Society for Testing and Materials (ASTM...

  8. Gradient microstructure in laser clad TiC-reinforced Ni-alloy composite coating

    NARCIS (Netherlands)

    Pei, Y.T.; Zuo, T.C.

    1998-01-01

    A gradient TiC–(Ni alloy) composite coating was produced by one step laser cladding with pre-placed mixture powder on a 1045 steel substrate. The clad layers consisted of TiC particles, γ-Ni primary dendrites and interdendritic eutectics. From the bottom to the top of the clad layer produced at 2000

  9. Behavior of spent nuclear fuel and storage system components in dry interim storage. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-02-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom; organic-cooled reactor (OCR) fuel (clad with a zirconium alloy) in silos in Canada; and boiling water reactor (BWR) fuel (clad with Zircaloy) in a metal storage cask in Germany. Dry storage demonstrations are under way for Zircaloy-clad fuel from BWRs, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions. 110 refs., 22 figs., 28 tabs.

  10. MATPRO: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, P.E.; Thompson, L.B. (eds.)

    1976-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Documentation and formulations that are generally semiempirical in nature are presented for uranium dioxide and mixed uranium-plutonium dioxide fuel, zircaloy cladding, gas mixture, and LWR fuel rod material properties.

  11. Results of High-Temperature Heating Test for Irradiated Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, June-Hyung; Cheon, Jin-Sik; Lee, Byoung-Oon; Kim, Jun-Hwan; Kim, Hee-Moon; Yoo, Boung-Ok; Jung, Yang-Hong; Ahn, Sang-Bok; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The U and Pu constituents in the fuel, however, tend to interact metallurgically with iron-based claddings at elevated temperatures during nominal steady-state operating conditions and off-normal reactor events. In particular, if the temperature is raised above the eutectic temperature of metallic fuel, e.g., in an off-normal reactor event, the fuel can form a mixture of liquid and solid phases that may promote further cladding interaction. Such fuel-cladding chemical interaction, in conjunction with fission gas pressure loading, can potentially shorten fuel pin lifetime and eventually cause cladding breach. In this work, microstructure observation results through microscope, SEM and EPMA are reported for the irradiated U-10Zr and U-10Zr-5Ce fuel slugs with T92 cladding after high-temperature heating test. Also, the measured eutectic penetration rate is compared with the prediction value by the existing eutectic penetration correlation being used for design and modelling purposes. Microstructure of the irradiated U-10Zr and U-10Zr-5Ce fuel slug with T92 cladding after high-temperature heating test were investigated through the microscope, SEM and EPMA. Also, the measured maximum eutectic penetration rate along cladding direction was compared with the prediction value by existing eutectic penetration correlation. In the case of U-10Zr/T92 specimen, migration phenomena of U, Zr, and Fe as well as Nd lanthanide fission product were observed at the eutectic melting region. The measured penetration rate was almost similar to prediction value by existing eutectic penetration rate correlation.

  12. Accident Tolerant Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; Heather Chichester; Jesse Johns; Melissa Teague; Michael Tonks; Robert Youngblood

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and

  13. IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

    Directory of Open Access Journals (Sweden)

    M.K. MEYER

    2014-04-01

    Full Text Available High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  14. Multiphonon ionization of traps formed in hafnium oxide by electrical stress

    Energy Technology Data Exchange (ETDEWEB)

    Danilyuk, A.L.; Migas, D.B.; Danilyuk, M.A.; Borisenko, V.E. [Belorussian State University of Informatics and Radioelectronics, P. Browka 6, 220013 Minsk (Belarus); Wu, X.; Pey, K.L. [Microelectronics Center, School of EEE, Nanyang Technological University, Nanyang Avenue, Singapore 639798 (Singapore); Concurrently with Engineering Product Development Pillar, Singapore University of Technology and Design, 20 Dover Drive, Singapore 138682 (Singapore); Raghavan, N. [Microelectronics Center, School of EEE, Nanyang Technological University, Nanyang Avenue, Singapore 639798 (Singapore)

    2013-02-15

    We have investigated behavior of traps formed in hafnium oxide (HfO{sub 2}) by electrical stress and their influence on the charge carrier transport through Si/SiO{sub 2}/HfO{sub 2}/poly-Si nanostructures. The traps govern the transport process assuming a capture of charge carriers followed by their ionization via the multiphonon transition mechanism. The multiphonon transitions via the Poole-Frenkel effect or electron tunneling as well as the multiphonon tunneling ionization of neutral traps have been carefully considered for charged traps. We also provide a set of parameters including the trap concentration, ionization energy, the frequency factor, the effective mass of charge carriers, optical energy, and phonon energy in order to reproduce and reasonably fit available experimental data. (Copyright copyright 2013 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  15. Specific features of the charge and mass transfer in a silver-intercalated hafnium diselenide

    Science.gov (United States)

    Pleshchev, V. G.; Selezneva, N. V.; Baranov, N. V.

    2013-07-01

    The specific features of the charge transfer in intercalated samples of Ag x HfSe2 have been studied for the first time by alternating current (ac) impedance spectroscopy. It has been found that relaxation processes in an ac field are accelerated with increasing silver content in the samples. The complex conductivity ( Y) shows a frequency dispersion described by power law Y ˜ ω s , which is characteristic of the hopping conductivity mechanism. The Ag x HfSe2 compounds demonstrate shorter relaxation times as compared to those observed in hafnium diselenide intercalated with copper atoms, and this fact indicates that the charge carrier mobility in the silver-intercalated compounds is higher. The possibility of silver ion transfer in Ag x HfSe2 is confirmed by the measurements performed by the method of electrochemical cell emf.

  16. Surface State Capture Cross-Section at the Interface between Silicon and Hafnium Oxide

    Directory of Open Access Journals (Sweden)

    Fu-Chien Chiu

    2013-01-01

    Full Text Available The interfacial properties between silicon and hafnium oxide (HfO2 are explored by the gated-diode method and the subthreshold measurement. The density of interface-trapped charges, the current induced by surface defect centers, the surface recombination velocity, and the surface state capture cross-section are obtained in this work. Among the interfacial properties, the surface state capture cross-section is approximately constant even if the postdeposition annealing condition is changed. This effective capture cross-section of surface states is about 2.4 × 10−15 cm2, which may be an inherent nature in the HfO2/Si interface.

  17. Dinitrogen cleavage and functionalization by carbon monoxide promoted by a hafnium complex.

    Science.gov (United States)

    Knobloch, Donald J; Lobkovsky, Emil; Chirik, Paul J

    2010-01-01

    Molecular nitrogen (N(2)) and carbon monoxide (CO) have the two strongest bonds in chemistry and present significant challenges in developing new transformations that exploit these two abundant feedstocks. At the core of this objective is the discovery of transition-metal compounds that promote the six-electron reductive cleavage of N(2) at ambient temperature and pressure and also promote new nitrogen-element bond formation. Here we show that an organometallic hafnium compound induces N(2) cleavage on the addition of CO, with a simultaneous assembly of new nitrogen-carbon and carbon-carbon bonds. Subsequent addition of a weak acid liberates oxamide, which demonstrates that an important agrochemical can be synthesized directly from N(2) and CO. These studies introduce an alternative paradigm for N(2) cleavage and functionalization in which the six-electron reductive cleavage is promoted by both the transition metal and the incoming ligand, CO, used for the new bond formations.

  18. Hafnium dioxide as a dielectric for highly-sensitive waveguide-coupled surface plasmon resonance sensors

    Directory of Open Access Journals (Sweden)

    Kunal Tiwari

    2016-04-01

    Full Text Available Hafnium dioxide has been recognized as an excellent dielectric for microelectronics. However, its usefulness for the surface plasmon based sensors has not yet been tested. Here we investigate its usefulness for waveguide-coupled bi-metallic surface plasmon resonance sensors. Several Ag/HfO2/Au multilayer structure sensors were fabricated and evaluated by optical measurements and computer simulations. The resulting data establish correlations between the growth parameters and sensor performance. The sensor sensitivity to refractive index of analytes is determined to be S n = ∂ θ SPR ∂ n ≥ 4 7 0 . The sensitivity data are supported by simulations, which also predict 314 nm for the evanescent field decay length in air.

  19. Zirconium and hafnium fractionation in differentiation of alkali carbonatite magmatic systems

    Science.gov (United States)

    Kogarko, L. N.

    2016-05-01

    Zirconium and hafnium are valuable strategic metals which are in high demand in industry. The Zr and Hf contents are elevated in the final products of magmatic differentiation of alkali carbonatite rocks in the Polar Siberia region (Guli Complex) and Ukraine (Chernigov Massif). Early pyroxene fractionation led to an increase in the Zr/Hf ratio in the evolution of the ultramafic-alkali magmatic system due to a higher distribution coefficient of Hf in pyroxene with respect to Zr. The Rayleigh equation was used to calculate a quantitative model of variation in the Zr/Hf ratio in the development of the Guli magmatic system. Alkali carbonatite rocks originated from rare element-rich mantle reservoirs, in particular, the metasomatized mantle. Carbonated mantle xenoliths are characterized by a high Zr/Hf ratio due to clinopyroxene development during metasomatic replacement of orthopyroxene by carbonate fluid melt.

  20. Pyroelectric response in crystalline hafnium zirconium oxide (Hf1-xZrxO2) thin films

    Science.gov (United States)

    Smith, S. W.; Kitahara, A. R.; Rodriguez, M. A.; Henry, M. D.; Brumbach, M. T.; Ihlefeld, J. F.

    2017-02-01

    Pyroelectric coefficients were measured for 20 nm thick crystalline hafnium zirconium oxide (Hf1-xZrxO2) thin films across a composition range of 0 ≤ x ≤ 1. Pyroelectric currents were collected near room temperature under zero applied bias and a sinusoidal oscillating temperature profile to separate the influence of non-pyroelectric currents. The pyroelectric coefficient was observed to correlate with zirconium content, increased orthorhombic/tetragonal phase content, and maximum polarization response. The largest measured absolute value was 48 μCm-2 K-1 for a composition with x = 0.64, while no pyroelectric response was measured for compositions which displayed no remanent polarization (x = 0, 0.91, and 1).

  1. A thermally robust and thickness independent ferroelectric phase in laminated hafnium zirconium oxide

    Directory of Open Access Journals (Sweden)

    S. Riedel

    2016-09-01

    Full Text Available Ferroelectric properties in hafnium oxide based thin films have recovered the scaling potential for ferroelectric memories due to their ultra-thin-film- and CMOS-compatibility. However, the variety of physical phenomena connected to ferroelectricity allows a wider range of applications for these materials than ferroelectric memory. Especially mixed HfxZr1-xO2 thin films exhibit a broad compositional range of ferroelectric phase stability and provide the possibility to tailor material properties for multiple applications. Here it is shown that the limited thermal stability and thick-film capability of HfxZr1-xO2 can be overcome by a laminated approach using alumina interlayers.

  2. Cesium hafnium chloride: A high light yield, non-hygroscopic cubic crystal scintillator for gamma spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Burger, Arnold, E-mail: aburger@fisk.edu [Department of Life and Physical Sciences, Fisk University, Nashville, Tennessee 37208 (United States); Department of Physics and Astronomy, Vanderbilt University, Nashville, Tennessee 37235 (United States); Rowe, Emmanuel; Groza, Michael; Morales Figueroa, Kristle [Department of Life and Physical Sciences, Fisk University, Nashville, Tennessee 37208 (United States); Cherepy, Nerine J.; Beck, Patrick R.; Hunter, Steven; Payne, Stephen A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

    2015-10-05

    We report on the scintillation properties of Cs{sub 2}HfCl{sub 6} (cesium hafnium chloride or CHC) as an example of a little-known class of non-hygroscopic compounds having the generic cubic crystal structure of K{sub 2}PtCl{sub 6}. The crystals are easily growable from the melt using the Bridgman method with minimal precursor treatments or purification. CHC scintillation is centered at 400 nm, with a principal decay time of 4.37 μs and a light yield of up to 54 000 photons/MeV when measured using a silicon CCD photodetector. The light yield is the highest ever reported for an undoped crystal, and CHC also exhibits excellent light yield nonproportionality. These desirable properties allowed us to build and test CHC gamma-ray spectrometers providing energy resolution of 3.3% at 662 keV.

  3. Separation of hafnium from zirconium in their tetrachloride solution in molten alkali metal chlorides

    Energy Technology Data Exchange (ETDEWEB)

    Salyulev, A.B.; Kudyakov, V.Ya.; Smirnov, M.V.; Moskalenko, N.I. (AN SSSR, Sverdlovsk. Inst. Ehlektrokhimii)

    1984-08-01

    The coefficient of HfCl/sub 4/ and ZrCl/sub 4/ separation in the process of vapour sublimation from their solutions in molten NaCl, KCl, CsCl, NaCl-KCl and NaCl-CsCl equimolar mixtures is found to vary in the series from approximately 1.10 to approximately 1.22 and practically not to depend on the temperature (in the 600-910 deg) range and concentration (2-25 mol.% ZrCl/sub 4/+HfCl/sub 4/). HfCl/sub 4/ and ZrCl/sub 4/ are shown to form almost perfect solutions with each other, which in their turn form imperfect solutions with molten alkali metal chlorides, with the strength of hafnium complex chloride anions increasing higher than that of zirconium in the series from NaCl to CsCl.

  4. Selected Isotopes for Optimized Fuel Assembly Tags

    Energy Technology Data Exchange (ETDEWEB)

    Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

    2008-10-01

    In support of our ongoing signatures project we present information on 3 isotopes selected for possible application in optimized tags that could be applied to fuel assemblies to provide an objective measure of burnup. 1. Important factors for an optimized tag are compatibility with the reactor environment (corrosion resistance), low radioactive activation, at least 2 stable isotopes, moderate neutron absorption cross-section, which gives significant changes in isotope ratios over typical fuel assembly irradiation levels, and ease of measurement in the SIMS machine 2. From the candidate isotopes presented in the 3rd FY 08 Quarterly Report, the most promising appear to be Titanium, Hafnium, and Platinum. The other candidate isotopes (Iron, Tungsten, exhibited inadequate corrosion resistance and/or had neutron capture cross-sections either too high or too low for the burnup range of interest.

  5. Characterization of a Novel Hafnium-Based X-ray Contrast Agent.

    Science.gov (United States)

    Frenzel, Thomas; Bauser, Marcus; Berger, Markus; Hilger, Christoph Stephan; Hegele-Hartung, Christa; Jost, Gregor; Neis, Christian; Hegetschweiler, Kaspar; Riefke, Björn; Suelzle, Detlev; Pietsch, Hubertus

    2016-12-01

    Characterization of BAY-576, a new x-ray contrast agent which is not based on iodine, but rather on the heavy metal hafnium. Compared with iodine, hafnium provides better x-ray absorption in the energy range of computed tomography (CT) and allows images of comparable quality to be acquired at a significantly reduced radiation dose. A range of standard methods were used to explore the physicochemistry of BAY-576 as well as its tolerability in in vitro assays, its pharmacokinetics and toxicology in rats, and its performance in CT imaging in rabbits. BAY-576 is an extraordinarily stable chelate with a metal content of 42% (wt/wt) and with excellent water solubility. Formulations of 300 mg Hf/mL exhibited viscosity (3.3-3.6 mPa) and osmolality (860-985 mOsm/kg) in the range of nonionic x-ray agents. No relevant effects on erythrocytes, the coagulation, or complement system or on a panel of 87 potential biological targets were observed. The compound did not bind to plasma proteins of a number of species investigated. After intravenous injection in rats, it was excreted fast and mainly via the kidneys. Its pharmacokinetics was comparable to known extracellular contrast agents. A dose of 6000 mg Hf/kg, approximately 10 to 20 times the expected diagnostic dose, was well tolerated by rats with only moderate adverse effects. Computed tomography imaging in rabbits bearing a tumor in the liver demonstrated excellent image quality when compared with iopromide at the same contrast agent dose in angiography during the arterial phase. At 70% of the radiation dose, BAY-576 provided a contrast-to-noise ratio of the tumor, which was equivalent to iopromide at 100% radiation dose. The profile of BAY-576 indicates its potential as the first compound in a new class of noniodine x-ray contrast agents, which can contribute to the reduction of the radiation burden in contrast-enhanced CT imaging.

  6. Foam coating on aluminum alloy with laser cladding

    NARCIS (Netherlands)

    Ocelik, V.; van Heeswijk, V.; de Hosson, J.T.M.; Csach, K.

    dThis article concentrates on the creation of a foam layer on an Al-Si substrate with laser technology. The cladding of At-Si powder in the front of a laser track has been separated from the side injection of mixture of Al-Si/TiH2 powder (foaming agent), which allows for fine tuning of the main

  7. Interfacial adhesion of laser clad functionally graded materials

    NARCIS (Netherlands)

    Pei, Y. T.; Ocelik, V.; De Hosson, J. T. M.

    2003-01-01

    Specially designed samples of laser clad AlSi40 functionally graded materials (FGM) are made for evaluating the interfacial adhesion. To obtain the interfacial bond strength notches are made right at the interface of the FGMs. In-situ microstructural observations during straining in a field-emission

  8. Laser cladding process development for high carbon steel substrates

    CSIR Research Space (South Africa)

    Lengopeng, T

    2014-11-01

    Full Text Available spot diameter and powder feed rate were kept constant throughout the experiment. The effect of multiple layers (up to a total of six clad layers built-up) and post-weld heat treatment by a defocussed laser beam on the HAZ width and hardness were...

  9. Widely tunable femtosecond solitonic radiation in photonic crystal fiber cladding

    DEFF Research Database (Denmark)

    Peng, J. H.; Sokolov, A. V.; Benabid, F.

    2010-01-01

    We report on a means to generate tunable ultrashort optical pulses. We demonstrate that dispersive waves generated by solitons within the small-core features of a photonic crystal fiber cladding can be used to obtain femtosecond pulses tunable over an octave-wide spectral range. The generation...

  10. Elimination of Start/Stop defects in laser cladding

    NARCIS (Netherlands)

    Ocelik, V.; Eekma, M.; Hemmati, I.; De Hosson, J. Th. M.

    2012-01-01

    Laser cladding represents an advanced hard facing technology for the deposition of hard, corrosion and wear resistant layers of controlled thickness onto a selected area of metallic substrate. When a circular geometry is required, the beginning and the end of the laser track coincide in the same

  11. Microstrain Determination in Individual Grains of Laser Deposited Cladding Layers

    NARCIS (Netherlands)

    de Oliveira, Uazir O. B.; Ocelik, Vaclav; De Hosson, Jeff T. M.; Chandra, T; Tsuzaki, K; Militzer, M; Ravindran, C

    2007-01-01

    The laser cladding technique makes the deposition of thick metallic, wear and corrosion resistant coatings feasible on weaker substrates. During the process, localized high thermal gradients generate internal stresses that may cause cracking when these overcome the fracture stress. To explain the

  12. Finite-width plasmonic waveguides with hyperbolic multilayer cladding

    DEFF Research Database (Denmark)

    Babicheva, Viktoriia; Shalaginov, Mikhail Y.; Ishii, Satoshi

    2015-01-01

    Engineering plasmonic metamaterials with anisotropic optical dispersion enables us to tailor the properties of metamaterial-based waveguides. We investigate plasmonic waveguides with dielectric cores and multilayer metal-dielectric claddings with hyperbolic dispersion. Without using any......, are strongly absorbed. By avoiding the resonant widths in the design of the actual waveguides, the strong absorption can be eliminated. (C) 2015 Optical Society of America...

  13. Ultrahigh temperature-sensitive silicon MZI with titania cladding

    Directory of Open Access Journals (Sweden)

    Jong-Moo eLee

    2015-05-01

    Full Text Available We present a possibility of intensifying temperature sensitivity of a silicon Mach-Zehnder interferometer (MZI by using a highly negative thermo-optic property of titania (TiO2. Temperature sensitivity of an asymmetric silicon MZI with a titania cladding is experimentally measured from +18pm/C to -340 pm/C depending on design parameters of MZI.

  14. Spatial mode-selective waveguide with hyperbolic cladding

    NARCIS (Netherlands)

    Tang, Y.; Xi, Z.; Xu, M.; Bäumer, S.M.B.; Adam, A.J.L.; Urbach, H.P.

    2016-01-01

    Hyperbolic metamaterials (HMMs) are anisotropic materials with a permittivity tensor that has both positive and negative eigenvalues. Here we report that by using a type II HMM as a cladding material, a waveguide that only supports higher-order modes can be achieved, while the lower-order modes

  15. Sodium fast reactor fuels and materials : research needs.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R.; Porter, Douglas (Idaho National Laboratory, Idaho Falls, ID); Wright, Art (Argonne National Laboratory Argonne, IL); Lambert, John (Argonne National Laboratory Argonne, IL); Hayes, Steven (Idaho National Laboratory, Idaho Falls, ID); Natesan, Ken (Argonne National Laboratory Argonne, IL); Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Garner, Frank (Radiation Effects Consulting. Richland, WA); Walters, Leon (Advanced Reactor Concepts, Idaho Falls, ID); Yacout, Abdellatif (Argonne National Laboratory Argonne, IL)

    2011-09-01

    An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

  16. Thermo-Elastic Finite Element Analyses of Annular Nuclear Fuels

    Science.gov (United States)

    Kwon, Y. D.; Kwon, S. B.; Rho, K. T.; Kim, M. S.; Song, H. J.

    In this study, we tried to examine the pros and cons of the annular type of fuel concerning mainly with the temperatures and stresses of pellet and cladding. The inner and outer gaps between pellet and cladding may play an important role on the temperature distribution and stress distribution of fuel system. Thus, we tested several inner and outer gap cases, and we evaluated the effect of gaps on fuel systems. We conducted thermo-elastic-plastic-creep analyses using an in-house thermo-elastic-plastic-creep finite element program that adopted the 'effective-stress-function' algorithm. Most analyses were conducted until the gaps disappeared; however, certain analyses lasted for 1582 days, after which the fuels were replaced. Further study on the optimal gaps sizes for annular nuclear fuel systems is still required.

  17. Lanthanides migration and immobilization in U-Zr nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bozzolo, G., E-mail: guille_bozzolo@yahoo.com [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Hofman, G.L.; Yacout, A.M. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Mosca, H.O. [Gerencia de Investigaciones y Aplicaciones, CNEA, Av. Gral Paz 1499, B165KNA, San Martin, Buenos Aires (Argentina)

    2012-06-15

    Redistribution of lanthanides fission products during irradiation and migration to the surface of U-Zr based metallic fuels is a concern due to their interaction with the cladding. The existing remedy for preventing this effect is the introduction of diffusion barriers on the cladding inner surface or by adding thermodynamically stable compound-forming elements to the fuel. Exploring this second option, in this work atomistic modeling with the Bozzolo-Ferrante-Smith (BFS) method for alloys is used to study the formation of lanthanide-rich precipitates in U-Zr fuel and the segregation patterns of all constituents to the surface. Surface energies for all elements were computed and, together with the underlying concepts of the computational methodology and large scale simulations, the migration of lanthanides to the surface region in U-Zr fuels is explained. The role of additions to the fuel such as In, Ga, and Tl for immobilization of lanthanides is discussed.

  18. Advanced Fuels Campaign FY 2014 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Braase, Lori [Idaho National Lab. (INL), Idaho Falls, ID (United States). INL Systems Analyses; May, W. Edgar [Idaho National Lab. (INL), Idaho Falls, ID (United States). INL Systems Analyses

    2014-10-01

    The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This includes development of a state-of-the art Research and Development (R&D) infrastructure to support the use of a “goal-oriented science-based approach.” In support of the Fuel Cycle Research and Development (FCRD) program, AFC is responsible for developing advanced fuels technologies to support the various fuel cycle options defined in the Department of Energy (DOE) Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. AFC uses a “goal-oriented, science-based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performance under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. The modeling and simulation activities for fuel performance are carried out under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, which is closely coordinated with AFC. In this report, the word “fuel” is used generically to include fuels, targets, and their associated cladding materials. R&D of light water reactor (LWR) fuels with enhanced accident tolerance is also conducted by AFC. These fuel systems are designed to achieve significantly higher fuel and plant performance to allow operation to significantly higher burnup, and to provide enhanced safety during design basis and beyond design basis accident conditions. The overarching goal is to develop advanced nuclear fuels and materials that are robust, have high performance capability, and are more tolerant to

  19. FAILED FUEL DISPOSITION STUDY

    Energy Technology Data Exchange (ETDEWEB)

    THIELGES, J.R.

    2004-12-20

    In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

  20. Final Report on TREAT Tests R4 and R5; Seven-pin, Loss-of-flow Tests with Full-length, Unirradiated FFTF-type Fuel Pins -- Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B. W.

    1979-04-01

    Two seven-pin loss-of-flow simulation tests have been performed in-pile in support of LMFBR safety analyses. These integral-type tests were executed in the R-series test apparatus at TREAT, designed to provide a prototypic thermal hydraulic flowing sodium system. The test fuel pins were full-length FFTF-type, containing unirradiated UO{sub 2} fuel. In LOF test R4, the sequence was run at constant fuel power well beyond the inception of molten fuel motion; in R5, the sequence was terminated prior to fuel melting to preserve evidence of early molten cladding motion. The tests were consistent with one another, showing the anticipated sequence of sodium boiling, channel voiding, cladding dryout, cladding failures, molten cladding motion and, for R4, fuel melting and motion. Thin planar upper cladding blockages were formed as were thick lower blockages, effectively plugging the bundle at the extremities of the active fuel region. There was no sodium reentry, and there were no significant pressurization events. Comparison of test results with SAS code calculations revealed differences associated with early voiding behavior, release of noncondensible fuel pin plenum gas, and the thickness and location of the upper blockage. In general, however, the overall sequence of significant events observed in R4 and R5 is in quite good agreement with SAS code analysis of the test conditions. Features of SAS related to overall voiding, cladding motion, and to some extent fuel motion have been substantiated in these results.

  1. Light water reactor fuel response during reactivity initiated accident experiments

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, P. E.; McCardell, R. K.; Martinson, Z. R.; Seiffert, S. L.

    1979-01-01

    Experimental results from six recent Power Burst Facility (PBF) reactivity initiated accident (RIA) tests are compared with data from previous Special Power Excursion Reactor Test (SPERT), and Japanese Nuclear Safety Research Reactor (NSRR) tests. The RIA fuel behavior experimental program recently started in the PBF is being conducted with coolant conditions typical of hot-startup conditions in a commercial boiling water reactor. The SPERT and NSRR test programs investigated the behavior of single or small clusters of light water reactor (LWR) type fuel rods under approximate room temperature and atmospheric pressure conditions in capsules containing stagnant water. As observed in the SPERT and NSRR tests, energy deposition, and consequent enthalpy increase in the PBF test fuel, appears to be the single most important variable. However, the consequences of failure at boiling water hot-startup system conditions appear to be more severe than previously observed in either the stagnant capsule SPERT or NSRR tests. Metallographic examination of both previously unirradiated and irradiated PBF fuel rod cross sections revealed extensive variation in cladding wall thicknesses (involving considerable plastic flow) and fuel shattering along grain boundaries in both restructured and unrestructured fuel regions. Oxidation of the cladding resulted in fracture at the location of cladding thinning and disintegration of the rods during quench. In addition,swelling of the gaseous and potentially volatile fission products in previously irradiated fuel resulted in volume increases of up to 180% and blockage of the coolant channels within the flow shrouds surrounding the fuel rods.

  2. Effect of Co - based Alloy on Properties of Laser Cladding Layer

    Science.gov (United States)

    Yang, Y.; Jiang, Z. P.; Li, H. Z.

    2017-11-01

    A large number of laser cladding experiments have been carried out using 20CrMnTi steel as substrate and Co-based alloy as cladding material. The influence of Co-based alloy on the laser cladding properties of 20CrMnTi steel was studied by analyzing the macroscopic and microscopic characteristics of cladding crack susceptibility, dilution rate, microstructure and friction and wear properties. The results show that the high-power laser cladding of Co-based material can obtain a flat defect-free cladding layer with compact structure and low crack susceptibility. A multi-layer cladding strategy with variable power can be used to fabricate thin wall structures without collapse Parts, the surface smooth without pores.

  3. Hafnium isotope evidence for a transition in the dynamics of continental growth 3.2 Gyr ago.

    Science.gov (United States)

    Næraa, T; Scherstén, A; Rosing, M T; Kemp, A I S; Hoffmann, J E; Kokfelt, T F; Whitehouse, M J

    2012-05-30

    Earth's lithosphere probably experienced an evolution towards the modern plate tectonic regime, owing to secular changes in mantle temperature. Radiogenic isotope variations are interpreted as evidence for the declining rates of continental crustal growth over time, with some estimates suggesting that over 70% of the present continental crustal reservoir was extracted by the end of the Archaean eon. Patterns of crustal growth and reworking in rocks younger than three billion years (Gyr) are thought to reflect the assembly and break-up of supercontinents by Wilson cycle processes and mark an important change in lithosphere dynamics. In southern West Greenland numerous studies have, however, argued for subduction settings and crust growth by arc accretion back to 3.8 Gyr ago, suggesting that modern-day tectonic regimes operated during the formation of the earliest crustal rock record. Here we report in situ uranium-lead, hafnium and oxygen isotope data from zircons of basement rocks in southern West Greenland across the critical time period during which modern-like tectonic regimes could have initiated. Our data show pronounced differences in the hafnium isotope-time patterns across this interval, requiring changes in the characteristics of the magmatic protolith. The observations suggest that 3.9-3.5-Gyr-old rocks differentiated from a >3.9-Gyr-old source reservoir with a chondritic to slightly depleted hafnium isotope composition. In contrast, rocks formed after 3.2 Gyr ago register the first additions of juvenile depleted material (that is, new mantle-derived crust) since 3.9 Gyr ago, and are characterized by striking shifts in hafnium isotope ratios similar to those shown by Phanerozoic subduction-related orogens. These data suggest a transitional period 3.5-3.2 Gyr ago from an ancient (3.9-3.5 Gyr old) crustal evolutionary regime unlike that of modern plate tectonics to a geodynamic setting after 3.2 Gyr ago that involved juvenile crust generation by plate

  4. Fuel failures in the Connecticut Yankee reactor (Haddam Neck). Addendum to NP-2119

    Energy Technology Data Exchange (ETDEWEB)

    Raven, L.F.A.; Howl, D.; Naylor, J.; Pitek, M.T.; Clink, L.J.

    1984-05-01

    Significant levels of fuel rod failures were observed in the batch 8 fuel assemblies of the Connecticut Yankee reactor. Results of detailed poolside and hot cell examinations, reported earlier in NP-2119, indicated the failure mechanism was stress corrosion cracking initiating on the clad outer surface. The sources of cladding stresses were believed to be (a) fuel pellet chips wedged in the cladding gap, (b) swelling of highly nondensifying batch 8 fuel, and (c) potentially harmful effects of a power change event that occurred near the end of the second cycle of irradiation for batch 8. This report reviews the 1977-78 experience and conclusions of the earlier investigations against the background of the total operational experience of the reactor from initial startup to the present time. It provides more details on operating conditions and uses the results of SLEUTH-SEER analyses to interpret the effects of operational maneuvers. The investigation adds more evidence for the importance of the ramp effects.

  5. Effect of the fuel element bundle statistical characteristics on the evaluation of temperature in the sodium-cooled fast-neutron reactor core

    Directory of Open Access Journals (Sweden)

    B.B. Tikhomirov

    2015-09-01

    Full Text Available Different fuel element bundle models used to calculate the coolant and fuel cladding temperatures inside fuel assemblies have been analyzed as applied to sodium-cooled fast-neutron reactors. The drawbacks of the existing models have been identified. A bundle model based on an experimental study into the actual arrangement of the fuel elements within the AF shroud has been proposed. The model's capabilities and advantages, as compared to conservative models, have been shown with regard for the need to raise the reliability of the fuel cladding working temperature estimation.

  6. Macroscopic behavior of fast reactor fuel subjected to simulated thermal transients

    Energy Technology Data Exchange (ETDEWEB)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.

    1983-06-01

    High-speed cinematography has been used to characterize the macroscopic behavior of irradiated and unirradiated fuel subjected to thermal transients prototypical of fast reactor transients. The results demonstrate that as the cladding melts, the fuel can disperse via spallation if the fuel contains in excess of approx. 16 ..mu..moles/gm of fission gas. Once the cladding has melted, the macroscopic behavior (time to failure and dispersive nature) was strongly influenced by the presence of volatile fission products and the heating rate.

  7. Fabrication of Tungsten-Rhenium Cladding materials via Spark Plasma Sintering for Ultra High Temperature Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Charit, Indrajit; Butt, Darryl; Frary, Megan; Carroll, Mark

    2012-11-05

    This research will develop an optimized, cost-effective method for producing high-purity tungsten-rhenium alloyed fuel clad forms that are crucial for the development of a very high-temperature nuclear reactor. The study will provide critical insight into the fundamental behavior (processing-microstructure- property correlations) of W-Re alloys made using this new fabrication process comprising high-energy ball milling (HEBM) and spark plasma sintering (SPS). A broader goal is to re-establish the U.S. lead in the research field of refractory alloys, such as W-Re systems, with potential applications in very high-temperature nuclear reactors. An essential long-term goal for nuclear power is to develop the capability of operating nuclear reactors at temperatures in excess of 1,000K. This capability has applications in space exploration and some special terrestrial uses where high temperatures are needed in certain chemical or reforming processes. Refractory alloys have been identified as being capable of withstanding temperatures in excess of 1,000K and are considered critical for the development of ultra hightemperature reactors. Tungsten alloys are known to possess extraordinary properties, such as excellent high-temperature capability, including the ability to resist leakage of fissile materials when used as a fuel clad. However, there are difficulties with the development of refractory alloys: 1) lack of basic experimental data on thermodynamics and mechanical and physical properties, and 2) challenges associated with processing these alloys.

  8. Aminopyridinate-FI hybrids, their hafnium and titanium complexes, and their application in the living polymerization of 1-hexene.

    Science.gov (United States)

    Haas, Isabelle; Dietel, Thomas; Press, Konstantin; Kol, Moshe; Kempe, Rhett

    2013-10-11

    Based on two well-established ligand systems, the aminopyridinato (Ap) and the phenoxyimine (FI) ligand systems, new Ap-FI hybrid ligands were developed. Four different Ap-FI hybrid ligands were synthesized through a simple condensation reaction and fully characterized. The reaction of hafnium tetrabenzyl with all four Ap-FI hybrid ligands exclusively led to mono(Ap-FI) complexes of the type [(Ap-FI)HfBn2 ]. The ligands acted as tetradentate dianionic chelates. Upon activation with tris(pentafluorophenyl)borane, the hafnium-dibenzyl complexes led to highly active catalysts for the polymerization of 1-hexene. Ultrahigh molecular weights and extremely narrow polydispersities support the living nature of this polymerization process. A possible deactivation product of the hafnium catalysts was characterized by single-crystal X-ray analysis and is discussed. The coordination modes of these new ligands were studied with the help of model titanium complexes. The reaction of titanium(IV) isopropoxide with ligand 1 led to a mono(Ap-FI) complex, which showed the desired fac-mer coordination mode. Titanium (IV) isopropoxide reacted with ligand 4 to give a complex of the type [(ApH-FI)2 Ti(OiPr)2 ], which featured the ligand in its monoanionic form. The two titanium complexes were characterized by X-ray crystal-structure analysis. Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  9. A study on the instrumentation technology for the irradiation of the DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chul Yong; Bae, K. K.; Moon, J. S.; Park, H. S.; Song, K. C.; Jung, I. H.; Kang, K. H.; Yang, M. S.; Ryu, J. S.; Cho, Y. G. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    The advanced countries have a great experience for the irradiation test of material and nuclear fuel from 1960's. Especially, instruments and instrument attachment technology and irradiation equipment have been developed for this purpose. The techniques described in this report are temperature measurement of fuel rod cladding surface and fuel centerline, cladding elongation, fuel stack elongation, fuel rod pressure by fission gas release. As a result of analysis we define the instrumentation techniques for the DUPIC fuel irradiation test at HANARO. The development of high temperature thermocouple, the method of attachment to fuel cladding and inside the fuel, the method of fuel rod elongation and fuel rod pressure measurement (by LVDT) are essential techniques. In addition to the instrumentation technique, re-instrumentation technique have been developed for high burn-up fuel in commercial reactor. The re-instrumentation is a technique of re-fabrication and instrumentation of irradiated fuel rod. It technology can be utilized as a reference for us to develop instrumented test rig with a remote manner. 19 refs., 38 figs. (Author)

  10. Thermal Analysis of ZPPR High Pu Content Stored Fuel

    Directory of Open Access Journals (Sweden)

    Charles W. Solbrig

    2014-01-01

    Full Text Available The Zero Power Physics Reactor (ZPPR operated from April 18, 1969, until 1990. ZPPR operated at low power for testing nuclear reactor designs. This paper examines the temperature of Pu content ZPPR fuel while it is in storage. Heat is generated in the fuel due to Pu and Am decay and is a concern for possible cladding damage. Damage to the cladding could lead to fuel hydriding and oxidizing. A series of computer simulations were made to determine the range of temperatures potentially occuring in the ZPPR fuel. The maximum calculated fuel temperature is 292°C (558°F. Conservative assumptions in the model intentionally overestimate temperatures. The stored fuel temperatures are dependent on the distribution of fuel in the surrounding storage compartments, the heat generation rate of the fuel, and the orientation of fuel. Direct fuel temperatures could not be measured but storage bin doors, storage sleeve doors, and storage canister temperatures were measured. Comparison of these three temperatures to the calculations indicates that the temperatures calculated with conservative assumptions are, as expected, higher than the actual temperatures. The maximum calculated fuel temperature with the most conservative assumptions is significantly below the fuel failure criterion of 600°C (1,112°F.

  11. Fuel performance modeling for proposed Th-based Canadian SCWR

    Energy Technology Data Exchange (ETDEWEB)

    Bell, J.S.; Chan, P.K. [Royal Military College of Canada, Chemistry and Chemical Engineering Department, Kingston, ON (Canada)

    2014-07-01

    The fuel assembly for the Canadian Super Critical Water Reactor (SCWR) is in the conceptual design phase. The proposed fuel pellets are made of ceramic Th-Pu mixed oxide ((Th,Pu)O{sub 2}). Neutronics and thermal hydraulics calculations are being undertaken by the nuclear industry to optimize the fuel assembly within a pressure tube. The SCWR working groups have established two conceptual fuel element designs, which defines outer diameter, fuel composition, cladding material, exit burnup etc. A detailed fuel element performance assessment under in-reactor conditions could be used to determine cladding material thickness and suitability and to optimize the fuel pellet geometry. This work reports the development of a fuel performance model to predict the behaviour of the Canadian SCWR fuel using the finite element method (COMSOL). An initial approach is to develop a thorium-uranium mixed-oxide ((Th,U)O{sub 2}) model. Preliminary results from this model agree with fuel irradiation data. Uranium dioxide (UO{sub 2}) fuel, under the same conditions, is also being modeled and compared. A plan to model (Th,Pu)O{sub 2} SCWR fuel will also be briefly presented here. (author)

  12. Filtration and Leach Testing for PUREX Cladding Sludge and REDOX Cladding Sludge Actual Waste Sample Composites

    Energy Technology Data Exchange (ETDEWEB)

    Shimskey, Rick W.; Billing, Justin M.; Buck, Edgar C.; Casella, Amanda J.; Crum, Jarrod V.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Hallen, Richard T.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Swoboda, Robert G.

    2009-03-02

    A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan (Barnes and Voke 2006). The test program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Hanford Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Under test plan TP RPP WTP 467 (Fiskum et al. 2007), eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. Under this test plan, a waste testing program was implemented that included: • Homogenizing the archive samples by group as defined in the test plan. • Characterizing the homogenized sample groups. • Performing parametric leaching testing on each group for compounds of interest. • Performing bench-top filtration/leaching tests in the hot cell for each group to simulate filtration and leaching activities if they occurred in the UFP2 vessel of the WTP Pretreatment Facility. This report focuses on a filtration/leaching test performed using two of the eight waste composite samples. The sample groups examined in this report were the plutonium-uranium extraction (PUREX) cladding waste sludge (Group 3, or CWP) and reduction-oxidation (REDOX) cladding waste sludge (Group 4, or CWR). Both the Group 3 and 4 waste composites were anticipated to be high in gibbsite, thus requiring caustic leaching. WTP RPT 167 (Snow et al. 2008) describes the homogenization, characterization, and parametric leaching activities before benchtop filtration/leaching testing of these two waste groups. Characterization and initial parametric data in that report were used to plan a single filtration/leaching test using a blend of both wastes. The test focused on filtration testing of the waste and caustic leaching for aluminum, in the form

  13. PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

    Science.gov (United States)

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

    1963-09-01

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

  14. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed.

  15. Multiresponse Optimization of Laser Cladding Steel + VC Using Grey Relational Analysis in the Taguchi Method

    Science.gov (United States)

    Zhang, Zhe; Kovacevic, Radovan

    2016-07-01

    Laser cladding of metal matrix composite coatings (MMCs) has become an effective and economic method to improve the wear resistance of mechanical components. The clad quality characteristics such as clad height, carbide fraction, carbide dissolution, and matrix hardness in MMCs determine the wear resistance of the coatings. These clad quality characteristics are influenced greatly by the laser cladding processing parameters. In this study, American Iron and Steel Institute (AISI) 420 + 20% vanadium carbide (VC) was deposited on mild steel with a high powder direct diode laser. The Taguchi-based Grey relational method was used to optimize the laser cladding processing parameters (laser power, scanning speed, and powder feed rate) with the consideration of multiple clad characteristics related to wear resistance (clad height, carbide volume fraction, and Fe-matrix hardness). A Taguchi L9 orthogonal array was designed to study the effects of processing parameters on each response. The contribution and significance of each processing parameter on each clad characteristic were investigated by the analysis of variance (ANOVA). The Grey relational grade acquired from Grey relational analysis was used as the performance characteristic to obtain the optimal combination of processing parameters. Based on the optimal processing parameters, the phases and microstructure of the laser-cladded coating were characterized by using x-ray diffraction (XRD) and scanning electron microscopy (SEM) with energy-dispersive spectroscopy (EDS).

  16. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

    2004-12-01

    This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

  17. Internal correction of hafnium oxide spectral interferences and mass bias in the determination of platinum in environmental samples using isotope dilution analysis.

    Science.gov (United States)

    Rodríguez-Castrillón, José Angel; Moldovan, Mariella; García Alonso, J Ignacio

    2009-05-01

    A method has been developed for the accurate determination of platinum by isotope dilution analysis, using enriched (194)Pt, in environmental samples containing comparatively high levels of hafnium without any chemical separation. The method is based on the computation of the contribution of hafnium oxide as an independent factor in the observed isotope pattern of platinum in the spiked sample. Under these conditions, the ratio of molar fractions between natural abundance and isotopically enriched platinum was independent of the amount of hafnium present in the sample. Additionally, mass bias was corrected by an internal procedure in which the regression variance was minimised. This was possible as the mass bias factor for hafnium oxide was very close to that of platinum. The final procedure required the measurement of three platinum isotope ratios (192/194, 195/194 and 196/194) to calculate the concentration of platinum in the sample. The methodology has been validated using the reference material "BCR-723 road dust" and has been applied to different environmental matrices (road dust, air particles, bulk wet deposition and epiphytic lichens) collected in the Aspe Valley (Pyrenees Mountains). A full uncertainty budget, using Kragten's spreadsheet method, showed that the total uncertainty was limited only by the uncertainty in the measured isotope ratios and not by the uncertainties of the isotopic composition of platinum and hafnium.

  18. MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    Hagrman, D.L.; Reymann, G.A. (comps.)

    1979-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.

  19. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

  20. Matpro--version 10: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    Reymann, G.A. (comp.)

    1978-02-01

    The materials properties correlations and computer subcodes (MATPRO--Version 10) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory are described. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.

  1. Hafnium-doped hydroxyapatite nanoparticles with ionizing radiation for lung cancer treatment.

    Science.gov (United States)

    Chen, Min-Hua; Hanagata, Nobutaka; Ikoma, Toshiyuki; Huang, Jian-Yuan; Li, Keng-Yuan; Lin, Chun-Pin; Lin, Feng-Huei

    2016-06-01

    Recently, photodynamic therapy (PDT) is one of the new clinical options by generating cytotoxic reactive oxygen species (ROS) to kill cancer cells. However, the optical approach of PDT is limited by tissue penetration depth of visible light. In this study, we propose that a ROS-enhanced nanoparticle, hafnium-doped hydroxyapatite (Hf:HAp), which is a material to yield large quantities of ROS inside the cells when the nanoparticles are bombarded with high penetrating power of ionizing radiation. Hf:HAp nanoparticles are generated by wet chemical precipitation with total doping concentration of 15mol% Hf(4+) relative to Ca(2+) in HAp host material. The results show that the HAp particles could be successfully doped with Hf ions, resulted in the formation of nano-sized rod-like shape and with pH-dependent solubility. The impact of ionizing radiation on Hf:HAp nanoparticles is assessed by using in-vitro and in-vivo model using A549 cell line. The 2',7'-dichlorofluorescein diacetate (DCFH-DA) results reveal that after being exposed to gamma rays, Hf:HAp could significantly lead to the formation of ROS in cells. Both cell viability (WST-1) and cytotoxicity (LDH) assay show the consistent results that A549 lung cancer cell lines are damaged with changes in the cells' ROS level. The in-vivo studies further demonstrate that the tumor growth is inhibited owing to the cells apoptosis when Hf:HAp nanoparticles are bombarded with ionizing radiation. This finding offer a new therapeutic method of interacting with ionizing radiation and demonstrate the potential of Hf:HAp nanoparticles in tumor treatment, such as being used in a palliative treatment after lung surgical procedure. Photodynamic therapy (PDT) is one of the new clinical options by generating cytotoxic reactive oxygen species (ROS) to kill cancer cells. Unfortunately, the approach of PDT is usually limited to the treatment of systemic disease and deeper tumor, due to the limited tissue penetration depth of visible

  2. Laser cladding crack repair of austenitic stainless steel

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2009-06-01

    Full Text Available @ 2 bar pressure Fig. 6: Crack sealing and overlay of 6.0mm plate 4 Typical hardness of the 316L crack seal layers is 160 HV1kg. Crack sealing can be performed with more noble Ni-base alloys to improved... of surface compressive stresses will avoid initiation of SCC. Due to primary austenitic solidification of laser cladded 316L, no delta ferrite is present in the microstructure. Although the susceptibility...

  3. Testing Method for External Cladding Systems - Incerc Romania

    Science.gov (United States)

    Simion, A.; Dragne, H.

    2017-06-01

    This research presents a new testing method in a natural scale for external cladding systems tested on buildings with minimum than 3 floors [1]. The testing method is unique in Romania and it is similar about many fire testing current methods from European Union states. Also, presents the fire propagation and the effect of fire smoke on the building façade composed of thermal insulation. Laboratory of testing and research for building fire safety from National Institute INCERC Bucharest, provides a test method for determining the fire performance characteristics of non-loadbearing external cladding systems and external wall insulation systems when applied to the face of a building and exposed to an external fire under controlled conditions [2]. The fire exposure is representative of an external fire source or a fully-developed (post-flashover) fire in a room, venting through an opening such as a window aperture that exposes the cladding to the effects of external flames, or an external fire source. On the future, fire tests will be experimented for answer demande a number of high-profile fires where the external facade of tall buildings provided a route for vertical fire spread.

  4. FRAPCON-1: a computer code for the steady state analysis of oxide fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G. A.; Bohn, M. P.; Coleman, D. R.; Lanning, D. D.

    1978-08-01

    FRAPCON is a FORTRAN IV computer code which predicts the steady state long-term burnup response of a light water reactor fuel rod. The coupled effects of fuel and cladding deformation, temperature, and internal gas pressure on the behavior of the fuel rod are considered in determining fuel rod response. The cladding deformation model includes multi-axial, elasto-plastic analysis and considers both primary and secondary creep. The fuel temperature model considers the effects of fuel cracking and relocation in determining the fuel temperature distribution. Burnup dependent fission gas generation and release is included in calculating fuel rod internal pressure. An integral fuel rod failure subcode determines failure and failure modes based on the operating conditions at each timestep. The material property subcode, MATPRO, provides gas, fuel and cladding properties to the computational subcodes in FRAPCON. No material properties need to be supplied by the code user. FRAPCON is a completely modular code with each major computational subcode isolated within the code and coupled to the main code by subroutine calls and data transfer through argument lists. FRAPCON is soft-coupled to the transient fuel rod code, FRAP-T, to provide initial conditions to initiate analysis of such off-normal transients as a loss-of-coolant accident. The code is presently programmed and running on a CDC 7600 computer.

  5. Study of the behaviour under neutron irradiation of hafnium diboride; Etude du comportement sous irradiation neutronique du diborure d`hafnium

    Energy Technology Data Exchange (ETDEWEB)

    Cheminant-Coatanlem, P

    1998-12-31

    Owing to its good neutron cross section and to its high melting point, hafnium diboride is a potential candidate for a use as neutron absorbing material in control rod of pressurized water reactor of the next generation. The main causes of damage under neutron irradiation in this ceramic are due to the {sup 10}B(n,{alpha}){sup 7}Li reaction that introduces in the crystal structure new atoms and point defects. The materials under consideration are the stoichiometric HfB{sub 2} compound and the HfB{sub 2} + 10 vol. % Hf compound. They are been irradiated with neutrons at several fluences and temperatures. Electron irradiations, helium and lithium implantations have been carried out in order to simulate the creation of point defects and/or fission products. Transmission and scanning electron microscopy have been used to determine damage mechanisms in HfB{sub 2}. At a low temperature (<500 deg C), irradiation defects precipitate in dislocation loops of both nature, interstitial and vacancy. Those loops have a particular organisation in the HfB{sub 2} lattice: vacancy loops are lying in the basal plane and interstitial loops in planes perpendicular to basal planes. This induces anisotropic deformation of grains that originates internal stress development. These stresses are associated with the dislocation staking and consequently with the cavity formation at grain boundaries. At a higher temperature (>700 deg C), the same dislocation loops are observed. But, in addition, the irradiation defects diffuse to grain boundaries where helium bubbles are formed. The damage caused by this latter mechanism becomes predominant. The HfB{sub 2} + 10 vol. % Hf materials is more resistant under neutron irradiation than the HfB{sub 2} pellets that display a very damaged surface. This result is explained by the fact that, on the one band, the HfB{sub 2} + 10 vol. % Hf pellets have a higher toughness than the HfB{sub 2} pellets and, on the other hand, the HfB{sub 2} + 10 vol. % Hf

  6. Characteristics of laser produced plasmas of hafnium and tantalum in the 1-7 nm region

    Science.gov (United States)

    Li, Bowen; Otsuka, Takamitsu; Sokell, Emma; Dunne, Padraig; O'Sullivan, Gerry; Hara, Hiroyuki; Arai, Goki; Tamura, Toshiki; Ono, Yuichi; Dinh, Thanh-Hung; Higashiguchi, Takeshi

    2017-11-01

    Soft X-ray (SXR) spectra from hafnium and tantalum laser produced plasmas were recorded in the 1-7 nm region using two Nd:YAG lasers with pulse lengths of 170 ps and 10 ns, respectively, operating at a range of power densities. The maximum focused peak power density was 2. 3 × 1014 W cm-2 for 170 ps pulses and 1. 8 × 1012 W cm-2 for 10 ns pulses, respectively. Two intense quasicontinuous intensity bands resulting from n = 4 - n = 4 and n = 4 - n = 5 unresolved transition arrays (UTAs) dominate both sets of experimental spectra. Comparison with calculations performed with the Cowan suite of atomic structure codes as well as consideration of previous experimental and theoretical results aided identification of the most prominent features in the spectra. For the 10 ns spectrum, the highest ion stage that could be identified from the n = 4 - n = 5 arrays were lower than silver-like Hf25+ and Ta26+ (which has a 4 d 104 f ground configuration) indicating that the plasma temperature attained was too low to produce ions with an outermost 4 d subshell, while for the 170 ps plasmas the presence of significantly higher stages was deduced and lines due to 4 d-5 p transitions were clearly evident. Furthermore, we show an enhancement of emission from tantalum using dual laser irradiation, and the effect of pre-pulse durations and delay times between two pulses are demonstrated.

  7. Extraction chromatographic separations of tantalum and tungsten from hafnium and complex matrix constituents.

    Science.gov (United States)

    Snow, Mathew S; Finck, Martha R; Carney, Kevin P; Morrison, Samuel S

    2017-02-10

    Tantalum (Ta), hafnium (Hf), and tungsten (W) analyses from complex matrices require high purification of these analytes from each other and major/trace matrix constituents, however, current state-of-the-art Ta/Hf/W separations rely on traditional anion exchange approaches that show relatively similar distribution coefficient (Kd) values for each element. This work reports an assessment of three commercially available extraction chromatographic resins (TEVA, TRU, and UTEVA) for Ta/Hf/W separations. Batch contact studies show differences in Ta/Hf and Ta/W Kd values of up to 106 and 104 (respectively), representing an improvement of a factor of 100 and 300 in Ta/Hf and Ta/W Kd values (respectively) over AG1×4 resin. Variations in the Kd values as a function of HCl concentration for TRU resin show that this resin is well suited for Ta/Hf/W separations, with Ta/Hf, Ta/W, and W/Hf Kd value improvements of 10, 200, and 30 (respectively) over AG1×4 resin. Analyses of digested soil samples (NIST 2710a) using TRU resin and tandem TEVA-TRU columns demonstrate the ability to achieve extremely high purification (>99%) of Ta and W from each other and Hf, as well as enabling very high purification of Ta and W from the major and trace elemental constituents present in soils using a single chromatographic step. Copyright © 2017 Elsevier B.V. All rights reserved.

  8. Silicon-doped hafnium oxide anti-ferroelectric thin films for energy storage

    Science.gov (United States)

    Ali, Faizan; Liu, Xiaohua; Zhou, Dayu; Yang, Xirui; Xu, Jin; Schenk, Tony; Müller, Johannes; Schroeder, Uwe; Cao, Fei; Dong, Xianlin

    2017-10-01

    Motivated by the development of ultracompact electronic devices as miniaturized energy autonomous systems, great research efforts have been expended in recent years to develop various types of nano-structural energy storage components. The electrostatic capacitors characterized by high power density are competitive; however, their implementation in practical devices is limited by the low intrinsic energy storage density (ESD) of linear dielectrics like Al2O3. In this work, a detailed experimental investigation of energy storage properties is presented for 10 nm thick silicon-doped hafnium oxide anti-ferroelectric thin films. Owing to high field induced polarization and slim double hysteresis, an extremely large ESD value of 61.2 J/cm3 is achieved at 4.5 MV/cm with a high efficiency of ˜65%. In addition, the ESD and the efficiency exhibit robust thermal stability in 210-400 K temperature range and an excellent endurance up to 109 times of charge/discharge cycling at a very high electric field of 4.0 MV/cm. The superior energy storage performance together with mature technology of integration into 3-D arrays suggests great promise for this recently discovered anti-ferroelectric material to replace the currently adopted Al2O3 in fabrication of nano-structural supercapacitors.

  9. Conduction Channel Formation and Dissolution Due to Oxygen Thermophoresis/Diffusion in Hafnium Oxide Memristors.

    Science.gov (United States)

    Kumar, Suhas; Wang, Ziwen; Huang, Xiaopeng; Kumari, Niru; Davila, Noraica; Strachan, John Paul; Vine, David; Kilcoyne, A L David; Nishi, Yoshio; Williams, R Stanley

    2016-12-27

    Transition-metal-oxide memristors, or resistive random-access memory (RRAM) switches, are under intense development for storage-class memory because of their favorable operating power, endurance, speed, and density. Their commercial deployment critically depends on predictive compact models based on understanding nanoscale physicochemical forces, which remains elusive and controversial owing to the difficulties in directly observing atomic motions during resistive switching, Here, using scanning transmission synchrotron X-ray spectromicroscopy to study in situ switching of hafnium oxide memristors, we directly observed the formation of a localized oxygen-deficiency-derived conductive channel surrounded by a low-conductivity ring of excess oxygen. Subsequent thermal annealing homogenized the segregated oxygen, resetting the cells toward their as-grown resistance state. We show that the formation and dissolution of the conduction channel are successfully modeled by radial thermophoresis and Fick diffusion of oxygen atoms driven by Joule heating. This confirmation and quantification of two opposing nanoscale radial forces that affect bipolar memristor switching are important components for any future physics-based compact model for the electronic switching of these devices.

  10. Carrier Transport at Metal/Amorphous Hafnium-Indium-Zinc Oxide Interfaces.

    Science.gov (United States)

    Kim, Seoungjun; Gil, Youngun; Choi, Youngran; Kim, Kyoung-Kook; Yun, Hyung Joong; Son, Byoungchul; Choi, Chel-Jong; Kim, Hyunsoo

    2015-10-14

    In this paper, the carrier transport mechanism at the metal/amorphous hafnium-indium-zinc oxide (a-HIZO) interface was investigated. The contact properties were found to be predominantly affected by the degree of interfacial reaction between the metals and a-HIZO; that is, a higher tendency to form metal oxide phases leads to excellent Ohmic contact via tunneling, which is associated with the generated donor-like oxygen vacancies. In this case, the Schottky-Mott theory is not applicable. Meanwhile, metals that do not form interfacial metal oxide, such as Pd, follow the Schottky-Mott theory, which results in rectifying Schottky behavior. The Schottky characteristics of the Pd contact to a-HIZO can be explained in terms of the barrier inhomogeneity model, which yields a mean barrier height of 1.40 eV and a standard deviation of 0.14 eV. The work function of a-HIZO could therefore be estimated as 3.7 eV, which is in good agreement with the ultraviolet photoelectron spectroscopy (3.68 eV). Our findings will be useful for establishing a strategy to form Ohmic or Schottky contacts to a-HIZO films, which will be essential for fabricating reliable high-performance electronic devices.

  11. Dinitrogen functionalization with bis(cyclopentadienyl) complexes of zirconium and hafnium.

    Science.gov (United States)

    Chirik, Paul J

    2007-01-07

    The rich chemistry of substituted bis(cyclopentadienyl)zirconium and hafnium complexes bearing side-on coordinated dinitrogen ligands is highlighted in this Perspective. Our studies in this area were initially motivated by the desire to understand side-on vs. end-on dinitrogen coordination in bimetallic zirconocene and hafnocene N2 compounds. In the cases where eta2,eta2-dinitrogen compounds were isolated, both structural and computational data have established significant imido character in the metal-nitrogen bonds. This additional bonding interaction, which is diminished in end-on complexes bearing both terminal and bridging N2 ligands, facilitates dinitrogen functionalization by non-polar reagents including dihydrogen, carbon-hydrogen bonds and weak Brønsted acids such as water and ethanol. In hafnocene chemistry, where unwanted side-on, end-on isomerization is suppressed, cycloaddition of phenylisocyanate to coordinated N2 has also been accomplished. For N-H bond forming reactions involving H2, kinetic measurements, in addition to isotopic labelling and computational studies, are consistent with dinitrogen functionalization by 1,2-addition involving a highly ordered, four-centred transition structure.

  12. Fatigue mechanism of yttrium-doped hafnium oxide ferroelectric thin films fabricated by pulsed laser deposition.

    Science.gov (United States)

    Huang, Fei; Chen, Xing; Liang, Xiao; Qin, Jun; Zhang, Yan; Huang, Taixing; Wang, Zhuo; Peng, Bo; Zhou, Peiheng; Lu, Haipeng; Zhang, Li; Deng, Longjiang; Liu, Ming; Liu, Qi; Tian, He; Bi, Lei

    2017-02-01

    Owing to their prominent stability and CMOS compatibility, HfO2-based ferroelectric films have attracted great attention as promising candidates for ferroelectric random-access memory applications. A major reliability issue for HfO2 based ferroelectric devices is fatigue. So far, there have been a few studies on the fatigue mechanism of this material. Here, we report a systematic study of the fatigue mechanism of yttrium-doped hafnium oxide (HYO) ferroelectric thin films deposited by pulsed laser deposition. The influence of pulse width, pulse amplitude and temperature on the fatigue behavior of HYO during field cycling is studied. The temperature dependent conduction mechanism is characterized after different fatigue cycles. Domain wall pinning caused by carrier injection at shallow defect centers is found to be the major fatigue mechanism of this material. The fatigued device can fully recover to the fatigue-free state after being heated at 90 °C for 30 min, confirming the shallow trap characteristic of the domain wall pinning defects.

  13. Arc Jet Testing of Hafnium Diboride Based Ultra High Temperature Ceramics

    Science.gov (United States)

    Ellerby, Don; Beckman, Sarah; Irby, Edward; Squire, Tom; Olejniczak, Joe; Johnson, Sylvia M.; Gusman, Michael; Gasch, Matthew

    2003-01-01

    Hafnium Diboride (HFB,) based materials have shown promise for use in a number of high temperature aerospace applications, including rocket nozzles and as leading edges on hypersonic reentry vehicles. The stability of the materials in relevant environments is key to determining their suitability for a particular application. In this program we have been developing HfB2/SiC materials for use as sharp leading edges. The program as a whole included processing and characterization of the HfBJSiC materials. The specific work discussed here will focus on studies of the materials oxidation behavior in simulated reentry environments through arc jet testing. Four flat face models were tested to examine the influence of heat flux and stagnation pressure on the materials oxidation behavior. The results from arc jet testing of two HfB2/SiC cone models will also be discussed. Each cone model was run multiple times with gradually increasing heat fluxes. Total run times on a single cone model exceeded 80 minutes. For both the flat face and cone models surface temperatures well in excess of 2200 C were measured. Post test microstructural examination of the models and correlations with measured temperatures will be discussed.

  14. Facing-target mid-frequency magnetron reactive sputtered hafnium oxide film: Morphology and electrical properties

    Science.gov (United States)

    Zhang, Yu; Xu, Jun; Wang, You-Nian; Choi, Chi Kyu; Zhou, Da-Yu

    2016-03-01

    Amorphous hafnium dioxide (HfO2) film was prepared on Si (100) by facing-target mid-frequency reactive magnetron sputtering under different oxygen/argon gas ratio at room temperature with high purity Hf target. 3D surface profiler results showed that the deposition rates of HfO2 thin film under different O2/Ar gas ratio remain unchanged, indicating that the facing target midfrequency magnetron sputtering system provides effective approach to eliminate target poisoning phenomenon which is generally occurred in reactive sputtering procedure. X-ray photoelectron spectroscopy (XPS) and Fourier transform infrared spectroscopy (FTIR) demonstrated that the gradual reduction of oxygen vacancy concentration and the densification of deposited film structure with the increase of oxygen/argon (O2/Ar) gas flow ratio. Atomic force microscopy (AFM) analysis suggested that the surface of the as-deposited HfO2 thin film tends to be smoother, the root-meansquare roughness (RMS) reduced from 0.876 nm to 0.333 nm while O2/Ar gas flow ratio increased from 1/4 to 1/1. Current-Voltage measurements of MOS capacitor based on Au/HfO2/Si structure indicated that the leakage current density of HfO2 thin films decreased by increasing of oxygen partial pressure, which resulted in the variations of pore size and oxygen vacancy concentration in deposited thin films. Based on the above characterization results the leakage current mechanism for all samples was discussed systematically.

  15. Conduction Channel Formation and Dissolution Due to Oxygen Thermophoresis/Diffusion in Hafnium Oxide Memristors

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Suhas; Wang, Ziwen; Huang, Xiaopeng; Kumari, Niru; Davila, Noraica; Strachan, John Paul; Vine, David; Kilcoyne, A. L. David; Nishi, Yoshio; Williams, R. Stanley

    2016-12-27

    Due to the favorable operating power, endurance, speed, and density., transition-metal-oxide memristors, or resistive random-access memory (RRAM) switches, a