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Sample records for burnup clad material

  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

    SciTech Connect

    GrandJean, C.; Cauvin, R.; Lebuffe, C.

    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. Development of New Cladding Materials Applied for Advanced LWR Aiming at Ultra-high Burn-up and Fast Neutron Spectrum

    SciTech Connect

    Kiuchi, K.; Ogawa, H.; Ioka, I.; Kuroda, Y.; Anegawa, T.

    2002-07-01

    The ultra-high burnup more than 100 GWd/t and fast neutron spectrum tailoring are considered to be the most promising technologies applied to the advanced MOX LWRs for minimizing the electrical cost and waste management. The development of new cladding materials with the excellent irradiation properties has been conducted to realize these needs. Comparing with UO{sub 2}, to increase the internal pressure by FP gas release is accelerated with the co-production of Xe and He in MOX fuels. New stainless steels with the excellent irradiation properties, creep strength and compatibilities to high temperature water were selected to attain the reliability. The irradiation assisted stress corrosion cracking through the past experience in LWR plants is possible to inhibit by new steel making process. The problems of tritium release and PCMI is possible to inhibit by ductile niobium alloy lining. (authors)

  3. Ductile-to-brittle transition temperature for high-burnup cladding alloys exposed to simulated drying-storage conditions

    NASA Astrophysics Data System (ADS)

    Billone, M. C.; Burtseva, T. A.; Einziger, R. E.

    2013-02-01

    Structural analyses of dry casks containing high-burnup fuel require cladding mechanical properties and failure limits to assess fuel behavior. 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). A test procedure was developed to simulate the effects of drying-storage temperature histories. Following drying-storage simulation, samples were subjected to ring-compression test (RCT) loading, which was used as a ductility screening test and to simulate pinch-type loading that may occur during cask transport. RCT samples with <2% offset strain prior to >50% wall cracking were assessed as brittle. Prior to testing high-burnup cladding, many tests were conducted with pre-hydrided Zircaloy-4 (Zry-4) and ZIRLO™ to determine target 400 °C hoop stresses for high-burnup rodlets. Zry-4 cladding segments, from a 67-GWd/MTU fuel rod, with 520-620 wppm hydrogen and ZIRLO™ cladding segments from a 70-GWd/MTU fuel rod, with 350-650 wppm hydrogen were defueled and tested. Following drying-storage simulation, the extent of radial-hydride precipitation was characterized by the radial-hydride continuity factor. It was found that the DBTT was dependent on: cladding material, irradiation conditions, and drying-storage histories (stress at maximum temperature). High-burnup ZIRLO™ exhibited higher susceptible to radial-hydride formation and embrittlement than high-burnup Zry-4. It was also observed that uniformly pre-hydrided, non-irradiated cladding was not a good surrogate for high-burnup cladding because of the high density of circumferential hydrides across the wall and the high metal-matrix ductility for

  4. Postirradiation examination of the HT9 clad fuel test X425 at 2.9% burnup

    SciTech Connect

    Pahl, R G; Beck, W N; Sanecki, J E

    1987-11-01

    The X425 experiment was the first EBR-II subassembly to be irradiated with U-Pu-Zr metallic fuel clad in the HT9 alloy. This report summarizes our initial postirradiation examination of selected elements from X425 at 2.9% peak burnup. Fuel microstructure, swelling behavior, fission gas release, and fuel/clad chemical interaction are discussed.

  5. Modeling Cladding-Coolant Heat Transfer of High-Burnup Fuel During RIA

    SciTech Connect

    Wenfeng Liu; Kazimi, Mujid S.

    2006-07-01

    This paper describes a model for the cladding-coolant heat transfer of high burnup fuel during a Reactivity Initiated Accident (RIA) which is implemented in the fuel performance code FRAPTRAN 1.2. The minimum stable film boiling temperature, affected by the subcooling and the clad oxidation, is modeled by a modified Henry correlation. This accounts for the effects of thermal properties of the cladding surface on the transient temperature drop during liquid-solid contact. The transition boiling regime is described as the interpolation of the heat flux between two anchor points on the boiling curve: the Critical Heat Flux (CHF) and minimum stable film boiling. The CHF correlation is based on the Zuber hydrodynamic model multiplied by a subcooling factor. Frederking correlation is chosen to model the film boiling regime. The heat conduction through the oxide layer of the cladding surface of high burnup fuel is calculated by solving heat conduction equations with thermal properties of zirconia taken from MATPRO. This model is validated in the FRAPTRAN code for test cases of both high burnup and fresh test fuel rods including the burnup level (0--56 MW d/kg), peak fuel enthalpy deposit (70--190 cal/g), degree of subcooling (0--80 deg. C), and extent of oxidation (0--25 micron). The modified code demonstrates the capability of differentiating between the departure from nucleate boiling (DNB) and none-DNB cases. The predicted peak cladding temperature (PCT) and duration of DNB achieves generally good agreement with the experimental data. It is found that the cladding surface oxidation of high burnup fuel causes an early rewetting of cladding or suppresses DNB due to two factors: 1) Thick zirconia layer may delay the heat conducted to the surface while keeping the surface heat transfer in the most effective nucleate boiling regime. 2) The transient liquid-solid contact resulting from vapor breaking down would cause a lower interface temperature for an oxidized surface

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

    SciTech Connect

    Billone, M. C.; Burtseva, T. A.

    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).

  7. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    SciTech Connect

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.

  8. Post Irradiation Examination for Advanced Materials at Burnups Exceeding the Current Limit

    SciTech Connect

    John H. Strumpell

    2004-12-31

    Permitting fuel to be irradiated to higher burnups limits can reduce the amount of spent nuclear fuel (SNF) requiring storage and/or disposal and enable plants to operate with longer more economical cycle lengths and/or at higher power levels. Therefore, Framatome ANP (FANP) and the B&W Owner's Group (BWOG) have introduced a new fuel rod design with an advanced M5 cladding material and have irradiated several test fuel rods through four cycles. The U.S. Department of Energy (DOE) joined FANP and the BWOG in supporting this project during its final phase of collecting and evaluating high burnup data through post irradiation examination (PIE).

  9. Temperature and burnup correlated fuel-cladding chemical interaction in U-10ZR metallic fuel

    NASA Astrophysics Data System (ADS)

    Carmack, William J.

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors and provide a number of advantages over other fuel types considering their fabricability, performance, recyclability, and safety. Resistance to cladding "breach" and subsequent release of fission products and fuel constituents to the nuclear power plant primary coolant system is a key performance parameter for a nuclear fuel system. In metallic fuel, FCCI weakens the cladding, especially at high power-high temperature operation, contributing to fuel pin breach. Empirical relationships for FCCI have been developed from a large body of data collected from in-pile (EBR-II) and out-of-pile experiments [1]. However, these relationships are unreliable in predicting FCCI outside the range of EBR-II experimental data. This dissertation examines new FCCI data extracted from the MFF-series of prototypic length metallic fuel irradiations performed in the Fast Flux Test Facility (FFTF). The fuel in these assemblies operated a temperature and burnup conditions similar to that in EBR-II but with axial fuel height three times longer than EBR-II experiments. Comparing FCCI formation data from FFTF and EBR-II provides new insight into FCCI formation kinetics. A model is developed combining both production and diffusion of lanthanides to the fuel-cladding interface and subsequent reaction with the cladding. The model allows these phenomena to be influenced by fuel burnup (lanthanide concentrations) and operating temperature. Parameters in the model are adjusted to reproduce measured FCCI layer thicknesses from EBR-II and FFTF. The model predicts that, under appropriate conditions, rate of FCCI formation can be controlled by either fission product transport or by the reaction rate of the interaction species at the fuel-cladding interface. This dissertation will help forward the design of metallic fuel systems for advanced sodium cooled fast reactors by allowing the prediction of FCCI layer formation in full

  10. Analysis of Burnup and Economic Potential of Alternative Fuel Materials in Thermal Reactors

    SciTech Connect

    Oggianu, Stella Maris; No, Hee Cheon; Kazimi, Mujid S.

    2003-09-15

    A strategy is proposed for the assessment of nuclear fuel material economic potential use in future light water reactors (LWRs). In this methodology, both the required enrichment and the fuel performance limits are considered. In order to select the best fuel candidate, the optimal burnup that produces the lowest annual fuel cost within the burnup potential for a given fuel material and smear density ratio is determined.Several nuclear materials are presented as examples of the application of the methodology proposed in this paper. The alternative fuels considered include uranium dioxide (UO{sub 2}), uranium carbide (UC), uranium nitride (UN), metallic uranium (U-Zr alloy), combined thorium and uranium oxides (ThO{sub 2}/UO{sub 2}), and combined thorium and uranium metals (U/Th). For these examples, a typical LWR lattice geometry in a zirconium-based cladding was assumed. The uncertainties in the results presented are large due to the scarcity of experimental data regarding the behavior of the considered materials at high burnups. Also, chemical compatibility issues are to be considered separately.The same methodology can be applied in the future to evaluate the economic potential of other nuclear fuel materials including different cladding designs, dispersions of ceramics into ceramics, dispersions of ceramics into metals, and also for geometries other than the traditional circular fuel pin.

  11. High temperature nanoindentation hardness and Young's modulus measurement in a neutron-irradiated fuel cladding material

    NASA Astrophysics Data System (ADS)

    Kese, K.; Olsson, P. A. T.; Alvarez Holston, A.-M.; Broitman, E.

    2017-04-01

    Nanoindentation, in combination with scanning probe microscopy, has been used to measure the hardness and Young's modulus in the hydride and matrix of a high burn-up neutron-irradiated Zircaloy-2 cladding material in the temperature range 25-300 °C. The matrix hardness was found to decrease only slightly with increasing temperature while the hydride hardness was essentially constant within the temperature range. Young's modulus decreased with increasing temperature for both the hydride and the matrix of the high burn-up fuel cladding material. The hydride Young's modulus and hardness were higher than those of the matrix in the temperature range.

  12. Screening of advanced cladding materials and UN-U3Si5 fuel

    NASA Astrophysics Data System (ADS)

    Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

    2015-07-01

    In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in

  13. Cladding and Structural Materials for Advanced Nuclear Energy Systems

    SciTech Connect

    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.

  14. Potential effects of gallium on cladding materials

    SciTech Connect

    Wilson, D.F.; Beahm, E.C.; Besmann, T.M.; DeVan, J.H.; DiStefano, J.R.; Gat, U.; Greene, S.R.; Rittenhouse, P.L.; Worley, B.A.

    1997-10-01

    This paper identifies and examines issues concerning the incorporation of gallium in weapons derived plutonium in light water reactor (LWR) MOX fuels. Particular attention is given to the more likely effects of the gallium on the behavior of the cladding material. The chemistry of weapons grade (WG) MOX, including possible consequences of gallium within plutonium agglomerates, was assessed. Based on the calculated oxidation potentials of MOX fuel, the effect that gallium may have on reactions involving fission products and possible impact on cladding performance were postulated. Gallium transport mechanisms are discussed. With an understanding of oxidation potentials and assumptions of mechanisms for gallium transport, possible effects of gallium on corrosion of cladding were evaluated. Potential and unresolved issues and suggested research and development (R and D) required to provide missing information are presented.

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

    DOEpatents

    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.

  16. Material Selection for Accident Tolerant Fuel Cladding

    SciTech Connect

    Pint, Bruce A.; Terrani, Kurt A.; Yamamoto, Yukinori; Snead, Lance Lewis

    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.

  17. Material selection for accident tolerant fuel cladding

    SciTech Connect

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

    2015-09-14

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

  18. Material selection for accident tolerant fuel cladding

    DOE PAGES

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

    2015-09-14

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

  19. Material Selection for Accident Tolerant Fuel Cladding

    NASA Astrophysics Data System (ADS)

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

    2015-09-01

    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 ≥1473 K (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 1073 K (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 1473 K (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 1748 K (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 >1673 K (1400 °C) are still being evaluated.

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

    SciTech Connect

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

    2007-07-01

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

  1. Process for producing clad superconductive materials

    DOEpatents

    Cass, Richard B.; Ott, Kevin C.; Peterson, Dean E.

    1992-01-01

    A process for fabricating superconducting composite wire by the steps of placing a superconductive precursor admixture capable of undergoing a self propagating combustion in stoichiometric amounts sufficient to form a superconductive product within a metal tube, sealing one end of said tube, igniting said superconductive precursor admixture whereby said superconductive precursor admixture endburns along the length of the admixture, and cross-section reducing said tube at a rate substantially equal to the rate of burning of said superconductive precursor admixture and at a point substantially planar with the burnfront of the superconductive precursor mixture, whereby a clad superconductive product is formed in situ, the product characterized as superconductive without a subsequent sintering stage, is disclosed.

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

    SciTech Connect

    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.

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

    SciTech Connect

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

    2015-08-28

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

  4. Review of Halden Reactor Project high burnup fuel data that can be used in safety analyses

    SciTech Connect

    Wiesenack, W.

    1996-03-01

    The fuels and materials testing programmes carried out at the OECD Halden Reactor Project are aimed at providing data in support of a mechanistic understanding of phenomena, especially as related to high burnup fuel. The investigations are focused on identifying long term property changes, and irradiation techniques and instrumentation have been developed over the years which enable to assess fuel behaviour and properties in-pile. The fuel-cladding gap has an influence on both thermal and mechanical behaviour. Improved gap conductance due to gap closure at high exposure is observed even in the case of a strong contamination with released fission gas. On the other hand, pellet-cladding mechanical interaction, which is measured with cladding elongation detectors and diameter gauges, is re-established after a phase with less interaction and is increasing. These developments are exemplified with data showing changes of fuel temperature, hydraulic diameter and cladding elongation with burnup. Fuel swelling and cladding primary and secondary creep have been successfully measured in-pile. They provide data for, e.g., the possible cladding lift-off to be accounted for at high burnup. Fuel conductivity degradation is observed as a gradual temperature increase with burnup. This affects stored heat, fission gas release and temperature dependent fuel behaviour in general. The Halden Project`s data base on fission gas release shows that the phenomenon is associated with an accumulation of gas atoms at the grain boundaries to a critical concentration before appreciable release occurs. This is accompanied by an increase of the surface-to-volume ratio measured in-pile in gas flow experiments. A typical observation at high burnup is also that a burst release of fission gas may occur during a power decrease. Gas flow and pressure equilibration experiments have shown that axial communication is severely restricted at high burnup.

  5. Ultrasonic monitoring of material processing using clad buffer rod sensors

    NASA Astrophysics Data System (ADS)

    Ramos Franca, Demartonne

    Ultrasonic sensors and techniques are developed for in-line monitoring of polymer extrusion, cleanliness of molten metals and liquid flow speed at elevated temperature. Pulse-echo mode is used for the first two processes, while the through-transmission mode is applied in the third one. The ultrasonic probe consists of high performance clad buffer rods with different dimensions to thermally isolate the commercial ultrasonic transducer from materials at high temperature. The clad buffer rods are made of steel, polymer and ceramic. Steel clad buffer rods are introduced for in-line monitoring of polymer extrusion processes. Owing to its superior performance in pulse-echo mode, for the first time such a probe is installed and performs ultrasonic monitoring in the die of a co-extrusion machine and in the barrel section of a twin-screw extruder. It can reveal a variety of information relevant to process parameters, such as polymer layer thickness, interface location and adhesion quality, stability, or polymer composition change. For the ultrasonic monitoring of polymer processes, probes with acoustic impedance that matches that of the processed polymer may offer certain advantages such as quantitative viscoelastic evaluation; thus high temperature polymer clad buffer rods, in particular PEEK, are developed. It is demonstrated that this new probe exhibits unique advantages for in-line monitoring of the cure of epoxies and polymer extrusion process. Long steel clad buffer rods with a spherical focus lens machined at the probing end are proposed for cleanliness evaluation of molten metals. The potential of this focusing probe is demonstrated by means of high-resolution imaging and particles detection in molten zinc at temperatures higher than 600°C, using a single probe operated at pulse-echo mode. A contrapropagating ultrasonic flowmeter employing steel clad buffer rods is devised to operate at high temperature. It is demonstrated that these rods guide ultrasonic signals

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

    SciTech Connect

    Bevard, Bruce Balkcom

    2014-03-28

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

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

    SciTech Connect

    Asmolov, V.; Yegorova, L.; Kaplar, E.; Lioutov, K.; Smirnov, V.; Prokhorov, V.; Goryachev, A.

    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.

  8. Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

    NASA Astrophysics Data System (ADS)

    Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.

    2014-06-01

    The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel-clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel-clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

  9. Comments on ""Contact Diffusion Interaction of Materials with Cladding''

    NASA Technical Reports Server (NTRS)

    Morris, J. F.

    1972-01-01

    A Russian paper by A. A. Babad-Zakhryapina contributes much to the understanding of fuel, clad interactions, and thus to nuclear thermionic technology. In that publication the basic diffusion expression is a simple one. A more general but complicated equation for this mass transport results from the present work. With appropriate assumptions, however, the new relation reduces to Babad-Zakhryapina's version.

  10. Preliminary assessment of the fracture behavior of weld material in full-thickness clad beams

    SciTech Connect

    Keeney, J.A.; Bass, B.R.; McAfee, W.J.; Iskander, S.K.

    1994-10-01

    This report describes a testing program that utilizes full-thickness clad beam specimens to quantify fracture toughness for shallow cracks in material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPVs). The beam specimens are fabricated from a section of an RPV wall (removed from a canceled nuclear plant) that includes weld, plate, and clad material. Metallurgical factors potentially influencing fracture toughness for shallow cracks in the beam specimens include material gradients due to welding and cladding applications, as well as material inhomogeneities in welded regions due to reheating in multiple weld passes. A summary of the testing program includes a description of the specimen geometry, material properties, the testing procedure, and the experimental results form three specimens. The yield strength of the weld material was determined to be 36% higher than the yield strength of the base material. An irradiation-induced increase in yield strength of the weld material could result in a yield stress that exceeds the upper limit where code curves are valid. The high yield strength for prototypic weld material may have implications for RPV structural integrity assessments. Analyses of the test data are discussed, including comparisons of measured displacements with finite-element analysis results, applications of toughness estimation techniques, and interpretations of constraint conditions implied by stress-based constraint methodologies. Metallurgical conditions in the region of the cladding heat-affected zone are proposed as a possible explanation for the lower-bound fracture toughness measured with one of the shallow-crack clad beam specimens. Fracture toughness data from the three clad beam specimens are compared with other shallow- and deep-crack uniaxial beam and cruciform data generated previously from A 533 Grade B plate material.

  11. Ion irradiation testing of Improved Accident Tolerant Cladding Materials

    SciTech Connect

    Anderoglu, Osman; Tesmer, Joseph R.; Maloy, Stuart A.

    2014-01-14

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

  12. Fracture assessment of weld material from a full-thickness clad RPV shell segment

    SciTech Connect

    Keeney, J.A.; Bass, B.R.; McAfee, W.J.

    1996-07-01

    Fracture analysis was applied to full-thickness clad beam specimens containing shallow cracks in material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPV) at beginning of life. The beam specimens were fabricated from a section of an RPV wall (removed from a canceled nuclear plant) that includes weld, plate, and clad material. Metallurgical factors potentially influencing fracture toughness for shallow cracks in the beam specimens include gradients of material properties and residual stresses due to welding and cladding applications. Fracture toughness estimates were obtained from load vs load-line displacement and load vs crack-mouth-opening displacement data using finite-element methods and estimation schemes based on the {eta}-factor method. One of the beams experienced a significant amount of precleavage stable ductile tearing. Effects of precleavage tearing on estimates of fracture toughness were investigated using continuum damage models. Fracture toughness results from the clad beam specimens were compared with other deep- and shallow-crack single-edge notch bend (SENB) data generated previously from A533 Grade B plate material. Range of scatter for the clad beam data is consistent with that from the laboratory-scale SENB specimens tested at the same temperature.

  13. Severe accident modeling of a PWR core with different cladding materials

    SciTech Connect

    Johnson, S. C.; Henry, R. E.; Paik, C. Y.

    2012-07-01

    The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)

  14. Demonstration of a high burnup heterogeneous core using ferritic/martensitic materials

    SciTech Connect

    Lovell, A J; Fox, G L; Sutherland, W H; Hecht, S L

    1986-04-11

    The purpose of the Core Demonstration Experiment (CDE) is to demonstrate the capability of a mixed-oxide fuel system to achieve a three year life in a prototypic LMR heterogeneous reactor environment. The CDE assemblies are fabricated using wire-wrapped, large-diameter, advanced-oxide fuel and blanket pins with tempered martensitic HT9 cladding, wire wrap, and duct. The highest power fuel assembly operates with a Beginning of Life (BOL) peak linear pin power of 445 W/cm and a peak cladding temperature of 615C. The fuel and blanket assembly irradiation will start in FFTF Cycle 9 and continue for about 900 Equivalent Full Power Days (EFPD). The successful utilization of the tempered martensitic HT9 alloy in an FFTF test assembly is fully anticipated. The low swelling, observed at intermediate neutron fluence and projected to higher fluences, together with reasonable creep behavior gives acceptable mechanical performance for fuel pins, blanket pins and ducts. Duct length increase, dilation and bow; plus fuel and blanket pin diameter increases remain within specified tolerances. In addition, stress rupture data from unirradiated HT9 imply cumulative damage fractions for the nominal fuel and blanket pins that are low.

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

    SciTech Connect

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

    2017-01-01

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

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

    DOE PAGES

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

    2017-01-01

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

  17. Irradiation of SiC Clad Fuel Rods in the HFIR

    SciTech Connect

    Ott, Larry J; Bell, Gary L; Ellis, Ronald James; McDuffee, Joel Lee; Morris, Robert Noel

    2013-01-01

    During 2009 and- 2010, new test capability for the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was developed that allows testing of advanced nuclear fuels and cladding under prototypic light-water-reactor (LWR) operating conditions (i.e., cladding and fuel temperatures, fuel average linear heat generation rates, and cladding fluence). For the initial experiments for this test program, ORNL teamed with commercial fuel/cladding vendors who have developed an advanced composite-wound SiC cladding material for possible use in LWRs. The first experiment, containing SiC-clad UN fuel, was inserted in HFIR in June 2010, and the second experiment, containing SiC-clad UO2 fuel, was inserted in October 2010. Two capsules (one containing UN fuel and the other UO2) were withdrawn from their respective assemblies in November 2011 at an estimated fuel burnup of approximately 10 GWd/MTHM; and two capsules (one containing UN fuel and the other UO2) were withdrawn from their respective assemblies in February 2013 at an estimated fuel burnup of approximately 20 GWd/MTHM. These capsules are currently awaiting PIE. This paper will describe the experiment, as-run operating conditions for these capsules, and current PIE plans and status.

  18. EPRI fuel cladding integrity program

    SciTech Connect

    Yang, R.

    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.

  19. Preliminary Investigation of Zircaloy-4 as a Research Reactor Cladding Material

    SciTech Connect

    Brian K Castle

    2012-05-01

    As part of a scoping study for the ATR fuel conversion project, an initial comparison of the material properties of Zircaloy-4 and Aluminum-6061 (T6 and O-temper) is performed to provide a preliminary evaluation of Zircaloy-4 for possible inclusion as a candidate cladding material for ATR fuel elements. The current fuel design for the ATR uses Aluminum 6061 (T6 and O temper) as a cladding and structural material in the fuel element and to date, no fuel failures have been reported. Based on this successful and longstanding operating history, Zircaloy-4 properties will be evaluated against the material properties for aluminum-6061. The preliminary investigation will focus on a comparison of density, oxidation rates, water chemistry requirements, mechanical properties, thermal properties, and neutronic properties.

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

    SciTech Connect

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

    2013-10-01

    High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column were observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.

  1. Regulatory Perspective on Potential Fuel Reconfiguration and Its Implication to High Burnup Spent Fuel Storage and Transportation - 13042

    SciTech Connect

    Li, Zhian; Rahimi, Meraj; Tang, David; Aissa, Mourad; Flaganan, Michelle; Wagner, John C.

    2013-07-01

    The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases. These results indicate that the current knowledge in cladding material property is insufficient to determine the structural performance of the cladding of high burnup fuel after it has been stored in a dry cask storage system for some time. The uncertainties in material property and the elevated ductile to brittle transition temperature impose a challenge to the storage cask and transportation packaging designs because the cask designs may not be able to rely on the structural integrity of the fuel assembly for control of fissile material, radiation source, and decay heat source distributions. The fuel may reconfigure during further storage and/or the subsequent transportation conditions. In addition, the fraction of radioactive materials available for release from spent fuel under normal condition of storage and transport may also change. The spent fuel storage and/or transportation packaging vendors, spent fuel shippers, and the regulator may need to consider this possible fuel reconfiguration and its impact on the packages' ability to meet the safety requirements of Part 72 and Part 71 of Title 10 of the Code of Federal Regulations. The United States Nuclear Regulatory Commission (NRC) is working with the scientists at Oak Ridge National Laboratory (ORNL) to assess the impact of fuel reconfiguration on the safety of the dry storage systems and transportation packages. The NRC Division of Spent Fuel Storage and Transportation has formed a task force to work on the safety and regulatory concerns in relevance to high burnup

  2. Pellet cladding mechanical interactions of ceramic claddings fuels under light water reactor conditions

    NASA Astrophysics Data System (ADS)

    Li, Bo-Shiuan

    Ceramic materials such as silicon carbide (SiC) are promising candidate materials for nuclear fuel cladding and are of interest as part of a potential accident tolerant fuel design due to its high temperature strength, dimensional stability under irradiation, corrosion resistance, and lower neutron absorption cross-section. It also offers drastically lower hydrogen generation in loss of coolant accidents such as that experienced at Fukushima. With the implementation of SiC material properties to the fuel performance code, FRAPCON, performances of the SiC-clad fuel are compared with the conventional Zircaloy-clad fuel. Due to negligible creep and high stiffness, SiC-clad fuel allows gap closure at higher burnup and insignificant cladding dimensional change. However, severe degradation of SiC thermal conductivity with neutron irradiation will lead to higher fuel temperature with larger fission gas release. High stiffness of SiC has a drawback of accumulating large interfacial pressure upon pellet-cladding mechanical interactions (PCMI). This large stress will eventually reach the flexural strength of SiC, causing failure of SiC cladding instantly in a brittle manner instead of the graceful failure of ductile metallic cladding. The large interfacial pressure causes phenomena that were previously of only marginal significance and thus ignored (such as creep of the fuel) to now have an important role in PCMI. Consideration of the fuel pellet creep and elastic deformation in PCMI models in FRAPCON provide for an improved understanding of the magnitude of accumulated interfacial pressure. Outward swelling of the pellet is retarded by the inward irradiation-induced creep, which then reduces the rate of interfacial pressure buildup. Effect of PCMI can also be reduced and by increasing gap width and cladding thickness. However, increasing gap width and cladding thickness also increases the overall thermal resistance which leads to higher fuel temperature and larger fission

  3. Boronization of nickel and nickel clad materials for potential use in polymer electrolyte membrane fuel cells

    SciTech Connect

    Weil, K. Scott; Kim, Jin Yong Y.; Xia, Gordon; Coleman, J. E.; Yang, Z Gary

    2006-12-20

    A new low-cost, nickel clad bipolar plate concept is currently being developed for use in polymer electrolyte membrane fuel cells. Reported in this paper are the details of a powder-pack boronization process that would be used to establish a passivation layer on the electrolyte exposed surfaces of the bipolar plate in the final stage of manufacture. Results from energy dispersive X-ray analysis, X-ray diffraction, and scanning electron microscopy indicate that under moderate boronization conditions a homogeneous Ni3B layer grows on the exposed surfaces of the nickel clad material, the thickness of which depends on the time and temperature of boronization according to a Wagner-type scale growth relationship. At higher temperatures and longer reaction times, a Ni2B overlayer forms on top of the Ni3B during boronization.

  4. Analysis of Nickel Based Hardfacing Materials Manufactured by Laser Cladding for Sodium Fast Reactor

    NASA Astrophysics Data System (ADS)

    Aubry, P.; Blanc, C.; Demirci, I.; Dal, M.; Malot, T.; Maskrot, H.

    For improving the operational capacity, the maintenance and the decommissioning of the future French Sodium Fast Reactor ASTRID which is under study, it is asked to find or develop a cobalt free hardfacing alloy and the associated manufacturing process that will give satisfying wear performances. This article presents recent results obtained on some selected nickel-based hardfacing alloys manufactured by laser cladding, particularly on Tribaloy 700 alloy. A process parameter search is made and associated the microstructural analysis of the resulting clads. A particular attention is made on the solidification of the main precipitates (chromium carbides, boron carbides, Laves phases,…) that will mainly contribute to the wear properties of the material. Finally, the wear resistance of some samples is evaluated in simple wear conditions evidencing promising results on tribology behavior of Tribaloy 700.

  5. WC/Ni bronze composite material formation by combined methods of laser cladding and cold spraying

    NASA Astrophysics Data System (ADS)

    Ryashin, N. S.; Malikov, A. G.; Gulyaev, I. P.; Klinkov, S. V.; Kosarev, V. F.; Orishich, A. M.

    2016-10-01

    Formation of composite material containing anti-friction bronze CuAl8.5Fe4Ni5Mn1.5 and reinforced by inner bulk profiled frame of WC/Ni was considered. Combined methods of laser cladding and cold spraying were used. Reinforced cold spraying copper-bronze blend deposits on profiled frames of WC/Ni produced by laser cladding were obtained. Dependence of bronze weight concentration in cold spraying copper-bronze deposit on bronze weight part in powder blend was analyzed. Results of non-contact profiling of reinforcing WC/Ni frame, EDS analysis and microhardness tests of obtained reinforced copper-bronze-WC/Ni composites were presented.

  6. Sensitivity improvement of optical-fiber temperature sensor with solid cladding material based on multimode interference

    NASA Astrophysics Data System (ADS)

    Fukano, Hideki; Kushida, Yohei; Taue, Shuji

    2015-03-01

    We have developed a simple, high-sensitivity optical-fiber temperature sensor based on multimode interference (MMI). The fabricated MMI structure comprises three segmented fibers: a single-mode fiber (SMF); a large-core multimode fiber (MMF), whose outer surface is coated with a temperature-sensitive material; and another SMF. Fluoroacrylate and silicone rubber are tested as temperature-sensitive cladding materials. The silicone rubber coating exhibits a large shift in interference wavelength with temperature, producing a very fine temperature resolution as low as 0.01 °C.

  7. FRAPCON-3: A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup

    SciTech Connect

    Berna, G.A.; Beyer, G.A.; Davis, K.L.; Lanning, D.D.

    1997-12-01

    FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2) cladding elastic and plastic deformation, (3) fuel-cladding mechanical interaction, (4) fission gas release, (5) fuel rod internal gas pressure, (6) heat transfer between fuel and cladding, (7) cladding oxidation, and (8) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat-transfer correlations. The codes` integral predictions of mechanical behavior have not been assessed against a data base, e.g., cladding strain or failure data. Therefore, it is recommended that the code not be used for analyses of cladding stress or strain. FRAPCON-3 is programmed for use on both mainframe computers and UNIX-based workstations such as DEC 5000 or SUN Sparcstation 10. It is also programmed for personal computers with FORTRAN compiler software and at least 8 to 10 megabytes of random access memory (RAM). The FRAPCON-3 code is designed to generate initial conditions for transient fuel rod analysis by the FRAPTRAN computer code (formerly named FRAP-T6).

  8. Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

    NASA Technical Reports Server (NTRS)

    Bowles, K. J.; Gluyas, R. E.

    1975-01-01

    The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

  9. In-situ laser material process monitoring using a cladding power detection technique

    NASA Astrophysics Data System (ADS)

    Su, Daoning; Norris, Ian; Peters, Chris; Hall, Denis R.; Jones, Julian D. C.

    Progress in laser material processing may require real-time monitoring and process control for consistent quality and productivity. We report a method of in-situ monitoring of laser metal cutting and drilling using cladding power monitoring of an optical fibre beam delivery system—a technique which detects the light reflected or scattered from the workpiece. The light signal carries information about the quality of the process. Experiments involving drilling and cutting of two samples, a thin aluminum foil and a 2-mm thick stainless steel plate, confirmed the effectiveness of this method.

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

    SciTech Connect

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

    2014-01-09

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

  11. Assessment of Failure Mechanisms for GFR Vented Fuel Pins Using Hexoloy Cladding

    SciTech Connect

    Jian Gan

    2008-09-01

    A near-term vented fuel pin concept as a back-up option for the gas-cooled fast reactor (GFR) system was evaluated. This work explored the feasibility of using mixed carbide fuel (U0.85P0.15)C with off-the-shelf monolithic SiC clad in order to meet requirements for GFR fuel with an average burnup of 10%. The stress loading on the SiC cladding due to fuel swelling and thermal stress due to temperature gradient were estimated based on the data from the development of carbide fuels in the 1970’s-1980’s and the materials properties for SiC tubes. The fuel swelling at the goal burnup (10%) is expected to produce a hoop stress of approximately 32 MPa in cladding, approaching the estimated maximum allowable hoop stress (~33 MPa) for a SiC cladding reliability of 99.99%. The estimated tensile thermal stress component (~121 MPa) near the outer surface of a monolithic SiC cladding is likely to limit its application at high temperatures.

  12. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    NASA Astrophysics Data System (ADS)

    Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

    2013-10-01

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  13. Tensile properties and flow behavior analysis of modified 9Cr-1Mo steel clad tube material

    NASA Astrophysics Data System (ADS)

    Singh, Kanwarjeet; Latha, S.; Nandagopal, M.; Mathew, M. D.; Laha, K.; Jayakumar, T.

    2014-11-01

    The tensile properties and flow behavior of modified 9Cr-1Mo steel clad tube have been investigated in the framework of various constitutive equations for a wide range of temperatures (300-923 K) and strain rates (3 × 10-3 s-1, 3 × 10-4 s-1 and 3 × 10-5 s-1). The tensile flow behavior of modified 9Cr-1Mo steel clad tube was most accurately described by Voce equation. The variation of instantaneous work hardening rate (θ = dσ/dε) and σθ with stress (σ) indicated two stage behavior characterized by rapid decrease at low stresses (transient stage) followed by a gradual decrease in high stresses (Stage III). The variation of work hardening parameters and work hardening rate in terms of θ vs. σ and σθ vs. σ with temperature exhibited three distinct regimes. Rapid decrease in flow stress and work hardening parameters and rapid shift of θ vs. σ and σθ vs. σ towards low stresses with increase in temperature indicated dynamic recovery at high temperatures. Tensile properties of the material have been best predicted from Voce equation.

  14. Summary of corrosion loop test results for the Institut Laue-Langevin cladding material

    SciTech Connect

    Pawel, R.E.; Pawel, S.J.

    1995-02-01

    Corrosion testing of Grenoble Institut Laue-Langevin (ILL) reactor fuel cladding material is briefly summarized. The test specimen was fabricated from Al-Ni-Fe alloy (VDM-4194). Chemical analysis of this alloy compared well with the supplied composition report and average values for the ILL cladding and for a standard 6061 Al alloy. For the most direct comparison with the data base of oxidation results for 6061 Al, moderately aggressive thermal-hydraulic conditions and a coolant pH of 5 were chosen for this test. Further, in an attempt to induce spallation (again for comparison with behavior of 6061), the power to the test section was increased twice during the duration of the test while holding other control parameters as constant as possible. Results of destructive examination of the sample were compared with correlation calculations. In general, the product film growth rates correlated extremely well, indicating that the general growth behavior of the alloy closely follows that of 6061 Al for this range of thermal-hydraulic conditions. 8 figs., 2 tabs.

  15. Possibility of Multi-material Laser Cladding Fabrication of Nickel Alloy and Stainless Steel

    NASA Astrophysics Data System (ADS)

    Kotoban, D.; Aramov, A.; Tarasova, T.

    There are some applications in the industry for multi-material components, including device engineering and multifunctional surface engineering, having to eliminate a brazing or welding technological step. This study investigates the laser cladding process parameters, related single track geometry and quality of multi-material samples. The optimal process parameters for steel were found of 0.21-0.26 J/mm2 and 0.25 10-2 g/mm under the scanning velocity of 1400-1700 mm/min and powder feeding of 4.2-4.5 g/min. The bimetal thin walls, cylinders and cubes were manufactured within the optimal conditions. The requirements of steel and nickel joining were explored. For the examination, the optical microscopy, SEM, EDX microelement analysis and hardness analysis were involved.

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

    SciTech Connect

    Rebak, Raul B.

    2014-09-30

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

  17. New high burnup fuel models for NRC`s licensing audit code, FRAPCON

    SciTech Connect

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1996-03-01

    Fuel behavior models have recently been updated within the U.S. Nuclear Regulatory Commission steady-state FRAPCON code used for auditing of fuel vendor/utility-codes and analyses. These modeling updates have concentrated on providing a best estimate prediction of steady-state fuel behavior up to the maximum burnup level s of current data (60 to 65 GWd/MTU rod-average). A decade has passed since these models were last updated. Currently, some U.S. utilities and fuel vendors are requesting approval for rod-average burnups greater than 60 GWd/MTU; however, until these recent updates the NRC did not have valid fuel performance models at these higher burnup levels. Pacific Northwest Laboratory (PNL) has reviewed 15 separate effects models within the FRAPCON fuel performance code (References 1 and 2) and identified nine models that needed updating for improved prediction of fuel behavior at high burnup levels. The six separate effects models not updated were the cladding thermal properties, cladding thermal expansion, cladding creepdown, fuel specific heat, fuel thermal expansion and open gap conductance. Comparison of these models to the currently available data indicates that these models still adequately predict the data within data uncertainties. The nine models identified as needing improvement for predicting high-burnup behavior are fission gas release (FGR), fuel thermal conductivity (accounting for both high burnup effects and burnable poison additions), fuel swelling, fuel relocation, radial power distribution, fuel-cladding contact gap conductance, cladding corrosion, cladding mechanical properties and cladding axial growth. Each of the updated models will be described in the following sections and the model predictions will be compared to currently available high burnup data.

  18. Steady-state irradiation testing of U-Pu-Zr fuel to >18% burnup

    SciTech Connect

    Pahl, R.G.; Wisner, R.S. ); Billone, M.C.; Hofman, G.L. )

    1990-01-01

    Tests of austenitic stainless steel clad U-xP-10Zr fuel (x=o, 8, 19 wt. %) to peak burnups as high as 18.4 at. % have been completed in the EBR-II. Fuel swelling and fractional fission gas release are slowly increasing functions of burnup beyond 2 at. % burnup. Increasing plutonium content in the fuel reduces swelling and decreases the amount of fission gas which diffuses from fuel to plenum. LIFE-METAL code modelling of cladding strains is consistent with creep by fission gas loading and irradiation-induced swelling mechanisms. Fuel/cladding chemical interaction involves the ingress of rare-earth fission products. Constituent redistribution in the fuel had not limited steady-state performance. Cladding breach behavior at closure welds, in the gas plenum, and in the fuel column region have been benign events. 3 refs., 5 figs.

  19. Effects of quenching, irradiation, and annealing processes on the radiation hardness of silica fiber cladding materials (I)

    NASA Astrophysics Data System (ADS)

    Wen, Jianxiang; Gong, Renxiang; Xiao, Zhongyin; Luo, Wenyun; Wu, Wenkai; Luo, Yanhua; Peng, Gang-ding; Pang, Fufei; Chen, Zhenyi; Wang, Tingyun

    2016-07-01

    Silica optical fiber cladding materials were experimentally treated by a series of processes. The treatments involved quenching, irradiation, followed by annealing and subsequent re-irradiation, and they were conducted in order to improve the radiation hardness. The microstructural properties of the treated materials were subsequently investigated. Following the treatment of the optical fiber cladding materials, the results from the electron spin resonance (ESR) analysis demonstrated that there was a significant decrease in the radiation-induced defect structures. The ESR signals became significantly weaker when the samples were annealed at 1000 °C in combination with re-irradiation. In addition, the microstructure changes within the silica optical fiber cladding material were also analyzed using Raman spectroscopy. The experimental results demonstrate that the Sisbnd Osbnd Si bending vibrations at ω3 = 800-820 cm-1 and ω4 = 1000-1200 cm-1 (with longitudinal optical (LO) and transverse optical (TO) splitting bands) were relatively unaffected by the quenching, irradiation, and annealing treatments. In particular, the annealing process resulted in the disappearance of the defect centers; however, the LO and TO modes at the ω3 and ω4 bands were relatively unchanged. With the additional support of the ESR test results, we can conclude that the combined treatment processes can significantly enhance the radiation hardness properties of the optical fiber cladding materials.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

  2. High Resolution Neutron Radiography and Tomography of Hydrided Zircaloy-4 Cladding Materials

    SciTech Connect

    Smith, Tyler S; Bilheux, Hassina Z; Ray, Holly B; Bilheux, Jean-Christophe; Yan, Yong

    2015-01-01

    Neutron radiography for hydrogen analysis was performed with several Zircaloy-4 cladding samples with controlled hydrogen concentrations up to 1100 ppm. Hydrogen charging was performed in a process tube that was heated to facilitate hydrogen absorption by the metal. A correlation between the hydrogen concentration in the hydrided tubes and the neutron intensity was established, by which hydrogen content can be determined precisely in a small area (55 m x 55 m). Radiography analysis was also performed to evaluate the heating rate and its correlation with the hydrogen distribution through hydrided materials. In addition to radiography analysis, tomography experiments were performed on Zircaloy-4 tube samples to study the local hydrogen distribution. Through tomography analysis a 3D reconstruction of the tube was evaluated in which an uneven hydrogen distribution in the circumferential direction can be observed.

  3. Experimental and numerical investigation on cladding of corrosion-erosion resistant materials by a high power direct diode laser

    NASA Astrophysics Data System (ADS)

    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

  4. FRAPTRAN Predictability of High Burnup Advanced Fuel Performance: Analysis of the CABRI CIP0-1 and CIP0-2 Experiments

    SciTech Connect

    Del Barrio, M.T.; Herranz, L.E.

    2007-07-01

    Adequacy of analytical tools to estimate advanced high burnup fuel during a power pulse need to be soundly proven. Most of models in codes dealing with transient are extrapolations of those developed for lower irradiations. In addition, lack of open information prevents often a proper account of mechanical properties of new advanced cladding material. These circumstances make experimental programs on high burnup fuel performance an indispensable tool to enhance safety codes predictability through building up sound databases on which models can be extended or developed and on which suitable code performance can be proven. The experiments CIP0-1 and CIP0-2, carried out on 2002 in the CABRI reactor, can be seen as reference tests to investigate high burnup fuel response to RIA transients. Fuel rods of up to 75 GWd/tU (average rod burnup) encapsulated in advanced cladding materials (ZIRLO and M5) were submitted to power pulses of about 30 ms of half maximum width that injected 90-100 cal/g after 1.2 s. None of the rodlets failed during the experiments, but they underwent deformation that was experimentally determined. The FRAPTRAN code has been used for the analysis of these RIA tests. The fuel rod characterization necessary for FRAPTRAN at the end of the base irradiation, prior to the transient, was provided by FRAPCON-3. An investigation of major deviations of fuel rod characterization at the end of the base irradiation has highlighted that thermal uncertainties could result in outstanding discrepancies in FGR estimates. Transient comparison with the available data shows that FRAPTRAN presents a relatively good agreement in permanent clad hoop strain and overestimates significantly the axial elongation of the cladding. The potential effect of approximations made in describing the cladding mechanical behavior, the fuel-to-clad relative movement and the pre-transient gap width, have been all discussed. Given existing uncertainties, a conclusive statement could not be

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

    SciTech Connect

    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.

  6. Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations

    SciTech Connect

    Wagner, J.C.; DeHart, M.D.

    2000-03-01

    This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ``end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified.

  7. Effects of hydride morphology on the embrittlement of Zircaloy-4 cladding

    NASA Astrophysics Data System (ADS)

    Kim, Ju-Seong; Kim, Tae-Hoon; Kook, Dong-Hak; Kim, Yong-Soo

    2015-01-01

    Spent nuclear fuel claddings discharged from water reactors contain hydrogen up to 800 wppm depending on the burn-up and power history. During long-term dry storage, the cladding temperature slowly decreases with diminishing decay heat and absorbed hydrogen atoms are precipitated in Zr-matrix according to the terminal solid solubility of hydrogen. Under these conditions, hydrides can significantly reduce cladding ductility and impact resistance, especially when the radial hydrides are massively present in the material. In this study, the effects of hydride morphology on the embrittlement of Zircaloy-4 cladding were investigated using a ring compression test. The results show that circumferentially hydrided Zircaloy-4 cladding is brittle at room temperature but its ductility is regained substantially as the temperature goes above 150 °C. On the other hand, radially hydrided cladding remains brittle at 150 °C and micro-cracks developed in the radial hydrides can act as crack propagation paths. Fracture energy analysis shows that ductile to brittle transition temperature is low in between 25 °C and 100 °C in the former case, whereas it lies in between 200 °C and 250 °C in the latter case.

  8. Incorporation of Integral Fuel Burnable Absorbers Boron and Gadolinium into Zirconium-Alloy Fuel Clad Material

    SciTech Connect

    Sridharan, K.; Renk, T.J.; Lahoda, E.J.; Corradini, M.L

    2004-12-14

    Long-lived fuels require the use of higher enrichments of 235U or other fissile materials. Such high levels of fissile material lead to excessive fuel activity at the beginning of life. To counteract this excessive activity, integral fuel burnable absorbers (IFBA) are added to some rods in the fuel assembly. The two commonly used IFBA elements are gadolinium, which is added as gadolinium-oxide to the UO2 powder, and boron, which is applied as a zirconium-diboride coating on the UO2 pellets using plasma spraying or chemical vapor deposition techniques. 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 very costly because of their small volume and can add from 20 to 30% to the manufacturing cost of the fuel. Other manufacturing issues that impact cost and performance are maintaining the correct levels of dosing, the reduction in fuel melting point due to gadolinium-oxide additions, and parasitic neutron absorption at fuel's end-of-life. The goal of the proposed research is to develop an alternative approach that involves incorporation of boron or gadolinium into the outer surface of the fuel cladding material rather than as an additive to the fuel pellets. This paradigm shift will allow for the introduction of the IFBA in a non-nuclear regulated environment and will obviate the necessity of additional handling and processing of the fuel pellets. This could represent significant cost savings and potentially lead to greater reproducibility and control of the burnable fuel in the early stages of the reactor operation. The surface alloying is being performed using the IBEST (Ion Beam Surface Treatment) process developed at Sandia National Laboratories. IBEST involves the delivery of energetic ion beam pulses onto the surface of a material, near-surface melting, and rapid solidification. The non-equilibrium nature of such processing allows for surface

  9. In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

    SciTech Connect

    Byun, Thak Sang; Yamamoto, Yukinori; Maloy, Stuart A.; Gussev, M. N.; Terrani, K. A.

    2015-08-25

    Here, one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) is high resistance of the cladding to plastic deformation and burst failure, since the deformation and burst behavior governs the cooling efficiency of flow channels and the process of fission product release. To simulate and evaluate the deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisting of a high-resolution video camera, a light filtering unit, and monochromatic light sources. The in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. As the first application, ten (10) candidate cladding materials for ATF, i.e., five FeCrAl alloys and five nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800 °C, while negligible strain rates were measured for higher strength alloys.

  10. In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

    SciTech Connect

    Gussev, Maxim N.; Byun, Thak Sang; Yamamoto, Yukinori; Maloy, Stuart A.; Terrani, Kurt A.

    2015-11-01

    The high resistance of cladding to plastic deformation and burst failure is one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) since the deformation and burst behavior governs the cooling efficiency of flow channels and process of fission product release. To simulate and evaluate such deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisted of a high-resolution video camera, light filtering unit, and monochromatic light sources, and the in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. In this study eleven (11) candidate cladding materials for ATF, i.e., 6 FeCrAl alloys and 5 nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800°C while negligible strain rates were measured for higher strength alloys and/or for relatively thick wall specimens.

  11. In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

    DOE PAGES

    Byun, Thak Sang; Yamamoto, Yukinori; Maloy, Stuart A.; ...

    2015-08-25

    Here, one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) is high resistance of the cladding to plastic deformation and burst failure, since the deformation and burst behavior governs the cooling efficiency of flow channels and the process of fission product release. To simulate and evaluate the deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisting of a high-resolution video camera, a light filteringmore » unit, and monochromatic light sources. The in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. As the first application, ten (10) candidate cladding materials for ATF, i.e., five FeCrAl alloys and five nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800 °C, while negligible strain rates were measured for higher strength alloys.« less

  12. High Temperature Expansion Due to Compression Test for the Determination of a Cladding Material Failure Criterion under RIA Loading Conditions

    SciTech Connect

    Le Saux, M.; Poussard, C.; Averty, X.; Sainte Catherine, C.; Carassou, S.

    2007-07-01

    This paper is mainly dedicated to the development of an out-of-pile test reproducing the thermo-mechanical loading conditions encountered during the first stage of a Reactivity Initiated Accidents (RIA) transient, dominated by Pellet Clad Mechanical Interaction (PCMI). In particular, the strain-controlled clad loading under high strain rate associated with temperatures up to 600 deg. C expected during the PCMI phase is simulated by an Expansion Due to Compression (EDC) test achievable at high temperature. The use of appropriate materials for the inner pellet made it possible to achieve the tests from 20 deg. C up to 900 deg. C. The interpretation of the test data is supported by Finite Element Analysis (FEA) including parameters tuned using an inverse method coupling FEA and tests results. A deformation model, identified upon the PROMETRA (Transient Mechanical Properties) experimental database and describing the anisotropic viscoplastic behavior of Cold-Worked Stress Relieved Zircaloy-4 cladding alloys under typical RIA loading conditions, is exploited. The combined analysis of experimental results and finite element simulations provides a deeper understanding of the deformation mode (near pure hoop tension) that arises during the tests. The failure mode appears to be representative of that obtained on tubes during the PCMI stage of RIA experiments. An appropriate device is currently developed in order to reach a bi-axiality of the loading path closer to that expected during the PCMI stage (between plane-strain and equal-biaxial tension). (authors)

  13. Extended Burnup Demonstration Reactor Fuels Program. Annual progress report, April 1983-March 1984. [BWR

    SciTech Connect

    Exarhos, C.A.

    1985-06-20

    The US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities Nuclear Corporation have participated since 1979 in a cooperative Extended Burnup Demonstration Program. Under the program, standard ENC-fabricated reload fuel in the Big Rock Point and Oyster Creek reactor cores has been irradiated to discharge burnups at or beyond 35,000 MWD/MTU, one to two cycles beyond its originally projected exposure life. The program provides for examination of the fuel at poolside before and after each extended burnup cycle as well as for limited destructive hot cell examination. The 1984 progress report covers work performed under the EBD program between April 1983 and March 1984. Major milestones reached during the period include completion of a hot cell examination on four high burnup rods from Big Rock Point and of a poolside on the Oyster Creek EBD fuel at discharge. The hot cell examination of four rods at burnups to 37.2 GWD/MTU confirmed poolside measurements on the same fuel, showing the urania and gadolinia-bearing fuel rods to be in excellent condition. No major cladding degradation, pellet restructuring, or pellet-clad interaction was found in any of the samples examined. The Oyster Creek fuel, examined at an assembly average exposure of 34.5 GWD/MTU, showed good performance with regard to both diametral creepdown and clad oxide accumulation.

  14. Detailed Burnup Calculations for Testing Nuclear Data

    SciTech Connect

    Leszczynski, F.

    2005-05-24

    A general method (MCQ) has been developed by introducing a microscopic burnup scheme that uses the Monte Carlo calculated fluxes and microscopic reaction rates of a complex system and a depletion code for burnup calculations as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy-dependent cross-section libraries and full 3D geometry of the system can be input for the calculations. The resulting predictions for the system at successive burnup time steps are thus based on a calculation route where both geometry and cross sections are accurately represented, without geometry simplifications and with continuous energy data, providing an independent approach for benchmarking other methods and nuclear data of actinides, fission products, and other burnable absorbers. The main advantage of this method over the classical deterministic methods currently used is that the MCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burnup codes adopted until now are the widely used MCNP and ORIGEN codes, but other codes can be used also. For using this method, there is need of a well-known set of nuclear data for isotopes involved in burnup chains, including burnable poisons, fission products, and actinides. For fixing the data to be included in this set, a study of the present status of nuclear data is performed, as part of the development of the MCQ method. This study begins with a review of the available cross-section data of isotopes involved in burnup chains for power and research nuclear reactors. The main data needs for burnup calculations are neutron cross sections, decay constants, branching ratios, fission energy, and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross

  15. Sodium fast reactor evaluation: Core materials

    NASA Astrophysics Data System (ADS)

    Cheon, Jin Sik; Lee, Chan Bock; Lee, Byoung Oon; Raison, J. P.; Mizuno, T.; Delage, F.; Carmack, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.

  16. Evaluation of intermediate phases formed on the bonding interface of hot pressed Cu/Al clad materials

    NASA Astrophysics Data System (ADS)

    Lee, Kwang Seok; Lee, Sangmok; Lee, Jong-Sup; Kim, Yong-Bae; Lee, Geun-An; Lee, Sang-Pill; Bae, Dong-Su

    2016-09-01

    The aim of the present study is to identify the properties of intermediate phases formed on the bonding interface of hot pressed Cu/Al clad materials by transmission electron microscopy and nano-indentation analyses. Cu/Al clad materials were fabricated by hot pressing under 200 MPa at 250 °C for 1 h and then heat treated at 400 °C for 1 h. Nano-indentation measurement was conducted to evaluate the nanohardness and modulus of the intermediate phases formed between the Cu/Al interfaces. A 3-tier diffusion layer was observed at the Cu/Al interfaces. Knoop microhardness values at the bonding interface were 7 to 11 times that of the Cu and Al matrix metals. The intermediate phases formed at the bonding interface were Al4Cu9, AlCu, and Al2Cu. A mapping analysis confirmed that the Al and Cu particles moved via mutual diffusion toward the intermediate phases formed at the bonding interface. The nanohardness values of η2-AlCu and γ1-Al4Cu9 were 4 to 7 times that of the Cu and Al matrix metals. Nanohardness and Knoop microhardness measurement curves exhibited similar tendencies. The rigidity values of the respective intermediate phases can be arranged in descending order as follows: γ1-Al4Cu9 > η2-AlCu > θ-Al2Cu.

  17. Fabrication of Tungsten-Rhenium Cladding materials via Spark Plasma Sintering for Ultra High Temperature Reactor Applications

    SciTech Connect

    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.

  18. Characterization of microstructure and property evolution in advanced cladding and duct: Materials exposed to high dose and elevated temperature

    SciTech Connect

    Allen, Todd R.; Kaoumi, Djamel; Wharry, Janelle P.; Jiao, Zhijie; Topbasi, Cem; Kohnert, Aaron; Barnard, Leland; Certain, Alicia; Field, Kevin G.; Was, Gary S.; Morgan, Dane L.; Motta, Arthur T.; Wirth, Brian D.; Yang, Y.

    2015-05-20

    Designing materials for performance in high-radiation fields can be accelerated through a carefully chosen combination of advanced multiscale modeling paired with appropriate experimental validation. Here, the studies reported in this work, the combined efforts of six universities working together as the Consortium on Cladding and Structural Materials, use that approach to focus on improving the scientific basis for the response of ferritic–martensitic steels to irradiation. A combination of modern modeling techniques with controlled experimentation has specifically focused on improving the understanding of radiation-induced segregation, precipitate formation and growth under radiation, the stability of oxide nanoclusters, and the development of dislocation networks under radiation. Experimental studies use both model and commercial alloys, irradiated with both ion beams and neutrons. Lastly, transmission electron microscopy and atom probe are combined with both first-principles and rate theory approaches to advance the understanding of ferritic–martensitic steels.

  19. Characterization of microstructure and property evolution in advanced cladding and duct: Materials exposed to high dose and elevated temperature

    DOE PAGES

    Allen, Todd R.; Kaoumi, Djamel; Wharry, Janelle P.; ...

    2015-05-20

    Designing materials for performance in high-radiation fields can be accelerated through a carefully chosen combination of advanced multiscale modeling paired with appropriate experimental validation. Here, the studies reported in this work, the combined efforts of six universities working together as the Consortium on Cladding and Structural Materials, use that approach to focus on improving the scientific basis for the response of ferritic–martensitic steels to irradiation. A combination of modern modeling techniques with controlled experimentation has specifically focused on improving the understanding of radiation-induced segregation, precipitate formation and growth under radiation, the stability of oxide nanoclusters, and the development of dislocationmore » networks under radiation. Experimental studies use both model and commercial alloys, irradiated with both ion beams and neutrons. Lastly, transmission electron microscopy and atom probe are combined with both first-principles and rate theory approaches to advance the understanding of ferritic–martensitic steels.« less

  20. MOX capsule post-irradiation examination. Volume 2: Test plan for 30-GWd/MT burnup fuel

    SciTech Connect

    Morris, R.N.

    1997-12-01

    This test plan is a Level-2 document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-Oxide Fuel Irradiation Test Project Plan. The planned post-irradiation examination (PIE) work to be performed on the mixed uranium and plutonium oxide fuel capsules that have received burnups of approximately 30 GWd/MT is described. The major emphasis of this PIE task will be material interactions, particularly indications of gallium transport and interactions. This PIE will include gamma scanning, ceramography, metallography, pellet radial gallium analysis, and clad gallium analysis. A preliminary PIE report will be generated before all the work is completed so that the progress of the fuel irradiation may be known in a timely manner.

  1. Theoretical analysis of swelling characteristics of cylindrical uranium dioxide fuel pins with a niobium - 1-percent-zirconium clad

    NASA Technical Reports Server (NTRS)

    Saltsman, J. F.

    1973-01-01

    The relations between clad creep strain and fuel volume swelling are shown for cylindrical UO2 fuel pins with a Nb-1Zr clad. These relations were obtained by using the computer code CYGRO-2. These clad-strain - fuel-volume-swelling relations may be used with any fuel-volume-swelling model, provided the fuel volume swelling is isotropic and independent of the clad restraints. The effects of clad temperature (over a range from 118 to 1642 K (2010 to 2960 R)), pin diameter, clad thickness and central hole size in the fuel have been investigated. In all calculations the irradiation time was 500 hours. The burnup rate was varied.

  2. Developmental status of thermionic materials.

    NASA Technical Reports Server (NTRS)

    Yang, L.; Chin, J.

    1972-01-01

    Description of the reference materials selected for the major components of the unit cell of a thermionic pile element (TFE), the out-of-pile and in-pile test results, and current efforts for improving the life and performance of thermionic fuel elements. The component materials are required to withstand the fuel burnup and fast neutron fluence dictated by the thermionic reactor system. Tungsten was selected as the cladding material because of its compatibility with both the carbide and the oxide fuel materials. Niobium was selected as the collector material because its thermal expansion coefficient matches closely with that of the thin aluminum oxide layer used to electrically insulate the collector from the TFE sheath. An unfueled converter has performed stably over 41,000 hr. Accelerated irradiation tests have attained burnups equivalent to that for 40,000 hr of the thermionic reactor under consideration.

  3. Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term Irradiation at Elevated Temperature: Critical Experiments

    SciTech Connect

    Was, Gary; Jiao, Zhijie; Allen, Todd; Yang, Yong

    2013-12-20

    The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by microchemistry changes due to radiation-induced segregation, dislocation loop formation and growth, radiation induced precipitation, destabilization of the existing precipitate structure, as well as the possibility for void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiation-induced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses to 200 dpa and beyond). Further, predictive modeling is not yet possible, as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. This project builds upon joint work at the proposing institutions, under a NERI-C program that is scheduled to end in September, to understand the effects of

  4. Nuclear Energy Research Initiative. Development of a Stabilized Light Water Reactor Fuel Matrix for Extended Burnup

    SciTech Connect

    BD Hanson; J Abrefah; SC Marschman; SG Prussin

    2000-09-08

    The main objective of this project is to develop an advanced fuel matrix capable of achieving extended burnup while improving safety margins and reliability for present operations. In the course of this project, the authors improve understanding of the mechanism for high burnup structure (HBS) formation and attempt to design a fuel to minimize its formation. The use of soluble dopants in the UO{sub 2} matrix to stabilize the matrix and minimize fuel-side corrosion of the cladding is the main focus.

  5. Effect of lithium hydroxide on stability of fuel cladding oxide film in simulated pressurized water reactor primary water environments

    SciTech Connect

    Saario, T.; Taehtinen, S.; Piippo, J.

    1997-09-01

    The trend in pressurized water reactors (PWR) toward higher burnups, increasing lithium concentrations, and higher coolant temperatures imposes a demand for better fuel cladding corrosion and hydriding properties. There is a lack of reliable and fast in-situ techniques to investigate zirconium alloys in high-temperature water environments. The contact electric resistance (CER) technique was used to measure the electric resistance of the oxide growing on a zirconium-based fuel cladding material. Lithium hydroxide (LiOH) decreased electric resistance of the oxide when LiOH was in excess of {approximately} 70 ppm in PWR water at 300 C. Electric resistance of the oxide was dependent upon LiOH concentration and was shown to correlate inversely with the effect of LiOH on weight gain. Kinetics of the decrease of electric resistance indicated the mechanism of degradation was a phase transformation rather than a diffusion-limited process.

  6. Hydrogen uptake in Zircaloy-2 reactor fuel claddings studied with elastic recoil detection

    NASA Astrophysics Data System (ADS)

    Rajasekhara, S.; Doyle, B. L.; Enos, D. G.; Clark, B. G.

    2013-04-01

    The recent trend towards a high burn-up discharge spent nuclear fuel necessitates a thorough understanding of hydrogen uptake in Zr-based cladding materials that encapsulate spent nuclear fuel. Although it is challenging to experimentally replicate exact conditions in a nuclear reactor that lead to hydrogen uptake in claddings, in this study we have attempted to understand the kinetics of hydrogen uptake by first electrolytically charging Zircaloy-2 (Zr-2) cladding material for various durations (100 to 2,600 s), and subsequently examining hydrogen ingress with elastic recoil detection (ERD) and transmission electron microscopy (TEM). To understand the influence of irradiation damage defects on hydrogen uptake, an analogous study was performed on ion - irradiated (0.1, 1 and 25 dpa) Zr-2. Analysis of ERD data from the un-irradiated Zr-2 suggests that the growth of the hydride layer is diffusion controlled, and preliminary TEM results support this assertion. In un-irradiated Zr-2, the diffusivity of hydrogen in the hydride phase was found to be approximately 1.1 × 10-11 cm2/s, while the diffusivity in the hydride phase for lightly irradiated (0.1 and 1 dpa) Zr-2 is an order of magnitude lower. Irradiation to 25 dpa results in a hydrogen diffusivity that is comparable to the un-irradiated Zr-2. These results are compared with existing literature on hydrogen transport in Zr - based materials.

  7. PWR AXIAL BURNUP PROFILE ANALYSIS

    SciTech Connect

    J.M. Acaglione

    2003-09-17

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).

  8. Novel Accident-Tolerant Fuel Meat and Cladding

    SciTech Connect

    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.

  9. The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

    SciTech Connect

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

    2016-09-26

    Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets to the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.

  10. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    NASA Astrophysics Data System (ADS)

    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.

  11. Development of Spray Coating Methods and Materials to Replace Aluminum Cladding of Aging Aircraft for Corrosion Protection

    DTIC Science & Technology

    2007-06-01

    Replace FA9550-06-C-0113 Aluminum Cladding of Aging Aircraft for Corrosion Protection 5b. GRANTNUMBER [AF STTR Phase I Final Technical Report] 5c. PROGRAM...development of glassy coatings is limited by non-availability of aluminum based BMG powder feedstock for spraying. 15. SUBJECT TERMS STTR Report Corrosion ... aluminum cladding of aging aircraft for corrosion protection ABSTRACT The objective of this AF STTR Phase I work was to develop spray coating methods and new

  12. Fuel clad chemical interactions in fast reactor MOX fuels

    NASA Astrophysics Data System (ADS)

    Viswanathan, R.

    2014-01-01

    Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements (oxygen, cesium, tellurium, iodine) in the clad-attack are discussed and many Fuel-Clad-Chemical-Interaction (FCCI) models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL (Hanford Engineering Development Laboratory) relation is recommended: d/μm = ({0.507 ṡ [B/(at.% fission)] ṡ (T/K-705) ṡ [(O/M)i-1.935]} + 20.5) for (O/M)i ⩽ 1.98. A new model is proposed for (O/M)i ⩾ 1.98: d/μm = [B/(at.% fission)] ṡ (T/K-800)0.5 ṡ [(O/M)i-1.94] ṡ [P/(W cm-1)]0.5. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, (O/M)i is the initial oxygen-to-(uranium + plutonium) ratio, and P is the linear power rating. For fuels with [n(Pu)/n(M = U + Pu)] > 0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.

  13. Examination of T-111 clad uranium nitride fuel pins irradiated up to 13,000 hours at a clad temperature of 990 C

    NASA Technical Reports Server (NTRS)

    Slaby, J. G.; Siegel, B. L.

    1973-01-01

    The examination of 27 fuel pins irradiated for up to 13,000 hours at 990 C is described. The fuel pin clad was a tantalum alloy with uranium nitride as the nuclear fuel. Two nominal fuel pin diameters were tested with a maximum burnup of 2.34 atom percent. Twenty-two fuel pins were tested for fission gas leaks; thirteen pins leaked. Clad ductility tests indicated clad embrittlement. The embrittlement is attributed to hydrogen from an n,p reaction in the fuel. Fuel swelling was burnup dependent, and the amount of fission gas release was low, generally less than 0.5 percent. No incompatibilities between fuel, liner, and clad were in evidence.

  14. A High Temperature, non-TRISO Fuel and Clad Design with Commercial-Grade Enrichment for the Prismatic Block Very High Temperature Reactor

    SciTech Connect

    James W. Sterbentz

    2005-11-01

    The prismatic block Very High Temperature Reactor (VHTR) is a leading Generation IV reactor concept. This reactor with its relatively low core power density and large graphite mass currently satisfies the fundamental goals of the Generation IV charter. However, modifications can be made to the fuel and clad design, such that (1) VHTR uranium enrichment can be lowered to near commercial-grade pressurized water reactor (PWR) enrichments, (2) fuel burnups are extended, and (3) the thermal safety margin under transient conditions is increased. This paper outlines a possible fuel and clad design concept for use in a VHTR prismatic block core which could lead to substantial improvements in overall VHTR economics and sustainability. The results of depletion calculations here will demonstrate comparable burnup between the new fuel and clad design with only 4-6 wt% enriched uranium and the current higher enriched 10-20 wt% VHTR fuel design. In addition, the new fuel and clad design concept uses high-temperature ceramic fuel and clad materials that have the potential to significantly increase the thermal margin under VHTR transient conditions. The current fuel block design for the VHTR is the hexagonal Fort Saint Vrain (FSV) fuel block with 108 coolant channels, 210 fuel rods, and six burnable poison holes drilled axially in the block. This basic FSV block is also part of the new design concept here. The basic hexagonal block dimensions remain fixed with only the fuel pellet and clad materials and radii changed. Further optimizations of the fuel block are in progress. Currently, the proposed nuclear fuel for the prismatic VHTR is the well-known TRISO-coated particle fuel. The TRISO-coated particle offers a nice spherical, high-integrity pressure vessel containment for the fission gases (SiC layer). However, due to the multiple particle coating layers, the fuel kernel represents only 9.4% of the total particle volume (350 {micro}m kernel diameter particle) and together with

  15. Clad metal joint closure

    SciTech Connect

    Siebert, O.W.

    1985-04-09

    A plasma arc spray overlay of cladding metals is used over joints between clad metal pieces to provide a continuous cladding metal surface. The technique permits applying an overlay of a high melting point cladding metal to a cladding metal surface without excessive heating of the backing metal.

  16. Recent view to the results of pulse tests in the IGR reactor with high burn-up fuel

    SciTech Connect

    Asmolov, V.; Yegorova, L.

    1996-03-01

    Testing of 43 fuel elements (13 fuel elements with high burn-up fuel, 10 fuel elements with preirradiated cladding and fresh fuel, and 20 non-irradiated fuel elements) was carried out in the IGR pulse reactor with a half width of the reactor power pulse of about 0.7 sec. Tests were conducted in capsules with no coolant flow and with standard initial conditions in the capsule of 20{degrees}C and 0.2 MPa. Two types of coolant were used: water and air. One purpose of the test program was to determine the thresholds and mechanisms of fuel rod failure under RIA conditions for VVER fuel rods over their entire exposure range, from zero to high burn-up. These failure thresholds are often used in safety analyses. The tests and analyses were designed to reveal the influence on fuel rod failure of (1) the mechanical properties of the cladding, (2) the pellet-to-cladding gap, (3) fuel burn-up, (4) fuel-to-coolant heat transfer, and other parameters. The resulting data base can also be used for validation of computer codes used for analyzing fuel rod behavior. Three types of test specimens were used in the tests, and diagrams of these specimens are shown in Fig. 1. {open_quotes}Type-C{close_quotes} specimens were re-fabricated from commercial fuel rods of the VVER-1000 type that had been subjected to many power cycles of operation in the Novovoronezh Nuclear Power Plant (NV NPP). {open_quotes}Type-D{close_quotes} specimens were fabricated from the same commercial fuel rods used above, but the high burn-up oxide fuel was removed from the cladding and was replaced with fresh oxide fuel pellets. {open_quotes}Type-D{close_quotes} specimens thus provided a means of separating the effects of the cladding and the oxide fuel pellets and were used to examine cladding effects only.

  17. Fuel pin cladding

    DOEpatents

    Vaidyanathan, S.; Adamson, M.G.

    1983-12-16

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

  18. Fuel pin cladding

    DOEpatents

    Vaidyanathan, S.; Adamson, M.G.

    1986-01-28

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

  19. Fuel pin cladding

    DOEpatents

    Vaidyanathan, Swaminathan; Adamson, Martyn G.

    1986-01-01

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

  20. Fuel cladding behavior under rapid loading conditions

    NASA Astrophysics Data System (ADS)

    Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

    2016-02-01

    A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

  1. Summary of high burnup fuel issues and NRC`s plan of action

    SciTech Connect

    Meyer, R.O.

    1997-01-01

    For the past two years the Office of Nuclear Regulatory Research has concentrated mostly on the so-called reactivity-initiated accidents -- the RIAs -- in this session of the Water Reactor Safety Information Meeting, but this year there is a more varied agenda. RIAs are, of course, not the only events of interest for reactor safety that are affected by extended burnup operation. Their has now been enough time to consider a range of technical issues that arise at high burnup, and a list of such issues being addressed in their research program is given here. (1) High burnup capability of the steady-state code (FRAPCON) used for licensing audit calculations. (2) General capability (including high burnup) of the transient code (FRAPTRAN) used for special studies. (3) Adequacy at high burnup of fuel damage criteria used in regulation for reactivity accidents. (4) Adequacy at high burnup of models and fuel related criteria used in regulation for loss-of-coolant accidents (LOCAs). (5) Effect of high burnup on fuel system damage during normal operation, including control rod insertion problems. A distinction is made between technical issues, which may or may not have direct licensing impacts, and licensing issues. The RIAs became a licensing issue when the French test in CABRI showed that cladding failures could occur at fuel enthalpies much lower than a value currently used in licensing. Fuel assembly distortion became a licensing issue when control rod insertion was affected in some operating plants. In this presentation, these technical issues will be described and the NRC`s plan of action to address them will be discussed.

  2. Creation of heterogeneous materials by laser cladding of heterogeneous powder compositions on the base of steel and nickel

    NASA Astrophysics Data System (ADS)

    Drozdov, V. O.; Malikov, A. G.; Orishich, A. M.; Cherepanov, A. N.

    2016-11-01

    The paper presents the experimental investigation of the creation of complex composition corrosion-resistant coatings with low abrasive wear which are used in oil industry, by the method of laser cladding additive technology. Availability of the additive laser cladding of the complex composition corrosion-resistant coatings with a strengthening phase is shown. Optimal conditions for the cladding layers creation regarding the laser radiation parameters and radiation direction are found. The changes in micro-hardness, micro-structure and structure-phase components are established against the strengthening phase content in the upper clad layer. Different content of the strengthening phase in the upper layer from 10% to 30% does not change the micro-hardness of the upper layer and is on the average 700 HV. The laser irradiation of the powder mixture containing the strengthening phase results in the partial rupture of the WC strengthening phase. Hence the chemical interaction of the tungsten-cobalt powder mixture and stainless steel occurs and results in complex carbide formation.

  3. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    SciTech Connect

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-07-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at {approx}2.4, {approx}7 and {approx}11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of {approx}7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10{sup 15} n/cm{sup 2}/s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between {approx}410 deg. C and {approx}645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

  4. Process development for cladding APT tungsten targets

    SciTech Connect

    Horner, M H; Barber, R; Dalder, E

    2000-11-27

    This report describes development of processes for cladding APT Target tungsten components with a thin layer (0.127-mm) of Alloy 718, Alloy 600 or 316L stainless steel alloy. The application requires that the cladding be thermally bonded to the tungsten in order to transfer heat generated in the tungsten volume to a surrounding coolant. High temperature diffusion bonding using the hot isostatic processing (HIP) technique was selected as the method for creating a metallurgical bond between pure tungsten tubes and rods and the cladding materials. Bonding studies using a uniaxially loaded vacuum hot press were conducted in preliminary experiments to determine acceptable time-temperature conditions for diffusion bonding. The results were successfully applied in cladding tungsten rods and tubes with these alloys. Temperatures 800-810 C were suitable for cladding tungsten with Alloy 600 and 316L stainless steel alloy, whereas tungsten was clad with Alloy 718 at 1020 C.

  5. Assessment of reactivity transient experiments with high burnup fuel

    SciTech Connect

    Ozer, O.; Yang, R.L.; Rashid, Y.R.; Montgomery, R.O.

    1996-03-01

    A few recent experiments aimed at determining the response of high-burnup LWR fuel during a reactivity initiated accident (RIA) have raised concerns that existing failure criteria may be inappropriate for such fuel. In particular, three experiments (SPERT CDC-859, NSRR HBO-1 and CABRI REP Na-1) appear to have resulted in fuel failures at only a fraction of the anticipated enthalpy levels. In evaluating the results of such RIA simulation experiments, however, it is necessary that the following two key considerations be taken into account: (1) Are the experiments representative of conditions that LWR fuel would experience during an in-reactor RIA event? (2) Is the fuel that is being utilized in the tests representative of the present (or anticipated) population of LWR fuel? Conducting experiments under conditions that can not occur in-reactor can trigger response modes that could not take place during in-reactor operation. Similarly, using unrepresentative fuel samples for the tests will produce failure information that is of limited relevance to commercial LWR fuel. This is particularly important for high-burnup fuel since the manner under which the test samples are base-irradiated prior to the test will impact the mechanical properties of the cladding and will therefore affect the RIA response. A good example of this effect can be seen in the results of the SPERT CDC-859 test and in the NSRR JM-4 and JM-5 tests. The conditions under which the fuel used for these tests was fabricated and/or base-irradiated prior to the RIA pulse resulted in the formation of multiple cladding defects in the form of hydride blisters. When this fuel was subjected to the RIA power pulse, it failed by developing multiple cracks that were closely correlated with the locations of the pre-existing hydride blisters. In the case of the JM tests, many of the cracks formed within the blisters themselves and did not propagate beyond the heavily hydrided regions.

  6. Optical bistability in a silicon nitride microring resonator with azo dye-doped liquid crystal as cladding material.

    PubMed

    Wang, Chun-Ta; Tseng, Chih-Wei; Yu, Jui-Hao; Li, Yuan-Cheng; Lee, Chun-Hong; Jau, Hung-Chang; Lee, Ming-Chang; Chen, Yung-Jui; Lin, Tsung-Hsien

    2013-05-06

    This investigation reports observations of optical bistability in a silicon nitride (SiN) micro-ring resonator with azo dye-doped liquid crystal cladding. The refractive index of the cladding can be changed by switching the liquid crystal between nematic (NLC) and photo-induced isotropic (PHI) states by. Both the NLC and the PHI states can be maintained for many hours, and can be rapidly switched from one state to the other by photo-induced isomerization using 532 nm and 408 nm addressing light, respectively. The proposed device exhibits optical bistable switching of the resonance wavelength without sustained use of a power source. It has a 1.9 nm maximum spectral shift with a Q-factor of over 10000. The hybrid SiN- LC micro-ring resonator possesses easy switching, long memory, and low power consumption. It therefore has the potential to be used in signal processing elements and switching elements in optically integrated circuits.

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

    SciTech Connect

    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.

  8. Probabilistic Failure Analysis for Wound Composite Ceramic Cladding Assembly

    SciTech Connect

    Hemrick, James Gordon; Lara-Curzio, Edgar

    2013-01-01

    Advanced ceramic matrix composites based on silicon carbide (SiC) are being considered as candidate material systems for nuclear fuel cladding in light water reactors. The SiC composite structure is considered due to its assumed exceptional performance under accident scenarios, where its excellent high-temperature strength and slow reaction kinetics with steam and associated mitigated hydrogen production are desirable. The specific structures of interest consist of a monolithic SiC cylinder surrounded by interphase-coated SiC woven fibers in a tubular form and infiltrated with SiC. Additional SiC coatings on the outermost surface of the assembly are also being considered to prevent hydrothermal corrosion of the fibrous structure. The inner monolithic cylinder is expected to provide a hermetic seal to contain fission products under normal conditions. While this approach offers the promise of higher burn-up rates and safer behavior in the case of LOCA events, the reliability of such structures must be demonstrated in advance. Therefore, a probability failure analysis study was performed of such monolithic-composite hybrid structures to determine the feasibility of these design concepts. This analysis will be used to predict the future performance of candidate systems in an effort to determine the feasibility of these design concepts and to make future recommendations regarding materials selection.

  9. 78 FR 7451 - Clad Steel Plate From Japan; Determination

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-01

    ... COMMISSION Clad Steel Plate From Japan; Determination On the basis of the record \\1\\ developed in the subject... order on clad steel plate from Japan would be likely to lead to continuation or recurrence of material... contained in USITC Publication 4370 (January 2013), entitled Clad Steel Plate from Japan: Investigation...

  10. Nuclear fuel elements having a composite cladding

    DOEpatents

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

    1983-09-20

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

  11. Experience on fuel and structural materials development in the USA

    SciTech Connect

    Laidler, J.J.; Last, G.A.

    1985-06-01

    The United States has conducted extensive LMFBR fuel and structural materials development programs since the mid-1960's. Fuels and materials irradiation tests conducted in EBR-II formed the basis for evaluating the expected performance of FFTF fuel and identified candidate fuels and materials for further full-scale testing in FFTF. The performance of FFTF fuel through the first three years of reactor operation (700 EFPD) has been outstanding. Peak fuel burnup has been limited by swelling of the cold worked AISI 316 austenitic stainless steel ducts rather than by fuel pin behavior. Standard FFTF driver fuel is currently being irradiated routinely to a peak burnup of approximately 100,000 MWd/MTM. Fuel assemblies fabricated with Alloy D9 (titanium modified austenitic stainless steel) cladding and ducts will complete irradiation to an exposure of 155,000 MWd/MTM by the end of 1985, and a large-scale FFTF test program to demonstrate the extended performance capability of fuel assemblies fabricated with the ferritic/martensitic stainless steel alloy HT9 is in progress. Current information indicates that mixed oxide fuel clad with alloy HT9, a modified D9 alloy or dispersion strengthened ferritic stainless steel, enclosed in an HT9 alloy duct, can be expected to achieve the extended burnup goals set for the US program. 3 refs., 3 figs.

  12. High Burn-Up Spent Nuclear Fuel Vibration Integrity Study

    SciTech Connect

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

    2015-01-01

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

  13. Performance of titanium stabilized D9 cladding and ducts

    SciTech Connect

    Makenas, B J

    1986-04-01

    A series of mixed oxide UO/sub 2/-PuO/sub 2/ Liquid Metal Reactor (LMR) Fuel assemblies using D9 alloy cladding and/or ducts have been irradiated to high exposure in the Fast Flux Test Facility (FFTF) with very encouraging results. D9 is a titanium modified variant of 316 stainless steel. The highest burnup attained for 316 stainless steel clad driver fuel pins in FFTF has been achieved in an assembly with a cold worked D9 duct while assemblies which simultaneously utilize both D9 ducts and D9 clad pins have achieved burnups as high as 155 MWd/kgM with a corresponding fast fluence of 25 x 10/sup 22/ n/cm/sup 2/. The cold worked D9 ducts demonstrated less length increase and less dilation than their Cold Worked 316 SS counterparts. The D9 pins also showed less length and diameter increase than comparable 316 SS pins with less evidence of deformation due to interference between the pin bundle and the duct. Corrosion behavior at the inner and outer surface of the D9 cladding has proved to be very similar to that of 316 stainless steel.

  14. Fundamental metallurgical aspects of axial splitting in zircaloy cladding

    SciTech Connect

    Chung, H. M.

    2000-04-03

    Fundamental metallurgical aspects of axial splitting in irradiated Zircaloy cladding have been investigated by microstructural characterization and analytical modeling, with emphasis on application of the results to understand high-burnup fuel failure under RIA situations. Optical microscopy, SEM, and TEM were conducted on BWR and PWR fuel cladding tubes that were irradiated to fluence levels of 3.3 x 10{sup 21} n cm{sup {minus}2} to 5.9 x 10{sup 21} n cm{sup {minus}2} (E > 1 MeV) and tested in hot cell at 292--325 C in Ar. The morphology, distribution, and habit planes of macroscopic and microscopic hydrides in as-irradiated and posttest cladding were determined by stereo-TEM. The type and magnitude of the residual stress produced in association with oxide-layer growth and dense hydride precipitation, and several synergistic factors that strongly influence axial-splitting behavior were analyzed. The results of the microstructural characterization and stress analyses were then correlated with axial-splitting behavior of high-burnup PWR cladding reported for simulated-RIA conditions. The effects of key test procedures and their implications for the interpretation of RIA test results are discussed.

  15. Oxygen potential measurements in high burnup LWR U0 2 fuel

    NASA Astrophysics Data System (ADS)

    Matzke, Hj.

    1995-05-01

    A miniature solid state galvanic cell was used to measure the oxygen potential Δ overlineG( O2) of reactor irradiated U0 2 fuel at different burnups in the range of 28 to ⩾ 150 GWd d/t M. This very high burnup was achieved in the rim region of a fuel with a cross section average burnup of 75 GWd d/t M. The fuels had different enrichments and therefore different contributions of fission of 235U and 239Pu. The temperature range covered was 900 to 1350 K. None of the fuels showed a significant oxidation. Rather, if allowance is made for the dissolved rare earth fission products and the Pu formed during irradiation, some of the fuels were very slightly substoichiometric and the highest possible degree of oxidation corresponded to U0 2.001. In general, the Δ overlineG( O2) at 750°C was about -400 kJ/mol, corresponding to the Δ overlineG( O2) of the reaction Mo + O 2 → MoO 2. The implication of these results which are in contrast to commonly assumed ideas that U0 2 fuel oxidizes due to burnup, are discussed and the importance of the fission product Mo and of the zircaloy clad as oxygen buffers is outlined.

  16. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    SciTech Connect

    Wagner, J.C.

    2002-10-23

    This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

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

    SciTech Connect

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

    2014-01-01

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

  18. Thermal properties of U-Mo alloys irradiated to moderate burnup and power

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-01

    A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U-Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U-Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U-Mo alloy decreased approximately 30% for a fission density of 3.30 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.52 × 1021 fissions cm-3 from unirradiated values at 200 °C. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  19. Draft evaluation of the frequency for gas sampling for the high burnup confirmatory data project

    SciTech Connect

    Stockman, Christine T.; Alsaed, Halim A.; Bryan, Charles R.

    2015-03-26

    This report fulfills the M3 milestone M3FT-15SN0802041, “Draft Evaluation of the Frequency for Gas Sampling for the High Burn-up Storage Demonstration Project” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations. Gas sampling will provide information on the presence of residual water (and byproducts associated with its reactions and decomposition) and breach of cladding, which could inform the decision of when to open the project cask.

  20. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    SciTech Connect

    Burkes, Douglas E.; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-01

    A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U-Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U-Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U-Mo alloy decreased approximately 30% for a fission density of 2.88 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.08 × 1021 fissions cm-3 from unirradiated values at 200 oC. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  1. Clad Degradation - FEPs Screening Arguments

    SciTech Connect

    E. Siegmann

    2004-03-17

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

  2. Materials considerations in accelerator targets

    SciTech Connect

    Peacock, H.B. Jr.; Iyer, N.C.; Louthan, M.R. Jr.

    1994-08-01

    Future nuclear materials production and/or the burn-up of long lived radioisotopes may be accomplished through the capture of spallation produced neutrons in accelerators. Aluminum clad-lead and/or lead alloys has been proposed as a spallation target. Aluminum was the cladding choice because of the low neutron absorption cross section, fast radioactivity decay, high thermal conductivity, and excellent fabricability. Metallic lead and lead oxide powders were considered for the target core with the fabrication options being casting or powder metallurgy (PM). Scoping tests to evaluate gravity casting, squeeze casting, and casting and swaging processes showed that, based on fabricability and heat transfer considerations, squeeze casting was the preferred option for manufacture of targets with initial core cladding contact. Thousands of aluminum clad aluminum-lithium alloy core targets and control rods for tritium production have been fabricated by coextrusion processes and successfully irradiated in the SRS reactors. Tritium retention in, and release from the coextruded product was modeled from experimental and operational data. Newly produced tritium atoms were trapped by lithium atoms to form a lithium tritide. The effective tritium pressure required for trap or tritide stability was the equilibrium decomposition pressure of tritium over a lithium tritide-aluminum mixture. The temperature dependence of tritium release was determined by the permeability of the cladding to tritium and the local equilibrium at the trap sites. The model can be used to calculate tritium release from aluminum clad, aluminum-lithium alloy targets during postulated accelerator operational and accident conditions. This paper describes the manufacturing technologies evaluated and presents the model for tritium retention in aluminum clad, aluminum-lithium alloy tritium production targets.

  3. Materials considerations in accelerator targets

    SciTech Connect

    Peacock, H. B. Jr.; Iyer, N. C.; Louthan, M. R. Jr.

    1995-09-15

    Future nuclear materials production and/or the burn-up of long lived radioisotopes may be accomplished through the capture of spallation produced neutrons in accelerators. Aluminum clad-lead and/or lead alloys has been proposed as a spallation target. Aluminum was the cladding choice because of the low neutron absorption cross section, fast radioactivity decay, high thermal conductivity, and excellent fabricability. Metallic lead and lead oxide powders were considered for the target core with the fabrication options being casting or powder metallurgy (PM). Scoping tests to evaluate gravity casting, squeeze casting, and casting and swaging processes showed that, based on fabricability and heat transfer considerations, squeeze casting was the preferred option for manufacture of targets with initial core cladding contact. Thousands of aluminum clad aluminum-lithium alloy core targets and control rods for tritium production have been fabricated by coextrusion processes and successfully irradiated in the SRS reactors. Tritium retention in, and release from, the coextruded product was modeled from experimental and operational data. The model assumed that tritium atoms, formed by the 6Li(n,a)3He reaction, were produced in solid solution in the Al-Li alloy. Because of the low solubility of hydrogen isotopes in aluminum alloys, the irradiated Al-Li rapidly became supersaturated in tritium. Newly produced tritium atoms were trapped by lithium atoms to form a lithium tritide. The effective tritium pressure required for trap or tritide stability was the equilibrium decomposition pressure of tritium over a lithium tritide-aluminum mixture. The temperature dependence of tritium release was determined by the permeability of the cladding to tritium and the local equilibrium at the trap sites. The model can be used to calculate tritium release from aluminum clad, aluminum-lithium alloy targets during postulated accelerator operational and accident conditions. This paper describes

  4. Materials considerations in accelerator targets

    NASA Astrophysics Data System (ADS)

    Peacock, H. B.; Iyer, N. C.; Louthan, M. R.

    1995-09-01

    Future nuclear materials production and/or the burn-up of long lived radioisotopes may be accomplished through the capture of spallation produced neutrons in accelerators. Aluminum clad-lead and/or lead alloys has been proposed as a spallation target. Aluminum was the cladding choice because of the low neutron absorption cross section, fast radioactivity decay, high thermal conductivity, and excellent fabricability. Metallic lead and lead oxide powders were considered for the target core with the fabrication options being casting or powder metallurgy (PM). Scoping tests to evaluate gravity casting, squeeze casting, and casting and swaging processes showed that, based on fabricability and heat transfer considerations, squeeze casting was the preferred option for manufacture of targets with initial core cladding contact. Thousands of aluminum clad aluminum-lithium alloy core targets and control rods for tritium production have been fabricated by coextrusion processes and successfully irradiated in the SRS reactors. Tritium retention in, and release from, the coextruded product was modeled from experimental and operational data. The model assumed that tritium atoms, formed by the 6Li(n,a)3He reaction, were produced in solid solution in the Al-Li alloy. Because of the low solubility of hydrogen isotopes in aluminum alloys, the irradiated Al-Li rapidly became supersaturated in tritium. Newly produced tritium atoms were trapped by lithium atoms to form a lithium tritide. The effective tritium pressure required for trap or tritide stability was the equilibrium decomposition pressure of tritium over a lithium tritide-aluminum mixture. The temperature dependence of tritium release was determined by the permeability of the cladding to tritium and the local equilibrium at the trap sites. The model can be used to calculate tritium release from aluminum clad, aluminum-lithium alloy targets during postulated accelerator operational and accident conditions. This paper describes

  5. ZIRCONIUM-CLADDING OF THORIUM

    DOEpatents

    Beaver, R.J.

    1961-11-21

    A method of cladding thorium with zirconium is described. The quality of the bond achieved between thorium and zirconium by hot-rolling is improved by inserting and melting a thorium-zirconium alloy foil between the two materials prior to rolling. (AEC)

  6. Experimental Test Plan for PWR Sister Rods in the High Burnup Spent Fuel Data Project

    SciTech Connect

    Montgomery, Rose; Scaglione, John M; Bevard, Bruce Balkcom; Hanson, Brady; Billone, Dr. Michael

    2016-01-01

    The High Burnup Spent Fuel Data project pulled 25 sister rods (9 from the project assemblies and 16 from similar HBU assemblies) for characterization. The 25 sister rods are all high burnup and cover the range of modern domestic cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encountered during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of their long-term storage in a TN-32 dry storage cask (the cask rods ) to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup spent fuel.

  7. Burnup credit issues in transportation and storage

    SciTech Connect

    Brady, M. C.; Sanders, T. L.; Seager, K. D.; Lake, W. H.

    1992-01-01

    Reliance on the reduced reactivity of spent fuel for criticality control during transportation and storage is referred to as burnup credit. This concept has attracted international interest and is being actively pursued in the United States in the development of a new generation of transport casks. An overview of the US experience in developing a methodology to implement burnup credit in an integrated approach to transport cask design is presented in this paper. Specifically, technical issues related to the analysis, validation and implementation of burnup credit are identified and discussed.

  8. Burnup calculation by the method of first-flight collision probabilities using average chords prior to the first collision

    NASA Astrophysics Data System (ADS)

    Karpushkin, T. Yu.

    2012-12-01

    A technique to calculate the burnup of materials of cells and fuel assemblies using the matrices of first-flight neutron collision probabilities rebuilt at a given burnup step is presented. A method to rebuild and correct first collision probability matrices using average chords prior to the first neutron collision, which are calculated with the help of geometric modules of constructed stochastic neutron trajectories, is described. Results of calculation of the infinite multiplication factor for elementary cells with a modified material composition compared to the reference one as well as calculation of material burnup in the cells and fuel assemblies of a VVER-1000 are presented.

  9. The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

    DOE PAGES

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

    2016-09-26

    Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets tomore » the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.« less

  10. Burnup Credit Approach Used in the Yucca Mountain License Application

    SciTech Connect

    Scaglione, John M; Wagner, John C

    2010-01-01

    The United States Department of Energy has submitted a license application (LA) for construction authorization of a deep geologic repository at Yucca Mountain, Nevada. The license application is currently under review by the United States Nuclear Regulatory Commission (NRC). This paper will describe the methodology and approach used in the LA to address the issue of criticality and the role of burnup credit during the postclosure period. The most significant and effective measures for prevention of criticality in the repository include multiple redundant barriers that act to isolate fissionable material from water (which can act as a moderator, corrosive agent, and transporter of fissile material); inherent geometry of waste package internals and waste forms; presence of fixed neutron absorbers in waste package internals; and fuel burnup for commercial spent nuclear fuel. A probabilistic approach has been used to screen criticality from the total system performance assessment. Within the probabilistic approach, criticality is considered an event, and the total probability of a criticality event occurring within 10,000 years of disposal is calculated and compared against the regulatory criterion. The total probability of criticality includes contributions associated with both internal (within waste packages) and external (external to waste packages) criticality for each of the initiating events that could lead to waste package breach. The occurrence of and conditions necessary for criticality in the repository have been thoroughly evaluated using a comprehensive range of parameter distributions. A simplified design-basis modeling approach has been used to evaluate the probability of criticality by using numerous significant and conservative assumptions. Burnup credit is used only for evaluations of in-package configurations and uses a combination of conservative and bounding modeling approximations to ensure conservatism. This paper will review the NRC regulatory

  11. Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term and Elevated Temperature Irradiation: Modeling and Experimental Investigation

    SciTech Connect

    Wirth, Brian; Morgan, Dane; Kaoumi, Djamel; Motta, Arthur

    2013-12-01

    The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by dislocation loop formation and growth, microchemistry changes due to radiation-induced segregation, radiation-induced precipitation, destabilization of the existing precipitate structure, and in some cases, void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiationinduced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses beyond 200 dpa). Further, predictive modeling is not yet possible as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. Predictive modeling relies on an understanding of the physical processes and also on the development of microstructure and microchemical models to describe their evolution under

  12. General corrosion properties of modified PNC1520 austenitic stainless steel in supercritical water as a fuel cladding candidate material for supercritical water reactor

    NASA Astrophysics Data System (ADS)

    Nakazono, Y.; Iwai, T.; Abe, H.

    2010-03-01

    The Super-Critical Water-cooled Reactor (SCWR) has been designed and investigated because of its high thermal efficiency and plant simplification. There are some advantages including the use of a single phase coolant with high enthalpy but there are numerous potential problems, particularly with materials. As the operating temperature of supercritical water reactor will be between 280°C and 620°C with a pressure of 25MPa, the selection of materials is difficult and important. Austenitic stainless steels were selected for possible use in supercritical water systems because of their corrosion resistance and radiation resistance. The PNC1520 austenitic stainless steel developed by Japan Atomic Energy Agency (JAEA) as a nuclear fuel cladding material for a Na-cooled fast breeder reactor. The corrosion data of PNC1520 in supercritical water (SCW) is required but does not exist. The purpose of the present study is to research the corrosion properties for PNC1520 austenitic stainless steel in supercritical water. The supercritical water corrosion test was performed for the standard PNC1520 (1520S) and the Ti-additional type of PNC1520 (1520Ti) by using a supercritical water autoclave. Corrosion tests on the austenitic 1520S and 1520Ti steels in supercritical water were performed at 400, 500 and 600°C with exposures up to 1000h. The amount of weight gain, weight loss and weight of scale were evaluated after the corrosion test in supercritical water for both austenitic steels. After 1000h corrosion test performed, the weight gains of both austenitic stainless steels were less than 2 g/m2 at 400°C and 500°C . But both weight gain and weight loss of 1520Ti were larger than those of 1520S at 600°C . By increasing the temperature to 600°C, the surface of 1520Ti was covered with magnetite formed in supercritical water and dissolution of the steel alloying elements has been observed. In view of corrosion, 1520S may have larger possibility than 1520Ti to adopt a

  13. Nuclear reactor fuel element with vanadium getter on cladding

    DOEpatents

    Johnson, Carl E.; Carroll, Kenneth G.

    1977-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

  14. ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

    SciTech Connect

    A.H. Wells

    2004-11-17

    The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent {sup 235}U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU).

  15. Irradiation of three T-111 clad uranium nitride fuel pins for 8070 hours at 990 C (1815 F)

    NASA Technical Reports Server (NTRS)

    Slaby, J. G.; Siegel, B. L.; Gedeon, L.; Galbo, R. J.

    1973-01-01

    The design and successful operation of three tantalum alloy (Ta-8W-2Hf) clad uranium mononitride (UN) fuel pins irradiated for 8070 hr at 990 C (1815 F) is described. Two pin diameters having measured burnups of 0.47 and 0.90 uranium atom percent were tested. No clad failures or swelling was detected; however, postirradiation clad samples tested failed with 1 percent strain. The fuel density decrease was 2 percent, and the fission gas release was less than 0.05 percent. Isotropic fuel swelling, which averaged about 0.5 percent, was less than fuel pin assembly clearances. Thus the clad was not strained. Thermocouples with a modified hot zone operated at average temperatures to 1100 C (2012 F) without failure. Factors that influence the ability to maintain uniform clad temperature as well as the results of the heat transfer calculations are discussed.

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

    SciTech Connect

    Unal, Cetin; Galloway, Jack D.

    2014-09-12

    In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermal swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/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. A PWR Thorium Pin Cell Burnup Benchmark

    SciTech Connect

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

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

    SciTech Connect

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

    2015-09-01

    It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritium labelled water (THO) or T2O. In a standard processing flowsheet, tritium management would be accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy® cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste for Zircaloy® clad fuel by recovering the zirconium from the cladding for reuse. These recycle processes could release the tritium in the cladding. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0–96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. To gain a better understanding of how tritium in cladding

  19. Investigation of Burnup Credit Modeling Issues Associated with BWR Fuel

    SciTech Connect

    Wagner, J.C.

    2000-10-12

    Although significant effort has been dedicated to the study of burnup-credit issues over the past decade, U.S. studies to-date have primarily focused on spent pressurized-water-reactor (PWR) fuel. The current licensing approach taken by the U.S. Department of Energy for burnup credit in transportation seeks approval for PWR fuel only. Burnup credit for boiling-water-reactor (BWR) fuel has not yet been formally sought. Burnup credit for PWR fuel was pursued first because: (1) nearly two-thirds (by mass) of the total discharged commercial spent fuel in the United States is PWR fuel, (2) it can substantially increase the fuel assembly capacity with respect to current designs for PWR storage and transportation casks, and (3) fuel depletion in PWRs is generally less complicated than fuel depletion in BWRs. However, due to international needs, the increased enrichment of modern BWR fuels, and criticality safety issues related to permanent disposal within the United States, more attention has recently focused on spent BWR fuel. Specifically, credit for fuel burnup in the criticality safety analysis for long-term disposal of spent nuclear fuel enables improved design efficiency, which, due to the large mass of fissile material that will be stored in the repository, can have substantial financial benefits. For criticality safety purposes, current PWR storage and transportation canister designs employ flux traps between assemblies. Credit for fuel burnup will eliminate the need for these flux traps, and thus, significantly increase the PWR assembly capacity (for a fixed canister volume). Increases in assembly capacity of approximately one-third are expected. In contrast, current BWR canister designs do not require flux traps for criticality safety, and thus, are already at their maximum capacity in terms of physical storage. Therefore, benefits associated with burnup credit for BWR storage and transportation casks may be limited to increasing the enrichment capacity and

  20. MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION

    SciTech Connect

    Nichols, T.; Sternat, M.; Charlton, W.

    2011-05-08

    MONTEBURNS is a Monte-Carlo depletion routine utilizing MCNP and ORIGEN 2.2. Uncertainties exist in the MCNP transport calculation, but this information is not passed to the depletion calculation in ORIGEN or saved. To quantify this transport uncertainty and determine how it propagates between burnup steps, a statistical analysis of a multiple repeated depletion runs is performed. The reactor model chosen is the Oak Ridge Research Reactor (ORR) in a single assembly, infinite lattice configuration. This model was burned for a 25.5 day cycle broken down into three steps. The output isotopics as well as effective multiplication factor (k-effective) were tabulated and histograms were created at each burnup step using the Scott Method to determine the bin width. It was expected that the gram quantities and k-effective histograms would produce normally distributed results since they were produced from a Monte-Carlo routine, but some of results do not. The standard deviation at each burnup step was consistent between fission product isotopes as expected, while the uranium isotopes created some unique results. The variation in the quantity of uranium was small enough that, from the reaction rate MCNP tally, round off error occurred producing a set of repeated results with slight variation. Statistical analyses were performed using the {chi}{sup 2} test against a normal distribution for several isotopes and the k-effective results. While the isotopes failed to reject the null hypothesis of being normally distributed, the {chi}{sup 2} statistic grew through the steps in the k-effective test. The null hypothesis was rejected in the later steps. These results suggest, for a high accuracy solution, MCNP cell material quantities less than 100 grams and greater kcode parameters are needed to minimize uncertainty propagation and minimize round off effects.

  1. Economic incentives and recommended development for commercial use of high burnup fuels in the once-through LWR fuel cycle

    SciTech Connect

    Stout, R.B.; Merckx, K.R.; Holm, J.S.

    1981-01-01

    This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels.

  2. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    SciTech Connect

    R. Schreiner

    2004-10-21

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

  3. White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

    SciTech Connect

    Sindelar, R.; Louthan, M.; PNNL, B.

    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

  4. Metal clad aramid fibers for aerospace wire and cable

    NASA Technical Reports Server (NTRS)

    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.

  5. Oxide-cladding aluminum nitride photonic crystal slab: Design and investigation of material dispersion and fabrication induced disorder

    NASA Astrophysics Data System (ADS)

    Melo, E. G.; Carvalho, D. O.; Ferlauto, A. S.; Alvarado, M. A.; Carreño, M. N. P.; Alayo, M. I.

    2016-01-01

    Photonic crystal slabs with a lower-index material surrounding the core layer are an attractive choice to circumvent the drawbacks in the fabrication of membranes suspended in air. In this work we propose a photonic crystal (PhC) slab structure composed of a triangular pattern of air holes in a multilayer thin film of aluminum nitride embedded in silicon dioxide layers designed for operating around 450 nm wavelengths. We show the design of an ideal structure and analyze the effects of material dispersion based on a first-order correction perturbation theory approach using dielectric functions obtained by experimental measurements of the thin film materials. Numerical methods were used to investigate the effects of fabrication induced disorder of typical nanofabrication processes on the bandgap size and spectral response of the proposed device. Deviation in holes radii and positions were introduced in the proposed PhC slab model with a Gaussian distribution profile. Impacts of slope in holes sidewalls that might result from the dry etching of AlN were also evaluated. The results show that for operation at the midgap frequency, slope in holes sidewalls is more critical than displacements in holes sizes and positions.

  6. Oxide-cladding aluminum nitride photonic crystal slab: Design and investigation of material dispersion and fabrication induced disorder

    SciTech Connect

    Melo, E. G. Alvarado, M. A.; Carreño, M. N. P.; Alayo, M. I.; Carvalho, D. O.; Ferlauto, A. S.

    2016-01-14

    Photonic crystal slabs with a lower-index material surrounding the core layer are an attractive choice to circumvent the drawbacks in the fabrication of membranes suspended in air. In this work we propose a photonic crystal (PhC) slab structure composed of a triangular pattern of air holes in a multilayer thin film of aluminum nitride embedded in silicon dioxide layers designed for operating around 450 nm wavelengths. We show the design of an ideal structure and analyze the effects of material dispersion based on a first-order correction perturbation theory approach using dielectric functions obtained by experimental measurements of the thin film materials. Numerical methods were used to investigate the effects of fabrication induced disorder of typical nanofabrication processes on the bandgap size and spectral response of the proposed device. Deviation in holes radii and positions were introduced in the proposed PhC slab model with a Gaussian distribution profile. Impacts of slope in holes sidewalls that might result from the dry etching of AlN were also evaluated. The results show that for operation at the midgap frequency, slope in holes sidewalls is more critical than displacements in holes sizes and positions.

  7. Impact investigation of reactor fuel operating parameters on reactivity for use in burnup credit applications

    NASA Astrophysics Data System (ADS)

    Sloma, Tanya Noel

    When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with the fuel burnup, cooling time, and the initial enrichment of fissile material in the fuel. With regard to long-term SNF disposal and transportation, tremendous benefits, such as increased capacity, flexibility of design and system operations, and reduced overall costs, provide an incentive to seek burnup credit for criticality safety evaluations. The Nuclear Regulatory Commission issued Interim Staff Guidance 8, Revision 2 in 2002, endorsing burnup credit of actinide composition changes only; credit due to actinides encompasses approximately 30% of exiting pressurized water reactor SNF inventory and could potentially be increased to 90% if fission product credit were accepted. However, one significant issue for utilizing full burnup credit, compensating for actinide and fission product composition changes, is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters can have a significant effect on the isotopic inventory of the fuel, and thus the residual reactivity. This research seeks to quantify the reactivity impact on a system from dominant depletion parameters (i.e., fuel temperature, moderator density, burnable poison rod, burnable poison rod history, and soluble boron concentration). Bounding depletion parameters were developed by statistical evaluation of a database containing reactor operating histories. The database was generated from summary reports of commercial reactor criticality data. Through depletion calculations, utilizing the SCALE 6 code package, several light

  8. Isotopic validation for PWR actinide-only burnup credit using Yankee Rowe data

    SciTech Connect

    1997-11-01

    Safety analyses of criticality control systems for transportation packages include an assumption that the spent nuclear fuel (SNF) loaded into the package is fresh or unirradiated. In other words, the spent fuel is assumed to have its original, as-manufactured U-235 isotopic content. The ``fresh fuel`` assumption is very conservative since the potential reactivity of the nuclear fuel is substantially reduced after being irradiated in the reactor core. The concept of taking credit for this reduction in nuclear fuel reactivity due to burnup of the fuel, instead of using the fresh fuel assumption in the criticality safety analysis, is referred to as ``Burnup Credit.`` Burnup credit uses the actual physical composition of the fuel and accounts for the net reduction of fissile material and the buildup of neutron absorbers in the fuel as it is irradiated. Neutron absorbers include actinides and other isotopes generated as a result of the fission process. Using only the change in actinide isotopes in the burnup credit criticality analysis is referred to as ``Actinide-Only Burnup Credit.`` The use of burnup credit in the design of criticality control systems enables more spent fuel to be placed in a package. Increased package capacity results in a reduced number of storage, shipping and disposal containers for a given number of SNF assemblies. Fewer shipments result in a lower risk of accidents associated with the handling and transportation of spent fuel, thus reducing both radiological and nonradiological risk to the public. This paper describes the modeling and the results of comparison between measured and calculated isotopic inventories for a selected number of samples taken from a Yankee Rowe spent fuel assembly.

  9. 46 CFR 111.60-23 - Metal-clad (Type MC) cable.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Metal-clad (Type MC) cable. 111.60-23 Section 111.60-23...-GENERAL REQUIREMENTS Wiring Materials and Methods § 111.60-23 Metal-clad (Type MC) cable. (a) Metal-clad (Type MC) cable permitted on board a vessel must be continuous corrugated metal-clad cable. (b)...

  10. 46 CFR 111.60-23 - Metal-clad (Type MC) cable.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 4 2014-10-01 2014-10-01 false Metal-clad (Type MC) cable. 111.60-23 Section 111.60-23...-GENERAL REQUIREMENTS Wiring Materials and Methods § 111.60-23 Metal-clad (Type MC) cable. (a) Metal-clad (Type MC) cable permitted on board a vessel must be continuous corrugated metal-clad cable. (b)...

  11. 46 CFR 111.60-23 - Metal-clad (Type MC) cable.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 false Metal-clad (Type MC) cable. 111.60-23 Section 111.60-23...-GENERAL REQUIREMENTS Wiring Materials and Methods § 111.60-23 Metal-clad (Type MC) cable. (a) Metal-clad (Type MC) cable permitted on board a vessel must be continuous corrugated metal-clad cable. (b)...

  12. 46 CFR 111.60-23 - Metal-clad (Type MC) cable.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 4 2013-10-01 2013-10-01 false Metal-clad (Type MC) cable. 111.60-23 Section 111.60-23...-GENERAL REQUIREMENTS Wiring Materials and Methods § 111.60-23 Metal-clad (Type MC) cable. (a) Metal-clad (Type MC) cable permitted on board a vessel must be continuous corrugated metal-clad cable. (b)...

  13. 46 CFR 111.60-23 - Metal-clad (Type MC) cable.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 4 2012-10-01 2012-10-01 false Metal-clad (Type MC) cable. 111.60-23 Section 111.60-23...-GENERAL REQUIREMENTS Wiring Materials and Methods § 111.60-23 Metal-clad (Type MC) cable. (a) Metal-clad (Type MC) cable permitted on board a vessel must be continuous corrugated metal-clad cable. (b)...

  14. French investigations of high burnup effect on LOCA thermomechanical behavior: Part 1. Experimental programmes in support of LOCA design methodologies

    SciTech Connect

    Waeckel, N.; Cauvin, R.; Lebuffe, C.

    1997-01-01

    Within the framework of Burn-Up extension request, EDF, FRAMATOME, CEA and IPSN have carried out experimental programmes in order to provide the design of fuel rods under LOCA conditions with relevant data. The design methods used in France for LOCA are based on standard Appendix K methodology updated to take into account some penalties related to the actual conditions of the Nuclear Power Plant. Best-Estimate assessments are used as well. Experimental programmes concern plastic deformation and burst behavior of advanced claddings (EDGAR) and thermal shock quenching behavior of highly irradiated claddings (TAGCIR). The former reveals the important role played by the {alpha}/{beta} transformation kinetics related to advanced alloys (Niobium alloys) and the latter the significative impact of hydrogen charged during in-reactor corrosion on oxidation kinetics and failure behavior in terms of cooling rates.

  15. JAEA Studies on High Burnup Fuel Behaviors during Reactivity-Initiated Accident and Loss-of-Coolant Accident

    SciTech Connect

    Fuketa, Toyoshi; Sugiyama, Tomoyuki; Nagase, Fumihisa; Suzuki, Motoe

    2007-07-01

    The objectives of fuel safety research program at Japan Atomic Energy Agency (JAEA) are; to evaluate adequacy of present safety criteria and safety margins; to provide a database for future regulation on higher burnup UO{sub 2} and MOX fuels, new cladding and pellets; and to provide reasonably mechanistic computer codes for regulatory application. The JAEA program is comprised of reactivity-initiated accident (RIA) studies including pulse-irradiation experiments in the NSRR and cladding mechanical tests, loss-of-coolant accident (LOCA) tests including integral thermal shock test and oxidation rate measurement, development and verification of computer codes FEMAXI-6 and RANNS, and so on. In addition to an overview of the fuel safety research at JAEA, most recent progresses in the RIA and LOCA tests programs and the codes development are described and discussed in the paper. (authors)

  16. High burnup effects in WWER fuel rods

    SciTech Connect

    Smirnov, V.; Smirnov, A.

    1996-03-01

    Since 1987 at the Research Institute of Atomic Reactors, the examinations of the WWER spent fuel assemblies has been carried out. These investigations are aimed to gain information on WWER spent fuel conditions in order to validate the fuel assemblies use during the 3 and 4 year fuel cycle in the WWER-440 and WWER-1000 units. At present time, the aim is to reach an average fuel burnup of 55 MWd/kgU. According to this aim, a new investigation program on the WWER spent fuel elements is started. The main objectives of this program are to study the high burnup effects and their influence on the WWER fuel properties. This paper presented the main statistical values of the WWER-440 and WWER-1000 reactors` fuel assemblies and their fragment parameters. Average burnup of fuel in the investigated fuel assemblies was in the range of 13 to 49.7 MWd/kgU. In this case, the numer of fuel cycles was from 1 to 4 during operation of the fuel assemblies.

  17. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    SciTech Connect

    Marshall, William BJ J

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

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

    SciTech Connect

    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.

  19. Effect of hydride orientation on fracture toughness of Zircaloy-4 cladding

    NASA Astrophysics Data System (ADS)

    Hsu, Hsiao-Hung; Tsay, Leu-Wen

    2011-01-01

    Hydrogen embrittlement is one of the major degradation mechanisms for high burnup fuel cladding during reactor service and spent fuel dry storage, which is related to the hydrogen concentration, morphology and orientation of zirconium hydrides. In this work, the J-integral values for X-specimens with different hydride orientations are measured to evaluate the fracture toughness of Zircaloy-4 (Zry-4) cladding. The toughness values for Zry-4 cladding with various percentages of radial hydrides are much smaller than those with circumferential hydrides only in the same hydrogen content level at 25 °C. The fractograghic features reveal that the crack path is influenced by the orientation of zirconium hydride. Moreover, the fracture toughness measurements for X-specimens at 300 °C are not sensitive to a variation in hydride orientation but to hydrogen concentration.

  20. Fuel modeling at high burn-up: recent development of the GERMINAL code

    NASA Astrophysics Data System (ADS)

    Melis, J.-C.; Piron, J.-P.; Roche, L.

    1993-09-01

    In the frame of research and development on fast breeder reactors fuels, CEA/DEC is developing the computer code GERMINAL to study fuel pin thermal and mechanical behaviour during steady-state and accidental conditions. The development of the GERMINAL 1 code is foreseen in two steps: (1) The GERMINAL 1-1 version which is presently delivered fully documented with a physical qualification guaranteed up to 8 at%. (2) The GERMINAL 1-2 version which, in addition to what is presently treated in GERMINAL 1-1, includes the treatment of high burn-up effects on the the fission gas release and the fuel-clad interface (called JOG). The validation of GERMINAL 1-2 is presently in progress and will include specific experiments (JOG tests) performed in the CABRI reactor.

  1. Laser powder technology for cladding and welding

    NASA Astrophysics Data System (ADS)

    Arnold, J.; Volz, R.

    1999-06-01

    Laser powder technology offers several advantages compared to conventional cladding and welding techniques and is attracting increasing industrial interest. The laser materials processing group of the German Aerospace Center at Stuttgart, Germany, is currently developing these new methods for application in industrial process engineering. Key areas of the work include the design and implementation of a modular working head that can be universally used for laser welding and surface treatment, the development of powder nozzles for cladding and welding, and the construction of new systems for special applications (e.g., for inner cladding). Some of these developments are described, as well as some important examples that highlight the potential of welding and surface treatment using laser powder techniques.

  2. Burnup concept for a long-life fast reactor core using MCNPX.

    SciTech Connect

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

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

    SciTech Connect

    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.

  4. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    NASA Astrophysics Data System (ADS)

    Makmal, T.; Aviv, O.; Gilad, E.

    2016-10-01

    A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections.

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

    SciTech Connect

    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.

  6. Benefits of actinide-only burnup credit for shutdown PWRs

    SciTech Connect

    Lancaster, D.; Fuentes, E.; Kang, C.; Rivard, D.

    1998-02-01

    Owners of PWRs that are shutdown prior to resolution of interim storage or permanent disposal issues have to make difficult decisions on what to do with their spent fuel. Maine Yankee is currently evaluating multiple options for spent fuel storage. Their spent fuel pool has 1,434 assemblies. In order to evaluate the value to a utility of actinide-only burnup credit, analysis of the number of canisters required with and without burnup credit was made. In order to perform the analysis, loading curves were developed for the Holtec Hi-Star 100/MPC-32. The MPC-32 is hoped to be representative of future burnup credit designs from many vendors. The loading curves were generated using the actinide-only burnup credit currently under NRC review. The canister was analyzed for full loading (32 assemblies) and with partial loadings of 30 and 28 assemblies. If no burnup credit is used the maximum capacity was assumed to be 24 assemblies. this reduced capacity is due to the space required for flux traps which are needed to sufficiently reduce the canister reactivity for the fresh fuel assumption. Without burnup credit the 1,343 assemblies would require 60 canisters. If all the fuel could be loaded into the 32 assembly canisters only 45 canisters would be required. Although the actinide-only burnup credit approach is very conservative, the total number of canisters required is only 47 which is only two short of the minimum possible number of canisters. The utility is expected to buy the canister and the storage overpack. A reasonable cost estimate for the canister plus overpack is $500,000. Actinide-only burnup credit would save 13 canisters and overpacks which is a savings of about $6.5 million. This savings is somewhat reduced since burnup credit requires a verification measurement of burnup. The measurement costs for these assemblies can be estimated as about $1 million. The net savings would be $5.5 million.

  7. Innovations in laser cladding and direct metal deposition

    NASA Astrophysics Data System (ADS)

    Brückner, Frank; Nowotny, Steffen; Leyens, Christoph

    2012-03-01

    The present paper reviews recent progress in productivity, precision and quality of laser-based cladding and additive layer manufacturing. Recently, we have demonstrated the great benefits obtained from induction assisted laser cladding. This novel hybrid technology combines high deposition rates with excellent cladding properties. Laser-based direct metal deposition is a novel concept for the fabrication of components and repair as well as geometrical surface modifications. Newly developed nozzle design allows focused powder spots to generate wall thicknesses of about 30 μm. An in-depth understanding of the processes and the resulting materials properties is key for the development of technically viable and economically reasonable customized solutions.

  8. Method and etchant to join Ag-clad BSSCO superconducting tape

    DOEpatents

    Balachandran, U.; Iyer, A.N.; Huang, J.Y.

    1999-03-16

    A method of removing a silver cladding from high temperature superconducting material clad in silver (HTS) is disclosed. The silver clad HTS is contacted with an aqueous solution of HNO{sub 3} followed by an aqueous solution of NH{sub 4}OH and H{sub 2}O{sub 2} for a time sufficient to remove the silver cladding from the superconducting material without adversely affecting the superconducting properties of the superconducting material. A portion of the silver cladding may be masked with a material chemically impervious to HNO{sub 3} and to a combination of NH{sub 4}OH and H{sub 2}O{sub 2} to preserve the Ag coating. A silver clad superconductor is disclosed, made in accordance with the method discussed. 3 figs.

  9. Method and etchant to join ag-clad BSSCO superconducting tape

    DOEpatents

    Balachandran, Uthamalingam; Iyer, Anand N.; Huang, Jiann Yuan

    1999-01-01

    A method of removing a silver cladding from high temperature superconducting material clad in silver (HTS) is disclosed. The silver clad HTS is contacted with an aqueous solution of HNO.sub.3 followed by an aqueous solution of NH.sub.4 OH and H.sub.2 O.sub.2 for a time sufficient to remove the silver cladding from the superconducting material without adversely affecting the superconducting properties of the superconducting material. A portion of the silver cladding may be masked with a material chemically impervious to HNO.sub.3 and to a combination of NH.sub.4 OH and H.sub.2 O.sub.2 to preserve the Ag coating. A silver clad superconductor is disclosed, made in accordance with the method discussed.

  10. Advanced ceramic cladding for water reactor fuel

    SciTech Connect

    Feinroth, H.

    2000-07-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of {approximately}60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies {ge}50% would be examined.

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

    NASA Astrophysics Data System (ADS)

    Metzger, Kathryn E.

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

  12. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    SciTech Connect

    Ade, Brian J; Marshall, William BJ J; Martinez-Gonzalez, Jesus S

    2015-05-01

    Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

  13. High Resolution Temperature Estimation During Laser Cladding of Stainless Steel

    NASA Astrophysics Data System (ADS)

    Devesse, Wim; De Baere, Dieter; Hinderdael, Michaël; Guillaume, Patrick

    Laser cladding is a technique that is used for the coating, repair and production of metallic parts. Material is added to the surface of the part by injecting a flow of powder into a melt pool that is created with a high power laser beam. When the beam scans the surface of the substrate, strong local heating and cooling results. A good knowledge of the temperature distribution history during the laser cladding process is vital to predict and optimize the material properties of the final part. This paper presents a contactless temperature measurement system with high temporal and spatial resolution based on a hyperspectral line camera. High temperature measurements were made during laser cladding of AISI 316L stainless steel. A good correlation is shown between the temperature measurements and microscope images taken after creation of the clad.

  14. A Simple Global View of Fuel Burnup

    NASA Astrophysics Data System (ADS)

    Sekimoto, Hiroshi

    2017-01-01

    Reactor physics and fuel burnup are discussed in order to obtain a simple global view of the effects of nuclear reactor characteristics to fuel cycle system performance. It may provide some idea of free thinking and overall vision, though it is still a small part of nuclear energy system. At the beginning of this lecture, governing equations for nuclear reactors are presented. Since the set of these equations is so big and complicated, it is simplified by imposing some extreme conditions and the nuclear equilibrium equation is derived. Some features of future nuclear equilibrium state are obtained by solving this equation. The contribution of a nucleus charged into reactor core to the system performance indexes such as criticality is worth for understanding the importance of each nuclide. It is called nuclide importance and can be evaluated by using the equations adjoint to the nuclear equilibrium equation. Examples of some importances and their application to criticalily search problem are presented.

  15. Properties of multilayer coatings produced by coaxial laser cladding

    NASA Astrophysics Data System (ADS)

    Petrovskiy, V. N.; Bykovskiy, D. P.; Dzhumaev, P. S.; Polskiy, V. I.; Prokopova, N. M.; Chirikov, S. N.

    2016-09-01

    This article contains results of the study of multilayer coatings produced by laser cladding on the substrate steel 34HMA using iron based powder PR-10R6M5 as the filler material. The coatings were produced with consistent application of the tracks with fixed overlapping. The dependencies between the characteristics of tracks and the technological mode of deposition were revealed. Properties of coatings were determined for various overlapping of tracks and directions of the cladding layers.

  16. Characterization of spent fuel approved testing material: ATM-106

    SciTech Connect

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thornhill, C.K.

    1988-10-01

    The characterization data obtained to date are described for Approved Testing Material (ATM)-106 spent fuel from Assembly BT03 of pressurized-water reactor Calvert Cliffs No. 1. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory 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 (OCWRM) program. ATM-106 consists of 20 full-length irradiated fuel rods with rod-average burnups of about 3700 GJ/kgM (43 MWd/kgM) and expected fission gas release of /approximately/10%. Characterization data include (1) 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) calculated nuclide inventories and radioactivities in the fuel and cladding; and (6) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel rod are being conducted and will be included in planned revisions of this report. 12 refs., 110 figs., 81 tabs.

  17. Characterization of spent fuel approved testing material: ATM-103

    SciTech Connect

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1988-04-01

    The characterization data obtained to date are described for Approved Testing Material (ATM)-103, which is spent fuel from Assembly D101 of pressurized-water reactor Calvert Cliffs, No. 1. This report is one in a series being written by the Materials Characterization Center (MCC) 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 nuclear waste repository program. ATM-103 consists of 176 full-length irradiated fuel rods with rod-average burnups of about 2600 GJ/kgM (30 MWd/kgM) and less than 1% fission gas release. Characterization data include 1) 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 fuels 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. 10 refs., 103 figs., 63 tabs.

  18. Investigation of semiconductor clad optical waveguides

    NASA Technical Reports Server (NTRS)

    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.

  19. Implementation and evaluation of fuel creep using advanced light-water reactor materials in FRAPCON 3.5

    NASA Astrophysics Data System (ADS)

    Carroll, Spencer

    As current reactors approach the end of their operable lifetime, new reactors are needed if nuclear power is to continue being generated in the United States. Some utilities have already began construction on newer, more advanced LWR reactors, which use the same fuel as current reactors and have a similar but updated design. Others are researching next generation (GEN-IV) reactors which have new designs that utilize alternative fuel, coolants and other reactor materials. Many of these alternative fuels are capable of achieving higher burnups and are designed to be more accident tolerant than the currently used UO2 fuel. However, before these new materials can be used, extensive research must be done in order to obtain a detailed understanding of how the new fuels and other materials will interact. New fuels, such as uranium nitride (UN) and uranium carbide (UC) have several advantages over UO2, such as increased burnup capabilities and higher thermal conductivities. However, there are issues with each that prevent UC and UN from being used as direct replacements for UO2. Both UC and UN swell at a significantly higher rate than UO2 and neither fuel reacts favorably when exposed to water. Due to this, UC and UN are being considered more for GEN-IV reactors that use alternative coolant rather than for current LWRs. In an effort to increase accident tolerance, silicon carbide (SiC) is being considered for use as an alternative cladding. The high strength, high melting point and low oxidation of SiC make it an attractive cladding choice, especially in an accident scenario. However, as a ceramic, SiC is not ductile and will not creep outwards upon pellet-clad mechanical interaction (PCMI) which can cause a large build up in interfacial pressure. In order to understand the interaction between the high swelling fuels and unyielding SiC cladding, data on the properties and behaviors of these materials must be gathered and incorporated into FRAPCON. FRAPCON is a fuel

  20. Aerogel-clad optical fiber

    DOEpatents

    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.

  1. Aerogel-clad optical fiber

    DOEpatents

    Sprehn, G.A.; Hrubesh, L.W.; Poco, J.F.; Sandler, P.H.

    1997-11-04

    An optical fiber is surrounded by an aerogel cladding. For a low density aerogel, the index of refraction of the aerogel is close to that of air, which provides a high numerical aperture to the optical fiber. Due to the high numerical aperture, the aerogel clad optical fiber has improved light collection efficiency. 4 figs.

  2. Triton burnup measurements in KSTAR using a neutron activation system

    NASA Astrophysics Data System (ADS)

    Jo, Jungmin; Cheon, MunSeong; Kim, Jun Young; Rhee, T.; Kim, Junghee; Shi, Yue-Jiang; Isobe, M.; Ogawa, K.; Chung, Kyoung-Jae; Hwang, Y. S.

    2016-11-01

    Measurements of the time-integrated triton burnup for deuterium plasma in Korea Superconducting Tokamak Advanced Research (KSTAR) have been performed following the simultaneous detection of the d-d and d-t neutrons. The d-d neutrons were measured using a 3He proportional counter, fission chamber, and activated indium sample, whereas the d-t neutrons were detected using activated silicon and copper samples. The triton burnup ratio from KSTAR discharges is found to be in the range 0.01%-0.50% depending on the plasma conditions. The measured burnup ratio is compared with the prompt loss fraction of tritons calculated with the Lorentz orbit code and the classical slowing-down time. The burnup ratio is found to increase as plasma current and classical slowing-down time increase.

  3. Laser cladding: repairing and manufacturing metal parts and tools

    NASA Astrophysics Data System (ADS)

    Sexton, Leo

    2003-03-01

    Laser cladding is presently used to repair high volume aerospace, automotive, marine, rail or general engineering components where excessive wear has occurred. It can also be used if a one-off high value component is either required or has been accidentally over-machined. The ultimate application of laser cladding is to build components up from nothing, using a laser cladding system and a 3D CAD drawing of the component. It is thus emerging that laser cladding can be classified as a special case of Rapid Prototyping (RP). Up to this point in time RP was seen, and is still seen, as in intermediately step between the design stage of a component and a finished working product. This can now be extended so that laser cladding makes RP a one-stop shop and the finished component is made from tool-steel or some alloy-base material. The marriage of laser cladding with RP is an interesting one and offers an alternative to traditional tool builders, re-manufacturers and injection mould design/repair industries. The aim of this paper is to discuss the emergence of this new technology, along with the transference of the process out of the laboratory and into the industrial workplace and show it is finding its rightful place in the manufacturing/repair sector. It will be shown that it can be used as a cost cutting, strategic material saver and consequently a green technology.

  4. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    SciTech Connect

    Wagner, J.C.

    2001-08-02

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses.

  5. Update on the PWR axial burnup profile database

    SciTech Connect

    Cacciapouti, R.F.; Volkinburg, S.V.

    1995-12-01

    A pressurized water reactor database was developed to evaluate the axial burnup profiles of various reactor types. The data showed that the various types exhibit similar behavior, especially at the top and bottom of the assembly. From the existing data, bounding axial burnup profiles can be developed to envelope the various pressurized water reactor assembly deigns. The database encompasses most of the PWR fuel designs and contains sufficient data to provide reliable statistics.

  6. Designing Critical Experiments in Support of Full Burnup Credit

    SciTech Connect

    Mueller, Don; Roberts, Jeremy A

    2008-01-01

    Burnup credit is the process of accounting for the negative reactivity due to fuel burnup and generation of parasitic absorbers over fuel assembly lifetime. For years, the fresh fuel assumption was used as a simple bound in criticality work for used fuel storage and transportation. More recently, major actinides have been included [1]. However, even this yields a highly conservative estimate in criticality calculations. Because of the numerous economical benefits including all available negative reactivity (i.e., full burnup credit) could provide [2], it is advantageous to work toward full burnup credit. Unfortunately, comparatively little work has been done to include non-major actinides and other fission products (FP) in burnup credit analyses due in part to insufficient experimental data for validation of codes and nuclear data. The Burnup Credit Criticality Experiment (BUCCX) at Sandia National Laboratory was a set of experiments with {sup 103}Rh that have relevance for burnup credit [3]. This work uses TSUNAMI-3D to investigate and adjust a BUCCX model to match isotope-specific, energy-dependent k{sub eff} sensitivity profiles to those of a representative high-capacity cask model (GBC-32) [4] for each FP of interest. The isotopes considered are {sup 149}Sm, {sup 143}Nd, {sup 103}Rh, {sup 133}Cs, {sup 155}Gd, {sup 152}Sm, {sup 99}Tc, {sup 145}Nd, {sup 153}Eu, {sup 147}Sm, {sup 109}Ag, {sup 95}Mo, {sup 150}Sm, {sup 101}Ru, and {sup 151}Eu. The goal is to understand the biases and bias uncertainties inherent in nuclear data, and ultimately, to apply these in support of full burnup credit.

  7. Explosion Clad for Upstream Oil and Gas Equipment

    SciTech Connect

    Banker, John G.; Massarello, Jack; Pauly, Stephane

    2011-01-17

    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 CO{sub 2} and/or H{sub 2}S 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.

  8. Development of the recovery technology for nickel superalloy blades of the aircraft engine by laser cladding

    NASA Astrophysics Data System (ADS)

    Bykovskiy, D. P.; Petrovskiy, V. N.; Polskiy, V. I.; Chirikov, S. N.; Dzhumaev, P. S.

    2016-09-01

    Development of cladding modes was performed with a superalloy nickel based powder on a flat substrate from material identical to compressor and turbine blades. Cross sections were made, and a visual inspection of the shape and the quality of the clad track as well as themetallographic analysis were performed. Microhardness of the deposition zone, chemical composition of the base, cladded metals, and the heat affected zone were determined.

  9. Thermal property change of MOX and UO2 irradiated up to high burnup of 74 GWd/t

    NASA Astrophysics Data System (ADS)

    Nakae, Nobuo; Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro; Kurematsu, Shigeru; Kosaka, Yuji; Yoshino, Aya; Kitagawa, Takaaki

    2013-09-01

    Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO2 fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO2. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO2 is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO2 at high burnup under the condition that the pellet-cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO2 before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO2. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

  10. Oxygen potential in the rim region of high burnup UO 2 fuel

    NASA Astrophysics Data System (ADS)

    Matzke, Hj.

    1994-01-01

    Small specimens from the rim region (< 200 μm from the fuel surface) of a UO 2 fuel rod with an average burnup of 7.6% FIMA were analysed in a miniaturized galvanic cell to determine their oxygen potential Δ Ḡ(O 2) . These fuel pieces represented the porous rim structure with very small grains known to be formed near the periphery of high burnup UO 2 fuel pellets. The oxygen potential of the rim material was very low, corresponding to that of unirradiated stoichiometric UO 2, or to that of slightly substoichiometric UO 2 containing rare earth fission products. No indication of oxidation due to fission was found, though most fission was that of Pu. Measurements on pieces from the inner, unrestructured fuel showed somewhat higher oxygen potentials corresponding to those of very slightly substoichiometric fuel if allowance is made for the incorporation of rare earths. These results are in contrast to some generally accepted ideas of burnup effects, and the possible reasons and implications are discussed.

  11. Preparation of higher-actinide burnup and cross section samples. [LMFBR

    SciTech Connect

    Adair, H.L.; Kobisk, E.H.; Quinby, T.C.; Thomas, D.K.; Dailey, J.M.

    1981-01-01

    A joint research program involving the United States and the United Kingdom was instigated about four years ago for the purpose of studying burnup of higher actinides using in-core irradiation in the fast reactor at Dounreay, Scotland. Simultaneously, determination of cross sections of a wide variety of higher actinide isotopes was proposed. Coincidental neutron flux and energy spectral measurements were to be made using vanadium encapsulated dosimetry materials in the immediate region of the burnup and cross section samples. The higher actinide samples chosen for the burnup study were /sup 241/Am and /sup 244/Cm in the forms of Am/sub 2/O/sub 3/, Cm/sub 2/O/sub 3/, and Am/sub 6/ Cm(RE)/sub 7/O/sub 21/, where (RE) represents a mixture of lanthanide sesquioxides. It is the purpose of this paper to describe technology development and its application in the preparation of the fuel specimens and the cross section specimens that are being used in this cooperative program.

  12. Spatial mode-selective waveguide with hyperbolic cladding

    NASA Astrophysics Data System (ADS)

    Tang, Y.; Xi, Z.; Xu, M.; Bäumer, S.; Adam, A. J. L.; Urbach, H. P.

    2016-09-01

    Hyperbolic Meta-Materials~(HMMs) are anisotropic materials with permittivity tensor that has both positive and negative eigenvalues. Here we report that by using a type II HMM as cladding material, a waveguide which only supports higher order modes can be achieved, while the lower order modes become leaky and are absorbed in the HMM cladding. This counter intuitive property can lead to novel application in optical communication and photonic integrated circuit. The loss in our HMM-Insulator-HMM~(HIH) waveguide is smaller than that of similar guided mode in a Metal-Insulator-Metal~(MIM) waveguide.

  13. HIP clad nickel base Alloy 625 for deep sour wells

    SciTech Connect

    Uhl, W.K.; Pendley, M.R.

    1984-05-01

    The hot isostatic pressing (HIP) process was used to clad nickel base Alloy 625 to AISI 4130 low alloy steel. The performance of the HIP clad material in the corrosive environment characteristic of deep, sour oil and gas wells was evaluated in laboratory tests. Included in the test program were NACE TM-01-77 sulfide stress cracking tests, chloride stress corrosion cracking tests in boiling MgCl /SUB 2'/ , and pitting and crevice corrosion tests. The HIP clad 625 performed excellently, displaying essentially the same corrosion resistance as wrought 625. Specifically the HIP clad 625 resisted sulfide stress cracking at applied stresses as high as 120% of yield strength and resisted chloride stress corrosion cracking at stresses exceeding 100% of yield. The HIP clad 625 also displayed immunity to pitting and crevice corrosion, with corrosion rates of <0.025 mm/y (1 mil/y). The 4130 base metal, however, was attacked severly in all tests. SEM/EDX analysis of the 625/4130 interface demonstrated that dilution of the cladding by the base metal was essentially eliminated.

  14. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    SciTech Connect

    Su'ud, Zaki; Sekimoto, H.

    2014-09-30

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  15. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    NASA Astrophysics Data System (ADS)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

  16. Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

    SciTech Connect

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

    2016-01-01

    Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermal hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the

  17. Special optical fiber for temperature sensing based on cladding-mode resonance.

    PubMed

    Pang, Fufei; Xiang, Wenchao; Guo, Hairun; Chen, Na; Zeng, Xianglong; Chen, Zhenyi; Wang, Tingyun

    2008-08-18

    A fiber-optic temperature sensor by using a multi-cladding special fiber is presented. It works on the basis of leaky mode resonance from fiber core to outer cladding. With the thin-thickness inner cladding, the cladding mode is strongly excited and the resonant spectrum is very sensitive to the refractive index variation of coating material. By coating the special fiber with temperature-sensitive silicone, the temperature response was investigated experimentally from -20 degrees C to 80 degrees C. The results show high temperature sensitivity (240 pm/degrees C at 20 degrees C) and good repeatability.

  18. Use of burnup credit for transportation and storage

    SciTech Connect

    Sanders, T.L.; Ewing, R.I. ); Lake, W.H. )

    1991-01-01

    Burnup credit is the application of the effects of fuel burnup to nuclear criticality design. When burnup credit is considered in the design of storage facilities and transportation casks for spent fuel, the objectives are to reduce the requirements for storage space and to increase the payload of casks with acceptable nuclear criticality safety margins. The spent-fuel carrying capacities of previous-generation transport casks have been limited primarily by requirements to remove heat and/or to provide shielding. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced significantly. Thus a considerable weight margin is available to the designer for increasing the payload capacity. One method to achieve an increase in capacity is to reduce fuel assembly spacing. The amount of reduction in assembly spacing is limited by criticality and fuel support structural concerns. The optimum fuel assembly spacing provides the maximum cask loading within a basket that has adequate criticality control and sufficient structural integrity for regulatory accident scenarios. The incorporation of burnup credit in cask designs could result in considerable benefits in the transport of spent fuel. The acceptance of burnup credit for the design of transport casks depends on the resolution of system safety issues and the uncertainties that affect the determination of criticality safety margins. The remainder of this report will examine these issues and the integrated approach under way to resolve them. 20 refs., 2 figs.

  19. High power cladding light strippers

    NASA Astrophysics Data System (ADS)

    Wetter, Alexandre; Faucher, Mathieu; Sévigny, Benoit

    2008-02-01

    The ability to strip cladding light from double clad fiber (DCF) fibers is required for many different reasons, one example is to strip unwanted cladding light in fiber lasers and amplifiers. When removing residual pump light for example, this light is characterized by a large numerical aperture distribution and can reach power levels into the hundreds of watts. By locally changing the numerical aperture (N.A.) of the light to be stripped, it is possible to achieve significant attenuation even for the low N.A. rays such as escaped core modes in the same device. In order to test the power-handling capability of this device, one hundred watts of pump and signal light is launched from a tapered fusedbundle (TFB) 6+1x1 combiner into a high power-cladding stripper. In this case, the fiber used in the cladding stripper and the output fiber of the TFB was a 20/400 0.06/0.46 N.A. double clad fiber. Attenuation of over 20dB in the cladding was measured without signal loss. By spreading out the heat load generated by the unwanted light that is stripped, the package remained safely below the maximum operating temperature internally and externally. This is achieved by uniformly stripping the energy along the length of the fiber within the stripper. Different adhesive and heat sinking techniques are used to achieve this uniform removal of the light. This suggests that these cladding strippers can be used to strip hundreds of watts of light in high power fiber lasers and amplifiers.

  20. Properties and features of structure formation CuCr-contact alloys in electron beam cladding

    SciTech Connect

    Durakov, Vasiliy G.; Dampilon, Bair V. E-mail: gnusov@rambler.ru; Gnyusov, Sergey F. E-mail: gnusov@rambler.ru

    2014-11-14

    The microstructure and properties of the contact CuCr alloy produced by electron-beam cladding have been investigated. The effect of the electron beam cladding parameters and preheating temperature of the base metal on the structure and the properties of the coatings has been determined. The bimodal structure of the cladding coating has been established. The short circuit currents tests have been carried out according to the Weil-Dobke synthetic circuit simulating procedure developed for vacuum circuit breakers (VCB) test in real electric circuits. Test results have shown that the electron beam cladding (EBC) contact material has better breaking capacity than that of commercially fabricated sintered contact material. The application of the technology of electron beam cladding for production of contact material would significantly improve specific characteristics and reliability of vacuum switching equipment.

  1. Radiochemical analyses of several spent fuel Approved Testing Materials

    SciTech Connect

    Guenther, R.J.; Blahnik, D.E.; Wildung, N.J.

    1994-09-01

    Radiochemical characterization data are described for UO{sub 2} and UO{sub 2} plus 3 wt% Gd{sub 2}O{sub 3} commercial spent nuclear fuel taken from a series of Approved Testing Materials (ATMs). These full-length nuclear fuel rods include MLA091 of ATM-103, MKP070 of ATM-104, NBD095 and NBD131 of ATM-106, and ADN0206 of ATM-108. ATMs 103, 104, and 106 were all irradiated in the Calvert Cliffs Nuclear Power Plant (Reactor No.1), a pressurized-water reactor that used fuel fabricated by Combustion Engineering. ATM-108 was part of the same fuel bundle designed as ATM-105 and came from boiling-water reactor fuel fabricated by General Electric and irradiated in the Cooper Nuclear Power Plant. Rod average burnups and expected fission gas releases ranged from 2,400 to 3,700 GJ/kgM. (25 to 40 Mwd/kgM) and from less than 1% to greater than 10%, respectively, depending on the specific ATM. The radiochemical analyses included uranium and plutonium isotopes in the fuel, selected fission products in the fuel, fuel burnup, cesium and iodine on the inner surfaces of the cladding, {sup 14}C in the fuel and cladding, and analyses of the gases released to the rod plenum. Supporting examinations such as fuel rod design and material descriptions, power histories, and gamma scans used for sectioning diagrams are also included. These ATMs were examined as part of the Materials Characterization Center Program conducted at Pacific Northwest Laboratory provide a source of well-characterized spent fuel for testing in support of the US Department of Energy Office of Civilian Radioactive Waste Management Program.

  2. An empirical formulation to describe the evolution of the high burnup structure

    NASA Astrophysics Data System (ADS)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-01

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  3. Method and system for edge cladding of laser gain media

    DOEpatents

    Bayramian, Andrew James; Caird, John Allyn; Schaffers, Kathleen Irene

    2014-03-25

    A gain medium operable to amplify light at a gain wavelength and having reduced transverse ASE includes an input surface and an output surface opposing the input surface. The gain medium also includes a central region including gain material and extending between the input surface and the output surface along a longitudinal optical axis of the gain medium. The gain medium further includes an edge cladding region surrounding the central region and extending between the input surface and the output surface along the longitudinal optical axis of the gain medium. The edge cladding region includes the gain material and a dopant operable to absorb light at the gain wavelength.

  4. Characterization of Hydrogen Content in ZIRCALOY-4 Nuclear Fuel Cladding

    NASA Astrophysics Data System (ADS)

    Pfeif, E. A.; Lasseigne, A. N.; Krzywosz, K.; Mader, E. V.; Mishra, B.; Olson, D. L.

    2010-02-01

    Assessment of hydrogen uptake of underwater nuclear fuel clad and component materials will enable improved monitoring of fuel health. Zirconium alloys are used in nuclear reactors as fuel cladding, fuel channels, guide tubes and spacer grids, and are available for inspection in spent fuel pools. With increasing reactor exposure zirconium alloys experience hydrogen ingress due to neutron interactions and water-side corrosion that is not easily quantified without destructive hot cell examination. Contact and non-contact nondestructive techniques, using Seebeck coefficient measurements and low frequency impedance spectroscopy, to assess the hydrogen content and hydride formation within zircaloy 4 material that are submerged to simulate spent fuel pools are presented.

  5. Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor

    SciTech Connect

    G. S. Chang

    2005-08-01

    A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is capable of handling a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) lobe powers, and irradiation time intervals. MCWO processes user input that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN-2, and data process module calculations are output in succession as MCWO executes. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN-2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN-2 back to MCNP in a repeated, cyclic fashion. The basic requirements of MCWO are a working MCNP input file and some additional input parameters; all interaction with ORIGEN-2 as well as other calculations are performed by CMO. This paper presents the MCWO-calculated results for the Reduced Enrichment Research and Test Reactor (RERTR) experiments RERTR-1 and RERTR-2 as well as the Weapons-Grade Mixed Oxide (WG-MOX) fuel testing in ATR. Calculations performed for the WG-MOX test irradiation, which is managed by the Oak Ridge National Laboratory (ORNL), supports the DOE Fissile Materials Disposition Program (FMDP). The MCWO-calculated results are compared with measured data.

  6. High-burnup fuel and the impact on fuel management

    SciTech Connect

    Cacciapouti, R.J.; Weader, R.J.

    1996-12-31

    Competition in the electric utility industry has forced utilities to reduce cost. For a nuclear utility, this means a reduction of both the nuclear fuel cost and the operating and maintenance cost. To this extent, utilities are pursuing longer cycles. To reduce the nuclear fuel cost, utilities are trying to reduce batch size while increasing cycle length. Yankee Atomic Electric Company has performed a number of fuel cycle studies to optimize both batch size and cycle length; however, certain burnup-related constraints are encountered. As a result of these circumstances, longer fuel cycles make it increasingly difficult to simultaneously meet the burnup-related fuel design constraints and the technical specification limits. Longer cycles require fuel assemblies to operate for longer times at relatively high power. If utilities continue to pursue longer cycles to help reduce nuclear fuel cost, changes may need to be made to existing fuel burnup limits.

  7. The National Shipbuilding Research Program, 1991 Ship Production Symposium Proceedings: Paper No. IVB-1: Using Fiber Optics for Laser Cladding

    DTIC Science & Technology

    1991-09-01

    6) which are used to hardface valve seats. The weld cladding of the valve seats with such materials can become very expensive if repeated cladding...submarine components which require hardfacing and corrosion protection. The laser cladding operation consists of depositing a powder directly in front of...developed and optimized for the cobalt- tungsten hardfacing material on 316L and 416 stainless steels and 4140 and DH-36 carbon steels. Metallurgical

  8. Blind prediction exercise on modeling of PHWR fuel at extended burnup

    NASA Astrophysics Data System (ADS)

    Sah, D. N.; Viswanathan, U. K.; Viswanadham, C. S.; Unnikrishnan, K.; Rath, B. N.

    2008-12-01

    A blind prediction exercise was organised on Indian Pressurised Heavy Water Reactor (PHWR) fuel to investigate the predictive capability of existing codes for their application at extended burnup and to identify areas of improvement. The blind problem for this exercise was based on a PHWR fuel bundle irradiated in Kakrapar Atomic Power Station-I (KAPS-I) up to about 15 000 MWd/tU and subjected to detailed post-irradiation examination (PIE) in the hot cells facility at BARC. Eleven computer codes from seven countries participated in this exercise. The participants provided blind predictions of fuel temperature, fission gas release, internal gas pressure and other performance parameters for the fuel pins. The predictions were compared with the experimental PIE data which included fuel temperature derived from fuel restructuring, fission gas release measured by fuel pin puncturing, internal gas pressure in pin, cladding oxidation and fuel microstructural data. The details of the blind problem and an analysis of the results of blind predictions by the codes vis-à-vis measured data are provided in this paper.

  9. Intercode Advanced Fuels and Cladding Comparison Using BISON, FRAPCON, and FEMAXI Fuel Performance Codes

    NASA Astrophysics Data System (ADS)

    Rice, Aaren

    As part of the Department of Energy's Accident Tolerant Fuels (ATF) campaign, new cladding designs and fuel types are being studied in order to help make nuclear energy a safer and more affordable source for power. This study focuses on the implementation and analysis of the SiC cladding and UN, UC, and U3Si2 fuels into three specific nuclear fuel performance codes: BISON, FRAPCON, and FEMAXI. These fuels boast a higher thermal conductivity and uranium density than traditional UO2 fuel which could help lead to longer times in a reactor environment. The SiC cladding has been studied for its reduced production of hydrogen gas during an accident scenario, however the SiC cladding is a known brittle and unyielding material that may fracture during PCMI (Pellet Cladding Mechanical Interaction). This work focuses on steady-state operation with advanced fuel and cladding combinations. By implementing and performing analysis work with these materials, it is possible to better understand some of the mechanical interactions that could be seen as limiting factors. In addition to the analysis of the materials themselves, a further analysis is done on the effects of using a fuel creep model in combination with the SiC cladding. While fuel creep is commonly ignored in the traditional UO2 fuel and Zircaloy cladding systems, fuel creep can be a significant factor in PCMI with SiC.

  10. Isotopic biases for actinide-only burnup credit

    SciTech Connect

    Rahimi, M.; Lancaster, D.; Hoeffer, B.; Nichols, M.

    1997-04-01

    The primary purpose of this paper is to present the new methodology for establishing bias and uncertainty associated with isotopic prediction in spent fuel assemblies for burnup credit analysis. The analysis applies to the design of criticality control systems for spent fuel casks. A total of 54 spent fuel samples were modeled and analyzed using the Shielding Analyses Sequence (SAS2H). Multiple regression analysis and a trending test were performed to develop isotopic correction factors for 10 actinide burnup credit isotopes. 5 refs., 1 tab.

  11. Power excursion analysis for BWR`s at high burnup

    SciTech Connect

    Diamond, D.J.; Neymoith, L.; Kohut, P.

    1996-03-01

    A study has been undertaken to determine the fuel enthalpy during a rod drop accident and during two thermal-hydraulic transients. The objective was to understand the consequences to high burnup fuel and the sources of uncertainty in the calculations. The analysis was done with RAMONA-4B, a computer code that models the neutron kinetics throughout the core along with the thermal-hydraulics in the core, vessel, and steamline. The results showed that the maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important parameters in each of these categories are discussed in the paper.

  12. Findings of an international study on burnup credit

    SciTech Connect

    Brady, M.C.; Takano, M.; Okuno, H.; DeHart, M.D.; Nouri, A.; Sartori, E.

    1996-08-01

    Findings from a four year study by an international benchmarking group in the comparison of computational methods for evaluating burnup credit in criticality safety analyses are presented in this paper. Approximately 20 participants from 11 countries have provided results for most problems. Four detailed benchmark problems for Pressurized Water Reactor (PWR) fuel have been completed and are summarized in this paper. Preliminary results from current work addressing burnup credit for Boiling Water Reactor (BWR) fuel will also be discussed as well as planned activities for additional benchmarks including Mixed-Oxide (MOX) fuels, subcritical benchmarks, international databases, and other activities.

  13. Acoustic emission monitoring of tensile testing of corroded and un-corroded clad aluminum 2024-T3 and characterization of effects of corrosion on AE source events and material tensile properties

    NASA Astrophysics Data System (ADS)

    Okafor, A. Chukwujekwu; Natarajan, Shridhar

    2014-02-01

    Corrosion damage affects structural integrity and deteriorates material properties of aluminum alloys in aircraft structures. Acoustic Emission (AE) is an effective nondestructive evaluation (NDE) technique for monitoring such damages and predicting failure in large structures of an aircraft. For successful interpretation of data from AE monitoring, sources of AE and factors affecting it need to be identified. This paper presents results of AE monitoring of tensile testing of corroded and un-corroded clad Aluminum 2024-T3 test specimens, and characterization of the effects of strain-rate and corrosion damage on material tensile properties and AE source events. Effect of corrosion was studied by inducing corrosion in the test specimens by accelerated corrosion testing in a Q-Fog accelerated corrosion chamber for 12 weeks. Eight (8) masked dog-bone shaped specimens were placed in the accelerated corrosion chamber at the beginning of the test. Two (2) dog-bone shaped specimens were removed from the corrosion chamber after exposure time of 3, 6, 9, and 12 weeks respectively, and subjected to tension testing till specimen failure along with AE monitoring, as well as two (2) reference samples not exposed to corrosion. Material tensile properties (yield strength, ultimate tensile strength, toughness, and elongation) obtained from tension test and AE parameters obtained from AE monitoring were analyzed and characterized. AE parameters increase with increase in exposure period of the specimens in the corrosive environment. Aluminum 2024-T3 is an acoustically silent material during tensile deformation without any damage. Acoustic emission events increase with increase of corrosion damage and with increase in strain rate above a certain value. Thus AE is suitable for structural health monitoring of corrosion damage. Ultimate tensile strength, toughness and elongation values decrease with increase of exposure period in corrosion chamber.

  14. MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION

    SciTech Connect

    Sternat, M.; Nichols, T.

    2011-06-09

    Reactor burnup or depletion codes are used thoroughly in the fields of nuclear forensics and nuclear safeguards. Two common codes include MONTEBURNS and MCNPX/CINDER. These are Monte-Carlo depletion routines utilizing MCNP for neutron transport calculations and either ORIGEN or CINDER for burnup calculations. Uncertainties exist in the MCNP steps, but this information is not passed to the depletion calculations or saved. To quantify this transport uncertainty and determine how it propagates between burnup steps, a statistical analysis of multiple repeated depletion runs is performed. The reactor model chosen is the Oak Ridge Research Reactor (ORR) in a single assembly, infinite lattice configuration. This model was burned for a 150 day cycle broken down into three steps. The output isotopics as well as effective multiplication factor (k-effective) were tabulated and histograms were created at each burnup step using the Scott Method to determine the bin width. The distributions for each code are a statistical benchmark and comparisons made. It was expected that the gram quantities and k-effective histograms would produce normally distributed results since they were produced from a Monte-Carlo routine, but some of the results appear to not. Statistical analyses are performed using the {chi}{sup 2} test against a normal distribution for the k-effective results and several isotopes including {sup 134}Cs, {sup 137}Cs, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 238}Pu, {sup 239}Pu, and {sup 240}Pu.

  15. Multi-clad black display panel

    DOEpatents

    Veligdan, James T.; Biscardi, Cyrus; Brewster, Calvin

    2002-01-01

    A multi-clad black display panel, and a method of making a multi-clad black display panel, are disclosed, wherein a plurality of waveguides, each of which includes a light-transmissive core placed between an opposing pair of transparent cladding layers and a black layer disposed between transparent cladding layers, are stacked together and sawed at an angle to produce a wedge-shaped optical panel having an inlet face and an outlet face.

  16. Need for higher fuel burnup at the Hatch Plant

    SciTech Connect

    Beckhman, J.T.

    1996-03-01

    Hatch is a BWR 4 and has been in operation for some time. The first unit became commercial about 1975. Obtaining higher burnups, or higher average discharge exposures, is nothing new at Hatch. Since we have started, the discharge exposure of the plant has increased. Now, of course, we are not approaching the numbers currently being discussed but, the average discharge exposure has increased from around 20,000 MWD/MTU in the early to mid-1980s to 34,000 MWD/MTU in 1994, I am talking about batch average values. There are also peak bundle and peak rod values. You will have to make the conversions if you think in one way or the other because I am talking in batch averages. During Hatch`s operating history we have had some problems with fuel failure. Higher burnup fuel raises a concern about how much fuel failure you are going to have. Fuel failure is, of course, an economic issue with us. Back in the early 1980s, we had a problem with crud-induced localized corrosion, known as CILC. We have gotten over that, but we had some times when it was up around 27 fuel failures a year. That is not a pleasant time to live through because it is not what you want from an economic viewpoint or any other. We have gotten that down. We have had some fuel failures recently, but they have not been related to fuel burnup or to corrosion. In fact, the number of failures has decreased from the early 1980s to the 90s even though burnup increased during that time. The fuel failures are more debris-related-type failures. In addition to increasing burnups, utilities are actively evaluating or have already incorporated power uprate and longer fuel cycles (e.g., 2-year cycles). The goal is to balance out the higher power density, longer cycles, higher burnup, and to have no leakers. Why do we as an industry want to have higher burnup fuel? That is what I want to tell you a little bit about.

  17. Iridium alloy Clad Vent Set manufacturing qualification studies

    NASA Astrophysics Data System (ADS)

    Ulrich, George B.

    Metallurgical qualification studies to demonstrate the manufacturing readiness of the iridium alloy Clad Vent Set (CVS) for the General Purpose Heat Source program at the Oak Ridge Y-12 Plant are described. Microstructural data for various materials/test conditions are presented.

  18. Core materials development for the fuel cycle R&D program

    NASA Astrophysics Data System (ADS)

    Maloy, S. A.; Toloczko, M.; Cole, J.; Byun, T. S.

    2011-08-01

    The Fuel Cycle Research and Development program is investigating methods of burning minor actinides in a transmutation fuel. One of the challenges of achieving this goal is to develop fuels capable of reaching extreme burnup levels (e.g. 40%). To achieve such high burnup levels' fast reactor core materials (cladding and duct) must be able to withstand very high doses (>300 dpa design goal) while in contact with the coolant and the fuel. Thus, these materials must withstand radiation effects that promote low temperature embrittlement, radiation induced segregation, high temperature helium embrittlement, swelling, accelerated creep, corrosion with the coolant, and chemical interaction with the fuel (FCCI). To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Test specimens of ferritic/martensitic alloys (T91/HT-9) previously irradiated in the FFTF reactor up to 210 dpa at a temperature range of 350-750 °C are presently being tested. This includes analysis of a duct made of HT-9 after irradiation to a total dose of 155 dpa at temperatures from 370 to 510 °C. Compact tension, charpy and tensile specimens have been machined from this duct and mechanical testing as well as SANS and Mossbauer spectroscopy are currently being performed. Initial results from compression testing and Charpy testing reveal a strong increase in yield stress (˜400 MPa) and a large increase in DBTT (up to 230 °C) for specimens irradiated at 383 °C to a dose of 28 dpa. Less hardening and a smaller increase in DBTT was observed for specimens irradiated at higher temperatures up to 500 °C. Advanced radiation tolerant materials are also being developed to enable the desired extreme fuel burnup levels. Specifically, coatings are being developed to minimize FCCI, and research is underway to fabricate large heats of radiation tolerant oxide dispersion steels with homogeneous oxide

  19. Comparison of PWR - Burnup calculations with SCALE 5.0/TRITON other burnup codes and experimental results

    SciTech Connect

    Oberle, P.; Broeders, C. H. M.; Dagan, R.

    2006-07-01

    The increasing tendency towards fuel lifetime extension in thermal nuclear reactors motivated validation work for available evaluation tools for nuclear fuel burnup calculations. In this study two deterministic codes with different transport solvers and one Monte Carlo method are investigated. The code system KAPROS/KARBUS uses the classical deterministic First Collision Probability method utilizing a cylinderized Wigner-Seitz cell. In the SCALES.0/TRITON/NEWT code the Extended Step Characteristic method is applied. In a first step the two deterministic codes are compared with experimental results from the KWO-Isotope Correlation Experiment up to 30 MWD/kg HM burnup, published in 1981. Two pin cell calculations are analyzed by comparison of calculated and experimental results for important heavy isotope vectors. The results are very satisfactory. Subsequently, further validation at higher burnup (< 80 MWD/kg HM) is provided by comparison of the two deterministic codes and the Monte Carlo based burnup code MONTEBURNS for PWR UO{sub 2} fuel assembly calculations. Possible reasons for differences in the results are analyzed and discussed. Especially the influence of cross section data and processing is presented. (authors)

  20. A comparative study on the wear behaviors of cladding candidates for accident-tolerant fuel

    SciTech Connect

    Lee, Young-Ho; Byun, Thak Sang

    2015-10-01

    Accident-tolerant fuels are expected to have considerably longer coping time to respond to the loss of active cooling under severe accidents and, at the same time, have comparable or improved fuel performance during normal operation. The wear resistance of accident tolerant fuels, therefore, needs to be examined to determine the applicability of these cladding candidates to the current operating PWRs because the most common failure of nuclear fuel claddings is still caused by grid-to-rod fretting during normal operations. In this study, reciprocating sliding wear tests on three kinds of cladding candidates for accident-tolerant fuels have been performed to investigate the tribological compatibilities of selfmated cladding candidates and to determine the direct applicability of conventional Zirconium-based alloys as supporting structural materials. The friction coefficients of the cladding candidates are strongly influenced by the test environments and coupled materials. The wear test results under water lubrication conditions indicate that the supporting structural materials for the cladding candidates of accident-tolerant fuels need to be replaced with the same cladding materials instead of using conventional Zirconium-based alloys.

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

    SciTech Connect

    Powers, Jeffrey J.

    2015-09-18

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

  2. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    DOE PAGES

    Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; ...

    2015-03-19

    FeCrAl is an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In our study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. Also, the total tritium inventory insidemore » the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.« less

  3. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    SciTech Connect

    Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

    2015-03-19

    FeCrAl is an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In our study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. Also, the total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

  4. Electron probe microanalysis of a METAPHIX UPuZr metallic alloy fuel irradiated to 7.0 at.% burn-up

    NASA Astrophysics Data System (ADS)

    Brémier, S.; Inagaki, K.; Capriotti, L.; Poeml, P.; Ogata, T.; Ohta, H.; Rondinella, V. V.

    2016-11-01

    The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15-20 μm and migration of cladding elements to the fuel.

  5. Reactor vessel cladding separate effects studies

    SciTech Connect

    Corwin, W.R.

    1985-01-01

    The existence of a layer of tough weld overlay cladding on the interior of a light-water reactor pressure vessel could mitigate damage caused during certain overcooling transients. The potential benefit of the cladding is that it could keep a short surface flaw, which would otherwise become long, from growing either by impeding crack initiation or by arresting a running crack. Two aspects critical to cladding behavior will be reported: irradiation effects on cladding toughness and the response of mechanically loaded, flawed structures in the presence of cladding. 15 refs., 24 figs., 6 tabs.

  6. Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

    NASA Astrophysics Data System (ADS)

    Yano, Y.; Tanno, T.; Oka, H.; Ohtsuka, S.; Inoue, T.; Kato, S.; Furukawa, T.; Uwaba, T.; Kaito, T.; Ukai, S.; Oono, N.; Kimura, A.; Hayashi, S.; Torimaru, T.

    2017-04-01

    Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900-1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

  7. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

    SciTech Connect

    George, Nathan Michael; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Maldonado, Ivan

    2014-09-29

    A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in the fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to

  8. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

    DOE PAGES

    George, Nathan Michael; Terrani, Kurt A.; Powers, Jeffrey J.; ...

    2014-09-29

    A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in themore » fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to the

  9. UV-curable low index hybrid glass as hard cladding for silica fibers

    NASA Astrophysics Data System (ADS)

    Wojcik, Anna B.; Wojcik, Jan

    2005-09-01

    UV-curable hybrid glass materials, when applied as thin coatings on optical fibers are already known to result in fibers enhanced mechanical strength as well as thermal and environmental stability [1]. These materials, when fully cured offer refractive index in the range 1.470-1.50 measured at 1300 nm. In search for low optical loss and lower refractive index claddings the hybrid glass composition was altered to result in UV curable two component formulation HG-LI (1+2) of refractive index 1. 380 The goal of this study was fabrication and valuation of the optical, mechanical and thermal properties of the silica fibers cladded with this novel hybrid glass material. The silica fibers were drawn and cladded by HG-LI-(1+2) and HG-LI-2. For control fibers, soft silicone resin and hard silicone resin were used as cladding for the same silica preform rod. The basic optical (spectral attenuation) and mechanical characteristics (tensile strength, n parameter) for hybrid glass cladded fibers were performed. Thermo Gravimetric Analysis (TGA) was performed as well. The results showed that hybrid glass cladding has the onset decomposition temperature exceeding 300 °C, whereas the NA values of the silica fiber cladded by HG-LI-(1+2) and HG-Li-2 were 0.31 and 0.33, respectively.

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

    SciTech Connect

    Jiang, Hao; Wang, Jy-An John

    2014-01-01

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

  11. Cladding burst behavior of Fe-based alloys under LOCA

    DOE PAGES

    Terrani, Kurt A.; Dryepondt, Sebastien N.; Pint, Bruce A.; ...

    2015-12-17

    Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. Themore » most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Furthermore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures.« less

  12. Cladding burst behavior of Fe-based alloys under LOCA

    SciTech Connect

    Terrani, Kurt A.; Dryepondt, Sebastien N.; Pint, Bruce A.; Massey, Caleb P.

    2015-12-17

    Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. The most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Furthermore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures.

  13. Laser cladding of bioactive glass coatings.

    PubMed

    Comesaña, R; Quintero, F; Lusquiños, F; Pascual, M J; Boutinguiza, M; Durán, A; Pou, J

    2010-03-01

    Laser cladding by powder injection has been used to produce bioactive glass coatings on titanium alloy (Ti6Al4V) substrates. Bioactive glass compositions alternative to 45S5 Bioglass were demonstrated to exhibit a gradual wetting angle-temperature evolution and therefore a more homogeneous deposition of the coating over the substrate was achieved. Among the different compositions studied, the S520 bioactive glass showed smoother wetting angle-temperature behavior and was successfully used as precursor material to produce bioactive coatings. Coatings processed using a Nd:YAG laser presented calcium silicate crystallization at the surface, with a uniform composition along the coating cross-section, and no significant dilution of the titanium alloy was observed. These coatings maintain similar bioactivity to that of the precursor material as demonstrated by immersion in simulated body fluid.

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

    SciTech Connect

    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.

  15. Microstructural examination of high temperature creep failure of Zircaloy-2 cladding in irradiated PHWR fuel pins

    NASA Astrophysics Data System (ADS)

    Mishra, Prerna; Sah, D. N.; Kumar, Sunil; Anantharaman, S.

    2012-10-01

    Cladding samples taken from the ballooned region of the irradiated Zircaloy-2 cladded PHWR fuel pins which failed during isothermal heating tests carried out at 800-900 °C were examined using optical and scanning electron microscopy. The examination of samples from the fuel pin tested at 900 °C showed an intergranular mode of failure in the cladding due to formation of cracks, cavities and zirconium hydride precipitates on the grain boundaries in the cladding material. A thin hard α-Zr(O) layer was observed on outer surface due to dissolution of the oxide layer formed during reactor operation. Grain boundary sliding was identified to be the main mode of creep deformation of Zircaloy-2 at 900 °C. Examination of the cladding tested at 800 °C showed absence of cracks or cavities in the deformed material and no localisation of hydrides was observed at the grain boundaries. The failure of the cladding occurred after necking followed by extensive wall thinning of the cladding tube.

  16. Model biases in high-burnup fast reactor simulations

    SciTech Connect

    Touran, N.; Cheatham, J.; Petroski, R.

    2012-07-01

    A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

  17. The Fork+ burnup measurement system: Design and first measurement campaign

    SciTech Connect

    Olson, C.E.; Bronowski, D.R.; McMurtry, W.; Ewing, R.; Jordan, R.; Rivard, D.

    1998-12-31

    Previous work with the original Fork detector showed that burnup as determined by reactor records could be accurately allocated to spent nuclear fuel assemblies. The original Fork detector, designed by Los Alamos National Laboratory, used an ion chamber to measure gross gamma count and a fission chamber to measure neutrons from an activation source, {sup 244}Cm. In its review of the draft Topical Report on Burnup Credit, the US Nuclear Regulatory Commission indicated it felt uncomfortable with a measurement system that depended on reactor records for calibration. The Fork+ system was developed at Sandia National Laboratories under the sponsorship of the Electric Power Research Institute with the aim of providing this independent measurement capability. The initial Fork+ prototype was used in a measurement campaign at the Maine Yankee reactor. The campaign confirmed the applicability of the sensor approach in the Fork+ system and the efficiency of the hand-portable Fork+ prototype in making fuel assembly measurements. It also indicated potential design modifications that will be necessary before the Fork+ can be used effectively on high-burnup spent fuel.

  18. Laser Cladding of Vanadium-Carbide Tool Steels for Die Repair

    NASA Astrophysics Data System (ADS)

    Leunda, J.; Soriano, C.; Sanz, C.; Navas, V. García

    A study of the laser cladding of powder metallurgical tool steels has been carried out. CPM 10 V and Vanadis 4 Extra tool steel powders have been deposited on Vanadis 4 Extra tool steel plates, for repairing purposes, using a Nd:YAG laser. The microstructure of the laser cladding samples was investigated using optical and scanning electron microscopes. The volumetric fraction of retained austenite was evaluated by X-ray diffraction and microhardness profiles were measured. Crack free 700 HV 0.3 cladding tracks were achieved with both materials and coatings show a microstructure of carbides embedded in a martensite plus retained austenite matrix.

  19. Summary of work on coatings and claddings for fossil energy applications

    SciTech Connect

    Swindeman, R.W.

    1993-05-01

    A summary of efforts to examine coatings and cladding materials for high-strength austenitic steels is provided. Chromized coatings on 17--14CuMo stainless steel and a modified type 316 (HT-UPS) stainless steel were investigated. Claddings included alloy 671, 690, and an iron-aluminide intermetallic alloy. Structural alloys that were clad included type 304 stainless steel, modified type 316 stainless steel, and modified alloy 800H. The capability of producing co-extruded tubing of the experimental alloys was demonstrated.

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

    SciTech Connect

    Rosenbaum, H.S.

    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.

  1. Extended burnup fuel - a beneficial environmental step for the nuclear fuel cycle

    SciTech Connect

    Tulenko, J.S. )

    1992-01-01

    The beneficial environmental effects generated by extending the burnup of light water reactors (LWR) fuel covers the entire fuel cycle from cradle (mining) to grave (fuel disposal). The purpose of this paper is to outline the beneficial efforts of extending burnup and show how a proposed financial incentive waste charge would encourage the development and utilization of extended burnup. Environmental consciousness demands that economic incentives be implemented to encourage utilities to extend burnup of their LWR fuel. This paper builds on earlier work and focuses and extends their results. The cornerstone of the positive environmental effects of extended burnup results from the increased plutonium energy generation as burnup is increased and residence time is increased.

  2. R&D Plan for RISMC Industry Application #1: ECCS/LOCA Cladding Acceptance Criteria

    SciTech Connect

    Szilard, Ronaldo Henriques; Zhang, Hongbin; Epiney, Aaron Simon; Tu, Lei

    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

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

    SciTech Connect

    Wang, Jy-An John; Jiang, Hao

    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.

  4. Weld overlay cladding with iron aluminides

    SciTech Connect

    Goodwin, G.M.

    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.

  5. Fully Coupled Modeling of Burnup-Dependent (U1- y , Pu y )O2- x Mixed Oxide Fast Reactor Fuel Performance

    NASA Astrophysics Data System (ADS)

    Liu, Rong; Zhou, Wenzhong; Zhou, Wei

    2016-03-01

    During the fast reactor nuclear fuel fission reaction, fission gases accumulate and form pores with the increase of fuel burnup, which decreases the fuel thermal conductivity, leading to overheating of the fuel element. The diffusion of plutonium and oxygen with high temperature gradient is also one of the important fuel performance concerns as it will affect the fuel material properties, power distribution, and overall performance of the fuel pin. In order to investigate these important issues, the (U1- y Pu y )O2- x fuel pellet is studied by fully coupling thermal transport, deformation, oxygen diffusion, fission gas release and swelling, and plutonium redistribution to evaluate the effects on each other with burnup-dependent models, accounting for the evolution of fuel porosity. The approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with fast reactor mixed oxide fuel performance analysis. The modeling results showed a consistent fuel performance comparable with the previous results. Burnup degrades the fuel thermal conductivity, resulting in a significant fuel temperature increase. The fission gas release increased rapidly first and then steadily with the burnup increase. The fuel porosity increased dramatically at the beginning of the burnup and then kept constant as the fission gas released to the fuel free volume, causing the fuel temperature to increase. Another important finding is that the deviation from stoichiometry of oxygen affects greatly not only the fuel properties, for example, thermal conductivity, but also the fuel performance, for example, temperature distribution, porosity evolution, grain size growth, fission gas release, deformation, and plutonium redistribution. Special attention needs to be paid to the deviation from stoichiometry of oxygen in fuel fabrication. Plutonium content will also affect the fuel material properties and performance

  6. Protective claddings for high strength chromium alloys

    NASA Technical Reports Server (NTRS)

    Collins, J. F.

    1971-01-01

    The application of a Cr-Y-Hf-Th alloy as a protective cladding for a high strength chromium alloy was investigated for its effectiveness in inhibiting nitrogen embrittlement of a core alloy. Cladding was accomplished by a combination of hot gas pressure bonding and roll cladding techniques. Based on bend DBTT, the cladding alloy was effective in inhibiting nitrogen embrittlement of the chromium core alloy for up to 720 ks (200hours) in air at 1422 K (2100 F). A significant increase in the bend DBTT occurred with longer time exposures at 1422 K or short time exposures at 1589 K (2400 F).

  7. Advanced Fuels Campaign Cladding & Coatings Meeting Summary

    SciTech Connect

    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.

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

    SciTech Connect

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

    2014-04-01

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

  9. Feasibility study of fuel cladding performance for application in ultra-long cycle fast reactor

    NASA Astrophysics Data System (ADS)

    Jung, Ju Ang; Kim, Seung Hyun; Shin, Sang Hun; Bang, In Cheol; Kim, Ji Hyun

    2013-09-01

    As a part of the research and development activities for long-life core sodium-cooled fast reactors, the cladding performance of the ultra-long cycle fast reactor (UCFR) is evaluated with two design power levels (1000 MWe and 100 MWe) and cladding peak temperatures (873 K and 923 K). The key design concept of the UCFR is that it is non-refueling during its 30-60 years of operation. This concept may require a maximum peak cladding temperature of 923 K and a cladding radiation damage of over 200 dpa (displacements per atom). Therefore, for the design of the UCFR, deformation due to thermal creep, irradiation creep, and swelling must be taken into consideration through quantitative evaluations. As candidate cladding materials for use in UCFRs, ferritic-martensitic (FM) steels, oxide dispersion strengthened (ODS) steels, and SiC-based composite materials are studied using deformation behavior modeling for a feasibility evaluation. The results of this study indicate that SiC is a potential UCFR cladding material, with the exception of irradiation creep due to high neutron fluence stemming from its long operating time of about 30-60 years.

  10. Propagation of Uncertainty in System Parameters of a LWR Model by Sampling MCNPX Calculations - Burnup Analysis

    NASA Astrophysics Data System (ADS)

    Campolina, Daniel de A. M.; Lima, Claubia P. B.; Veloso, Maria Auxiliadora F.

    2014-06-01

    For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95th percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input.

  11. Optical Fiber Geometry: Accurate Measurement of Cladding Diameter

    PubMed Central

    Young, Matt; Hale, Paul D.; Mechels, Steven E.

    1993-01-01

    We have developed three instruments for accurate measurement of optieal fiber cladding diameter: a contact micrometer, a scanning confocal microscope, and a white-light interference microscope. Each instrument has an estimated uncertainty (3 standard deviations) of 50 nm or less, but the confocal microscope may display a 20 nm systematic error as well. The micrometer is used to generate Standard Reference Materials that are commercially available. PMID:28053467

  12. Determination of IRT-2M fuel burnup by gamma spectrometry.

    PubMed

    Koleška, Michal; Viererbl, Ladislav; Marek, Milan; Ernest, Jaroslav; Šunka, Michal; Vinš, Miroslav

    2016-01-01

    A spectrometric system was developed for evaluating spent fuel in the LVR-15 research reactor, which employs highly enriched (36%) IRT-2M-type fuel. Such system allows the measurement of detailed fission product profiles. Within these measurements, nuclides such as (137)Cs, (134)Cs, (144)Ce, (106)Ru and (154)Eu may be detected in fuel assemblies with different cooling times varying between 1.67 and 7.53 years. Burnup calculations using the MCNPX Monte Carlo code data showed good agreement with measurements, though some discrepancies were observed in certain regions. These discrepancies are attributed to the evaluation of irradiation history, reactor regulation pattern and buildup schemes.

  13. New results from the NSRR experiments with high burnup fuel

    SciTech Connect

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide

    1996-03-01

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed.

  14. Reliability of hard plastic clad silica fibers

    NASA Astrophysics Data System (ADS)

    Skutnik, Bolesh J.; Spaniol, Stefan

    2006-04-01

    New formulations of cladding materials have become available in recent times for Hard Plastic Clad Silica (HPCS) fibers, Initial data showed gains in some properties, particularly dynamic strength, especially for high numerical aperture (NA) fibers. A systematic study has been undertaken to determine the full strength and fatigue behavior of these HPCS fibers and to make comparisons to earlier HPCS fibers. Preliminary results, now confirmed, has shown improved median dynamic strength and higher Weibull slope. Full results are presented below including fatigue behavior and optical properties. These fibers have many applications and benefits in the high power delivery and medical laser uses as highlighted below. High power diode laser systems with their laser diode bars and arrays not only require special fibers to couple directly to the diode emitters, but also require special fibers to couple from the laser to application sites. These latter power delivery fibers are much larger than the internal fibers but still must be flexible, and have not only good strength but also good fatigue behavior. This particularly important industrial systems using robotic arms to apply the high power laser energy at a treatment site. The optical properties of HPCS fibers are well suited for the needs of the delivery of high power from diode laser bars and arrays to an application site. Benefits of strong median dynamic strengths and tighter flaw distributions in such cases will be discussed. Many medical applications, especially endoscopic ones, can benefit from the use of highly flexible, high NA, cost effective, HPCS optical fibers. Benefits of high strength and good fatigue behavior for such fibers in endoscopic procedures, including laser surgery, are discussed briefly including implications for mechanical reliability in medical and industrial settings.

  15. Microbial Biofilm Growth on Irradiated, Spent Nuclear Fuel Cladding

    SciTech Connect

    S.M. Frank

    2009-02-01

    A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 × 103 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments.

  16. Cladding Alloys for Fluoride Salt Compatibility

    SciTech Connect

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

    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.

  17. Consolidation of cladding hulls from the electrometallurgical treatment of spent fuel.

    SciTech Connect

    Keiser, D. D., Jr.

    1998-04-10

    To consolidate metallic waste that is residual from Argonne National Laboratory's electrometallurgical treatment of spent nuclear fuel, waste ingots are currently being cast using an induction furnace located in a hot cell. These ingots, which have been developed to serve as final waste forms destined for repository disposal, are stainless steel (SS)-Zr alloys (the Zr is very near 15 wt.%). The charge for the alloys consists of stainless steel cladding hulls, Zr from the fuel being treated, noble metal fission products, and minor amounts of actinides that are present with the cladding hulls. The actual in-dated cladding hulls have been characterized before they were melted into ingots, and the final as-cast ingots have been characterized to determine the degree of consolidation of the charge material. It has been found that ingots can be effectively cast from irradiated cladding hulls residual from the electrometallurgical treatment process by employing an induction furnace located in a hot cell.

  18. Compatibility study between U-UO2 cermet fuel and T91 cladding

    NASA Astrophysics Data System (ADS)

    Mishra, Sudhir; Kaity, Santu; Khan, K. B.; Sengupta, Pranesh; Dey, G. K.

    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-UO2 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 U6(Fe,Cr) and U(Fe,Cr)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.

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

    SciTech Connect

    Sweet, Ryan; George, Nathan M.; Terrani, Kurt A.; Wirth, Brian

    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

  20. Corrosion inhibition of steam generator tubesheet by Alloy 690 cladding in secondary side environments

    NASA Astrophysics Data System (ADS)

    Hur, Do Haeng; Choi, Myung Sik; Lee, Deok Hyun; Han, Jung Ho; Shim, Hee Sang

    2013-11-01

    Denting is a phenomenon that a steam generator tube is distorted by a volume expansion of corrosion products of the tube support and tubesheet materials adjacent to the tube. Although denting has been mitigated by a modification of the design and material of the tube support structures, it has been an inevitable concern in the crevice region of the top of tubesheet. This paper provides a new technology to prevent denting by cladding the secondary surface of the tubesheet with a corrosion resistant material. In this study, Alloy 690 material was cladded onto the surface of an SA508 tubesheet to a thickness of about 9 mm. The corrosion rates of the original SA508 tubesheet and the Alloy 690 clad material were measured in acidic and alkaline simulated environments. Using Alloy 690 cladding, the corrosion rate of the tubesheet within a magnetite sludge pile decreased by a factor of 680 in 0.1 M NiCl2 solution at 300 °C, and by a factor of 58 in 2 M NaOH solution at 315 °C. This means that denting can drastically be prevented by cladding the secondary tubesheet surface with corrosion resistant materials.

  1. Development of Monteburns: A Code That Links MCNP and ORIGEN2 in an Automated Fashion for Burnup Calculations

    SciTech Connect

    Holly R. Trellue

    1998-12-01

    Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.

  2. Attainable Burnup in a LIFE Engine Loaded with Depleted Uranium

    SciTech Connect

    Fratoni, M; Kramer, K J; Latkowski, J F

    2009-11-30

    The Laser Inertial Fusion-based Energy (LIFE) system uses a laser-based fusion source for electricity production. The (D,T) reaction, beside a pure fusion system, allows the option to drive a sub-critical fission blanket in order to increase the total energy gain. In a typical fusion-fission LIFE engine the fission blanket is a spherical shell around the fusion source, preceded by a beryllium shell for neutron multiplications by means of (n,2n) reactions. The fuel is in the form of TRISO particles dispersed in carbon pebbles, cooled by flibe. The optimal design features 80 cm thick blanket, 16 cm multiplier, and 20% TRISO packing factor. A blanket loaded with depleted uranium and depleted in a single batch with continuous mixing can achieve burnup as high as {approx}85% FIMA while generating 2,000 MW of total thermal power and producing enough tritium to be used for fusion. A multi-segment blanket with a central promotion shuffling scheme enhances burnup to {approx}90% FIMA, whereas a blanket that is operated with continuous refueling achieves only 82% FIMA under the same constraints of thermal power and tritium self-sufficiency. Both, multi-segment and continuous refueling eliminate the need for a fissile breeding phase.

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

    NASA Astrophysics Data System (ADS)

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

    2016-05-01

    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.

  4. Secondary ion mass spectrometry of irradiated nuclear fuel and cladding

    NASA Astrophysics Data System (ADS)

    Portier, S.; Brémier, S.; Walker, C. T.

    2007-06-01

    The principles and operating modes of secondary ion mass spectrometry (SIMS) are first described after which the different methods of quantification are summarised. Some current applications of SIMS in nuclear fuel and cladding research are then reviewed after briefly considering the modifications that are needed to allow a SIMS instrument to be used for the analysis of highly radioactive materials. Amongst the applications reported are the investigation of the behaviour of fission gas xenon and the volatile fission products tellurium, iodine and caesium in UO2 nuclear fuel, measurement of the radial distribution of Pu isotopes in mixed oxide (MOX) fuel and of the radial distribution of Gd isotopes in (U,Gd)O2 fuel, and determination of the distribution of Li and B in the external oxide layer on Zircaloy cladding. It is evident from the large amount of new information gained that SIMS is a powerful complementary technique to electron probe microanalysis (EPMA) in these fields of study.

  5. Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance

    SciTech Connect

    Wagner, John C; Parks, Cecil V; Mueller, Don; Gauld, Ian C

    2010-01-01

    Taking credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transport, disposal, and reprocessing of spent nuclear fuel (SNF) while maintaining sufficient subcritical margin to establish an adequate safety basis. Consequently, there continues to be considerable interest in the United States (U.S.), as well as internationally, in the increased use of burnup credit in SNF operations, particularly related to storage, transport, and disposal of commercial SNF. This interest has motivated numerous technical studies related to the application of burnup credit, both domestically and internationally, as well as the design of SNF storage, transport and disposal systems that rely on burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission (NRC) initiated a burnup credit research program in 1999, with support from the Oak Ridge National Laboratory (ORNL), to develop regulatory guidance and the supporting technical bases for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Although this NRC research program has not been continuous since its inception, considerable progress has been achieved in many key areas in terms of increased understanding of relevant phenomena and issues, availability of relevant information and data, and subsequently updated regulatory guidance for expanded use of burnup credit. This paper reviews technical studies performed by ORNL for the U.S. NRC burnup credit research program. Examples of topics include reactivity effects associated with reactor operating characteristics, fuel assembly characteristics, burnable absorbers, control rods, spatial burnup distributions, cooling time, and assembly misloading; methods and data for validation of isotopic composition predictions; methods and data for validation of criticality calculations; and

  6. Metallography of pitted aluminum-clad, depleted uranium fuel

    SciTech Connect

    Nelson, D.Z.; Howell, J.P.

    1994-12-01

    The storage of aluminum-clad fuel and target materials in the L-Disassembly Basin at the Savannah River Site for more than 5 years has resulted in extensive pitting corrosion of these materials. In many cases the pitting corrosion of the aluminum clad has penetrated in the uranium metal core, resulting in the release of plutonium, uranium, cesium-137, and other fission product activity to the basin water. In an effort to characterize the extent of corrosion of the Mark 31A target slugs, two unirradiated slug assemblies were removed from basin storage and sent to the Savannah River Technology Center for evaluation. This paper presents the results of the metallography and photographic documentation of this evaluation. The metallography confirmed that pitting depths varied, with the deepest pit found to be about 0.12 inches (3.05 nun). Less than 2% of the aluminum cladding was found to be breached resulting in less than 5% of the uranium surface area being affected by corrosion. The overall integrity of the target slug remained intact.

  7. Behavior of Zircaloy Cladding in the Presence of Gallium

    SciTech Connect

    DiStefano, J.R.; King, J.F.; Manneschmidt, E.T.; Strizak, J.P.; Wilson, D.F.

    1998-09-28

    The U.S. Department of Energy has established a dual-track approach to the disposition of plutonium arising from the dismantling of nuclear weapons. Both immobilization and reactor-based mixed-oxide (MOX) fuel technologies are being evaluated. The reactor-based MOX fuel option requires assessment of the potential impact of concentrations of gallium (on the order of 1 to 10 ppm), not present in conventional MOX fuel, on cladding material performance. An experimental program was designed to evaluate the performance of prototypic Zircaloy cladding materials against (1) liquid gallium, and (2) various concentrations of G~03. Three types of tests were performed: (1) corrosion, (2) liquid metal embrittlement, and (3) corrosion-mechanical. These tests were to determine corrosion mechanisms, thresholds for temperature and concentration of gallium that delineate behavioral regimes, and changes in the mechanical properties of Zircaloy. Results have generally been favorable for the use of weapons-grade (WG) MOX fhel. The Zircaloy cladding does react with gallium to form intermetallic compounds at >3000 C; however, this reaction is limited by the mass of gallium and is therefore not expected to be significant with a low level (parts per million) of gallium in the MOX fuel. Furthermore, no evidence for grain boundary penetration by gallium or liquid metal embrittlement was observed.

  8. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    NASA Astrophysics Data System (ADS)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  9. Research and development status of laser cladding on magnesium alloys: A review

    NASA Astrophysics Data System (ADS)

    Liu, Jianli; Yu, Huijun; Chen, Chuanzhong; Weng, Fei; Dai, Jingjie

    2017-06-01

    Magnesium alloys are one of the most promising lightweight structural materials. However, the poor corrosion and wear resistance restrain their further application. As a kind of surface modification technique, laser cladding treatment is superior to others owing to its unique characteristics such as high efficiency and the metallurgical bonding between the coatings and substrates. In this paper, the laser cladding process and the effects of processing parameters, including laser power, scanning velocity, beam focal position, feeding ways of the material etc., are discussed in detail. The material systems preplaced on magnesium alloys are summarized. Except for the traditional metallic materials, novel ternary alloys, amorphous alloys and high entropy alloys (HEAs) are widely used and apparent advantages are exhibited. In terms of the problems existing in the laser cladding process of magnesium alloys, some potential solutions and the development tendency are reviewed.

  10. Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

    SciTech Connect

    William J. Carmack

    2012-05-01

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

  11. Cascaded-cladding-pumped cascaded Raman fiber amplifier.

    PubMed

    Jiang, Huawei; Zhang, Lei; Feng, Yan

    2015-06-01

    The conversion efficiency of double-clad Raman fiber laser is limited by the cladding-to-core area ratio. To get high conversion efficiency, the inner-cladding-to-core area ratio has to be less than about 8, which limits the brightness enhancement. To overcome the problem, a cascaded-cladding-pumped cascaded Raman fiber laser with multiple-clad fiber as the Raman gain medium is proposed. A theoretical model of Raman fiber amplifier with multiple-clad fiber is developed, and numerical simulation proves that the proposed scheme can improve the conversion efficiency and brightness enhancement of cladding pumped Raman fiber laser.

  12. Experimental Study on Surface Characteristics of Laser Cladding Layer Regulated by High-Frequency Microforging

    NASA Astrophysics Data System (ADS)

    Fan, Xiang Fang; Zhou, Ju; Qiu, Chang Jun; He, Bin; Ye, Jiang; Yuan, Bo; Pi, Zhengqing

    2011-03-01

    High-frequency microforging technology is used to produce micrometer-scale plastic deformation on the surface of material out of the vibration impact of a forging punch, and the cumulative effect of its various frequencies on micrometer-scale plastic deformation can cause changes of surface microstructure and mechanical properties. This study used (1) a self-made machine to treat NiCrBSi alloy, (2) a mechanical comparator and optical microscopy (OM) to study the geometric characteristics of plastic deformation, (3) OM and scanning electric microscopy (SEM) to observe influence on surface microstructure and cracking behavior of the laser cladding layer under microforging, (4) x-ray diffractometer (XRD) to measure the surface residual stress of laser cladding layer before and after forging, and (5) microhardness tester and wearing experimental machine to study changes of microhardness, friction coefficient, and wear characteristics of laser cladding layer after microforging. The results have shown that high-frequency microforging could produce plastic deformation about 150 μm deep on the surface of NiCrBSi alloy clad by laser. Regular dendrite and eutectic crystallization microstructure, which is a peculiar characteristic of the laser cladding layer, was broken into pieces and formed residual compression residual stress on the surface. Resistance to cracking of laser cladding layer improved greatly, microhardness and wearability increased, and the friction coefficient did not under go a noticeable change.

  13. The burnup dependence of light water reactor spent fuel oxidation

    SciTech Connect

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies

  14. Development of a used fuel cladding damage model incorporating circumferential and radial hydride responses

    NASA Astrophysics Data System (ADS)

    Chen, Qiushi; Ostien, Jakob T.; Hansen, Glen

    2014-04-01

    At the completion of the fuel drying process, used fuel Zry4 cladding typically exhibits a significant population of δ-hydride inclusions. These inclusions are in the form of small platelets that are generally oriented both circumferentially and radially within the cladding material. There is concern that radially-oriented hydride inclusions may weaken the cladding material and lead to issues during used fuel storage and transportation processes. A high fidelity model of the mechanical behavior of hydrides has utility in both designing fuel cladding to be more resistant to this hydride-induced weakening and also in suggesting modifications to drying, storage, and transport operations to reduce the impact of hydride formation and/or the avoidance of loading scenarios that could overly stress the radial inclusions. We develop a mechanical model for the Zry4-hydride system that, given a particular morphology of hydride inclusions, allows the calculation of the response of the hydrided cladding under various loading scenarios. The model treats the Zry4 matrix material as J2 elastoplastic, and treats the hydrides as platelets oriented in predefined directions (e.g., circumferentially and radially). The model is hosted by the Albany analysis framework, where a finite element approximation of the weak form of the cladding boundary value problem is solved using a preconditioned Newton-Krylov approach. Instead of forming the required system Jacobian operator directly or approximating its action with a differencing operation, Albany leverages the Trilinos Sacado package to form the Jacobian via automatic differentiation. We present results that describe the performance of the model in comparison with as-fabricated Zry4 as well as HB Robinson fuel cladding. Further, we also present performance results that demonstrate the efficacy of the overall solution method employed to host the model.

  15. Burnup calculations for KIPT accelerator driven subcritical facility using Monte Carlo computer codes-MCB and MCNPX.

    SciTech Connect

    Gohar, Y.; Zhong, Z.; Talamo, A.; Nuclear Engineering Division

    2009-06-09

    Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron accelerator. The neutron source of the subcritical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and electron energy in the range of 100 to 200 MeV. The main functions of the subcritical assembly are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron physics experiments and material structure analyses are planned using this facility. With the 100 KW electron beam power, the total thermal power of the facility is {approx}375 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products reduce continuously the reactivity during the operation, which reduces the neutron flux level and consequently the facility performance. To preserve the neutron flux level during the operation, fuel assemblies should be added after long operating periods to compensate for the lost reactivity. This process requires accurate prediction of the fuel burnup, the decay behavior of the fission produces, and the introduced reactivity from adding fresh fuel assemblies. The recent developments of the Monte Carlo computer codes, the high speed capability of the computer processors, and the parallel computation techniques made it possible to perform three-dimensional detailed burnup simulations. A full detailed three-dimensional geometrical model is used for the burnup simulations with continuous energy nuclear data libraries for the transport calculations and 63-multigroup or one group cross sections libraries for the depletion calculations. Monte Carlo Computer code MCNPX and MCB are utilized for this study. MCNPX transports the

  16. Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

    SciTech Connect

    Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi

    2014-09-30

    Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided.

  17. Mechanical Fatigue Testing of High Burnup Fuel for Transportation Applications

    SciTech Connect

    Wang, Jy-An John; Wang, Hong

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

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

    NASA Astrophysics Data System (ADS)

    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.

  19. Fuel Performance Calculations for FeCrAl Cladding in BWRs

    SciTech Connect

    George, Nathan; Sweet, Ryan; Maldonado, G. Ivan; Wirth, Brian D.; Powers, Jeffrey J.; Worrall, Andrew

    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.

  20. Glass-clad semiconductor core optical fibers

    NASA Astrophysics Data System (ADS)

    Morris, Stephanie Lynn

    Glass-clad optical fibers comprising a crystalline semiconductor core have garnered considerable recent attention for their potential utility as novel waveguides for applications in nonlinear optics, sensing, power delivery, and biomedicine. As research into these fibers has progressed, it has become evident that excessive losses are limiting performance and so greater understanding of the underlying materials science, coupled with advances in fiber processing, is needed. More specifically, the semiconductor core fibers possess three performance-limiting characteristics that need to be addressed: (a) thermal expansion mismatches between crystalline core and glass cladding that lead to cracks, (b) the precipitation of oxide species in the core upon fiber cooling, which results from partial dissolution of the cladding glass by the core melt, and (c) polycrystallinity; all of which lead to scattering and increased transmission losses. This dissertation systematically studies each of these effects and develops both a fundamental scientific understanding of and practical engineering methods for reducing their impact. With respect to the thermal expansion mismatch and, in part, the dissolution of oxides, for the first time to our knowledge, oxide and non-oxide glass compositions are developed for a series of semiconductor cores based on two main design criteria: (1) matching the thermal expansion coefficient between semiconductor core and glass cladding to minimize cracking and (2) matching the viscosity-temperature dependences, such that the cladding glass draws into fiber at a temperature slightly above the melting point of the semiconductor in order to minimize dissolution and improve the fiber draw process. The x[Na 2O:Al2O3] + (100 - 2x)SiO2 glass compositional family was selected due to the ability to tailor the glass properties to match the aforementioned targets through slight variations in composition and adjusting the ratios of bridging and non-bridging oxygen

  1. DISSOLUTION OF ZIRCALOY 2 CLAD UO2 COMMERCIAL REACTOR FUEL

    SciTech Connect

    Kessinger, G.; Thompson, M.

    2009-08-07

    The primary goal of this investigation was to evaluate the effectiveness of the chop-leach process, with nitric acid solvent, to produce a nominally 300 g/L [U] and 1 M [H{sup +}] product solution. The results of this study show that this processing technique is appropriate for applications in which a low free acid and moderately high U content are desired. The 7.75 L of product solution, which was over 450 g/L in U, was successfully diluted to produce about 13 L of solvent extraction feed that was 302 g/L in U with a [H{sup +}] in the range 0.8-1.2 M. A secondary goal was to test the effectiveness of this treatment for the removal of actinides from Zircaloy cladding to produce a low-level radioactive waste (LLW) cladding product. Analysis of the cladding shows that actinides are present in the cladding at a concentration of about 5000 {eta}Ci/g, which is about 50 times greater than the acceptable transuranium element limit in low level radioactive waste. It appears that the concentration of nitric acid used for this dissolution study (initial concentration 4 M, with 10 M added as the dissolution proceeded) was inadequate to completely digest the UO{sub 2} present in the spent fuel. The mass of insoluble material collected from the initial treatments with nitric acid, 340 g, was much higher than expected, and analysis of this insoluble residue showed that it contained at least 200 g U.

  2. Interactions of Zircaloy cladding with gallium: 1998 midyear status

    SciTech Connect

    Wilson, D.F.; DiStefano, J.R.; Strizak, J.P.; King, J.F.; Manneschmidt, E.T.

    1998-06-01

    A program has been implemented to evaluate the effect of gallium in mixed-oxide (MOX) fuel derived from weapons-grade (WG) plutonium on Zircaloy cladding performance. The objective is to demonstrate that low levels of gallium will not compromise the performance of the MOX fuel system in a light-water reactor. The graded, four-phase experimental program was designed to evaluate the performance of prototypic Zircaloy cladding materials against (1) liquid gallium (Phase 1), (2) various concentrations of Ga{sub 2}O{sub 3} (Phase 2), (3) centrally heated surrogate fuel pellets with expected levels of gallium (Phase 3), and (4) centrally heated prototypic MOX fuel pellets (Phase 4). This status report describes the results of a series of tests for Phases 1 and 2. Three types of tests are being performed: (1) corrosion, (2) liquid metal embrittlement, and (3) corrosion-mechanical. These tests will determine corrosion mechanisms, thresholds for temperature and concentration of gallium that may delineate behavioral regimes, and changes in the mechanical properties of Zircaloy. Initial results have generally been favorable for the use of WG-MOX fuel. The MOX fuel cladding, Zircaloy, does react with gallium to form intermetallic compounds at {ge}300 C; however, this reaction is limited by the mass of gallium and is therefore not expected to be significant with a low level (parts per million) of gallium in the MOX fuel. Although continued migration of gallium into the initially formed intermetallic compound can result in large stresses that may lead to distortion, this was shown to be extremely unlikely because of the low mass of gallium or gallium oxide present and expected clad temperatures below 400 C. Furthermore, no evidence for grain boundary penetration by gallium has been observed.

  3. Interactions of zircaloy cladding with gallium -- 1997 status

    SciTech Connect

    Wilson, D.F.; DiStefano, J.R.; King, J.F.; Manneschmidt, E.T.; Strizak, J.P.

    1997-11-01

    A four phase program has been implemented to evaluate the effect of gallium in mixed oxide (MOX) fuel derived from weapons grade (WG) plutonium on Zircaloy cladding performance. The objective is to demonstrate that low levels of gallium will not compromise the performance of the MOX fuel system in LWR. This graded, four phase experimental program will evaluate the performance of prototypic Zircaloy cladding materials against: (1) liquid gallium (Phase 1), (2) various concentrations of Ga{sub 2}O{sub 3} (Phase 2), (3) centrally heated surrogate fuel pellets with expected levels of gallium (Phase 3), and (4) centrally heated prototypic MOX fuel pellets (Phase 4). This status report describes the results of an initial series of tests for phases 1 and 2. Three types of tests are being performed: (1) corrosion, (2) liquid metal embrittlement (LME), and (3) corrosion mechanical. These tests are designed to determine the corrosion mechanisms, thresholds for temperature and concentration of gallium that may delineate behavioral regimes, and changes in mechanical properties of Zircaloy. Initial results have generally been favorable for the use of WG-MOX fuel. The MOX fuel cladding, Zircaloy, does react with gallium to form intermetallic compounds at {ge} 300 C; however, this reaction is limited by the mass of gallium and is therefore not expected to be significant with a low level (in parts per million) of gallium in the MOX fuel. While continued migration of gallium into the initially formed intermetallic compound results in large stresses that can lead to distortion, this is also highly unlikely because of the low mass of gallium or gallium oxide present and expected clad temperatures below 400 C. Furthermore, no evidence for grain boundary penetration by gallium has been observed.

  4. The effect of stainless steel overlay cladding on corrosion fatigue crack propagation in pressure vessel steel in PWR primary coolant

    SciTech Connect

    Bramwell, I.L.; Tice, D.R.; Worswick, D.; Heys, G.B.

    1995-12-31

    The growth of sub-critical cracks in pressure boundary materials in light water reactors is assessed using codified procedures, but the presence of the overlay-welded stainless steel cladding on the pressure vessel is not normally taken into consideration because of the difficulty in demonstrating clad integrity for the lifetime of the plant. In order to investigate any possible effect of the cladding layer on crack propagation, tests have been performed using two types of specimen. The first was sputter ion plated with a thin layer of austenitic stainless steel to simulate the electrochemical and oxide effects due to the cladding, whilst the second used an overlay clad specimen to investigate the behavior of a crack propagating from the austenitic into the ferritic material. Testing was carried out under cyclic loading conditions in well controlled simulated PWR primary water. At 288 C, the presence of stainless steel in contact with the low alloy steel did not enhance crack propagation in PWR primary coolant compared to unclad or unplated specimens. There was limited evidence that at 288 C under certain loading conditions, in both air and PWR water, there may be an effect of the cladding which reduces crack growth rates, at least for a short distance of crack propagation into the low alloy steel. Crack growth rates in the ferritic steel at 130 C were higher for both the plated and clad specimens than found in previous tests under similar conditions on the unclad material. However, the crack growth rates were bounded by current ASME 11 Appendix A recommendations for defects exposed to water and at low R ratio. There was no evidence of environmental enhancement of crack propagation in the stainless steel in clad specimens. The results indicate that the current approach of ignoring the cladding for assessment purposes is conservative at plant operating temperature.

  5. Double-clad nuclear fuel safety rod

    DOEpatents

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

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

  6. Interactions of Zircaloy Cladding with Gallium: Final Report

    SciTech Connect

    D.F. Wilson; E.T. Manneschmidt; J.F. King; J.P. Strizak; J.R. DiStefano

    1998-09-01

    The U.S. Department of Energy has established a dual-track approach to the disposition of plutonium arising from the dismantling of nuclear weapons. Both immobilization and reactor-based mixed-oxide (MOX) fuel technologies are being evaluated. The reactor-based MOX fuel option requires assessment of the potential impact of concentrations of gallium (on the order of 1 to 10 ppm), not present in conventional MOX fhel, on cladding material performance. Three previous repmts"3 identified several compatibility issues relating to the presence of gallium in MOX fuel and its possible reaction with fiel cladding. Gallium initially present in weapons-grade (WG) plutonium is largely removed during processing to produce MOX fhel. After blending the plutonium with uranium, only 1 to 10 ppm gallium is expected in the sintered MOX fuel. Gallium present as gallium oxide (G~OJ could be evolved as the suboxide (G~O). Migration of the evolved G~O and diffusion of gallium in the MOX matrix along thermal gradients could lead to locally higher concentrations of G~03. Thus, while an extremely low concentration of gallium in MOX fiel almost ensures a lack of significant interaction of gallium whh Zircaloy fhel cladding, there remains a small probability that corrosion effects will not be negligible. General corrosion in the form of surface alloying resulting from formation of intermetallic compounds between Zircaloy and gallium should be ma& limited and, therefore, superficial because of the expected low ratio of gallium to the surface area or volume of the Zircaloy cladding. Although the expected concentration of gallium is low and there is very limited volubility of gallium in zirconium, especially at temperatures below 700 "C,4 grain boundary penetration and liquid metal embrittlement (LME) are forms of localized corrosion that were also considered. One fuel system darnage mechanism, pellet clad interaction, has led to some failure of the Zircaloy cladding in light-water reactors (LWRS

  7. Clad Degradation- Summary and Abstraction for LA

    SciTech Connect

    D. Stahl

    2004-10-01

    The purpose of this model report is to develop the summary cladding degradation abstraction that will be used in the Total System Performance Assessment for the License Application (TSPA-LA). Most civilian commercial nuclear fuel is encased in Zircaloy cladding. The model addressed in this report is intended to describe the postulated condition of commercial Zircaloy-clad fuel as a function of postclosure time after it is placed in the repository. Earlier total system performance assessments analyzed the waste form as exposed UO{sub 2}, which was available for degradation at the intrinsic dissolution rate. Water in the waste package quickly became saturated with many of the radionuclides, limiting their release rate. In the total system performance assessments for the Viability Assessment and the Site Recommendation, cladding was analyzed as part of the waste form, limiting the amount of fuel available at any time for degradation. The current model is divided into two stages. The first considers predisposal rod failures (most of which occur during reactor operation and associated activities) and postdisposal mechanical failure (from static loading of rocks) as mechanisms for perforating the cladding. Other fuel failure mechanisms including those caused by handling or transportation have been screened out (excluded) or are treated elsewhere. All stainless-steel-clad fuel, which makes up a small percentage of the overall amount of fuel to be stored, is modeled as failed upon placement in the waste packages. The second stage of the degradation model is the splitting of the cladding from the reaction of water or moist air and UO{sub 2}. The splitting has been observed to be rapid in comparison to the total system performance assessment time steps and is modeled to be instantaneous. After the cladding splits, the rind buildup inside the cladding widens the split, increasing the diffusion area from the fuel rind to the waste package interior. This model report summarizes

  8. In Situ Production of Hard Metal Matrix Composite Coating on Engineered Surfaces Using Laser Cladding Technique

    NASA Astrophysics Data System (ADS)

    Raza, Mohammad Shahid; Hussain, Manowar; Kumar, Vikash; Das, Alok Kumar

    2016-11-01

    The growing need for high wear-resistant surface with enhanced physical properties has led to extensive researches in the field of surface engineering. Laser cladding emerged to be a promising method to achieve these objectives in a cost-effective way. The present paper studies the viability of cladding of tungsten disulfide (WS2) powder by using 400 W continuous-wave fiber laser. WS2 was used as a coating material, which was decomposed at higher temperature and underwent several chemical reactions. By this process, in situ formation of metal matrix composites and hard face coating on the substrate surface were attained. The characterization of laser cladded surface was done to study its morphological, microstructural, mechanical and tribological properties. It was observed that cladding of WS2 powder on 304 SS resulted in the formation of Cr-W-C-Fe metal matrix composite having improved mechanical and tribological properties. The value of microhardness of the coated surface was found to increase three to four times in comparison with the parent material surface. Wear test results indicated a decrease in wear by 1/9th (maximum) as compared to the parent 304 SS surface. The volume fractions of tungsten particles on the cladded surface were also investigated through EDS analysis.

  9. Friction and wear behaviors of compacted graphite iron with different biomimetic units fabricated by laser cladding

    NASA Astrophysics Data System (ADS)

    Sun, Na; Shan, Hongyu; Zhou, Hong; Chen, Darong; Li, Xiaoyan; Xia, Wen; Ren, Luquan

    2012-07-01

    Mimicking the biological characters on the cuticles of pangolin scales, biomimetic units were fabricated on the surfaces of compacted graphite cast iron (CGI) with different unit materials using laser cladding process. The influences of various unit materials including TiC, WC, B4C and Al2O3 powders on the friction and wear behaviors of CGI were investigated. The wear resistance mechanism of biomimetic specimens was discussed. The results indicated that the wear resistance of biomimetic specimens cladding TiC was the best; the specimens cladding WC or B4C were in the middle; and the specimens cladding Al2O3 was the worst. The sequence of friction coefficient values of biomimetic specimens cladding different ceramic powders from high to low was B4C, TiC, WC and Al2O3. The wear mechanism of untreated specimen was mainly adhesion wear, abrasive wear as well as the oxidation wear, whereas the adhesive wear and abrasive wear was the main wear mechanism of the regions of substrate in biomimetic specimens and slight adhesion, abrasive wear and fatigue wear on the regions of biomimetic units.

  10. In Situ Production of Hard Metal Matrix Composite Coating on Engineered Surfaces Using Laser Cladding Technique

    NASA Astrophysics Data System (ADS)

    Raza, Mohammad Shahid; Hussain, Manowar; Kumar, Vikash; Das, Alok Kumar

    2017-01-01

    The growing need for high wear-resistant surface with enhanced physical properties has led to extensive researches in the field of surface engineering. Laser cladding emerged to be a promising method to achieve these objectives in a cost-effective way. The present paper studies the viability of cladding of tungsten disulfide (WS2) powder by using 400 W continuous-wave fiber laser. WS2 was used as a coating material, which was decomposed at higher temperature and underwent several chemical reactions. By this process, in situ formation of metal matrix composites and hard face coating on the substrate surface were attained. The characterization of laser cladded surface was done to study its morphological, microstructural, mechanical and tribological properties. It was observed that cladding of WS2 powder on 304 SS resulted in the formation of Cr-W-C-Fe metal matrix composite having improved mechanical and tribological properties. The value of microhardness of the coated surface was found to increase three to four times in comparison with the parent material surface. Wear test results indicated a decrease in wear by 1/9th (maximum) as compared to the parent 304 SS surface. The volume fractions of tungsten particles on the cladded surface were also investigated through EDS analysis.

  11. Technical Development on Burn-up Credit for Spent LWR Fuel

    SciTech Connect

    Gauld, I.C.

    2001-12-26

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

  12. Development of Technical Basis for Burnup Credit Regulatory Guidance in the United States

    SciTech Connect

    Parks, Cecil V; Wagner, John C; Mueller, Don; Gauld, Ian C

    2011-01-01

    In the United States (U.S.) there has been and continues to be considerable interest in the increased use of burnup credit as part of the safety basis for SNF systems and this interest has motivated numerous technical studies related to the application of burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission initiated a burnup credit research program, with support from the Oak Ridge National Laboratory, to develop regulatory guidance and the supporting technical basis for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. The objective of this paper is to summarize the work and significant accomplishments, with references to the technical reports and publications for complete details.

  13. High Burnup Effects Program A State-of-the-Technology Assessment

    SciTech Connect

    Rising, K. H.; Bradley, E. R.; Williford, R. E.; Freshley, M D.

    1982-06-01

    Various analytical models and empirical correlations describing the fission gas release phenomenon were examined. An evaluation was made of the current pertinent experimental data on the subject of high burnup fission gas release. Data reported by individual investigators were compared and evaluated in relation to their applicability to the content and scope of the High Burnup Effects Program. These evaluations then form the bases for defining the data needs, and the selection of variables to be studied in this program.

  14. LWR fuel-cycle costs as a function of burnup. Final report

    SciTech Connect

    Franks, W.; Goldstein, L.; Joseph, L.; Nikmohammadian, N.

    1984-11-01

    Utilities may be able to decrease fuel-cycle costs as much as 5% in PWRs and 6% in BWRs by increasing discharge burnup to optimum practical limits. With one exception, this analysis of 12- and 18-month fuel cycles indicated a potential for still further cost reductions at higher burnup rates than those considered (39 GWd/MtU for BWRs and 55 GWd/MtU for PWRs).

  15. High frequency acoustic microscopy for the determination of porosity and Young's modulus in high burnup uranium dioxide nuclear fuel

    SciTech Connect

    Marchetti, M.; Laux, D.; Cappia, F.; Laurie, M.; Van Uffelen, P.; Rondinella, V.V.; Despaux, G.

    2015-07-01

    During irradiation UO{sub 2} nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of the porosity and of elastic properties in high burnup UO{sub 2} pellet can be investigated via high frequency acoustic microscopy. Ultrasound waves are generated by a piezoelectric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave (VR). A 67 MWd/kgU UO{sub 2} pellet was characterized using the acoustic microscope installed in the hot cells of the Institute of Transuranium Elements: 90 MHz frequency was applied, methanol was used as coupling liquid and VR was measured at different radial positions. By comparing the porosity values obtained via acoustic microscopy with those determined using ceramographic image analysis a good agreement was found, especially in the areas close to the centre. In addition Young's modulus was calculated and its radial profile was correlated to the corresponding burnup profile. (authors)

  16. Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

    SciTech Connect

    Ketusky, E.

    2016-08-09

    The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

  17. Efficient operation of double-clad Yb3+-doped fiber lasers with a novel circular cladding geometry.

    PubMed

    Martínez-Rios, A; Starodumov, A N; Po, H; Wang, Y; Demidov, A A; Li, X

    2003-09-15

    A highly efficient double-clad Yb3+-doped fiber laser with a novel cladding geometry is described. A round double-clad fiber with a small D-shaped hole for breaking a circular symmetry in a cladding combines the advantages of ease of manufacture and handling round fibers with efficient absorption of pump light. Fiber lasers with a double-D cladding shape and a D-hole cladding are compared. We report what are to our knowledge the highest slope efficiencies of 73% and 69%, respectively, pumping at 915 nm. Output powers in excess of 13 W are demonstrated.

  18. Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

    SciTech Connect

    Enercon Services, Inc.

    2011-03-14

    Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in

  19. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    SciTech Connect

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria

  20. Design and optimization of the combination film in 10kW diode laser cladding source

    NASA Astrophysics Data System (ADS)

    Zhu, Hong-bo; Lin, Xing-chen; Hao, Ming-ming; Zhang, Jin-sheng; Ning, Yong-qiang

    2015-08-01

    According to the special requirements of combination film in 10kW diode laser cladding source, the polarization combination film at 915nm was designed and grew. Film system is designed at different film materials based on the design theory. The non-QWOT film is optimized using the needle optimization and double sided coating by Optilayer software. The film was used in the 10kW diode laser source after high temperature aging testing. The film formed by Ta2O5 is very stable under IBAD, which can meet the reliability of 10kW diode laser cladding source in industry

  1. Cladding waveguide gratings in standard single-mode fiber for 3D shape sensing.

    PubMed

    Waltermann, Christian; Doering, Alexander; Köhring, Michael; Angelmahr, Martin; Schade, Wolfgang

    2015-07-01

    Femtosecond laser pulses were used for the direct point-by-point inscription of waveguides into the cladding of standard single-mode fibers. Homogeneous S-shaped waveguides have been processed as a bundle of overlapping lines without damaging the surrounding material. Within these structures, FBGs have been successfully inscribed and characterized. A sensor device to measure the bending direction of a fiber was created by two perpendicular inscribed cladding waveguides with FBG. Finally, a complete 3D shape sensor consisting of several bending sensor planes, capable of detecting bending radii even below 2.5 cm is demonstrated.

  2. Laser cladding of stellite {number_sign}6: A detailed analysis

    SciTech Connect

    Kathuria, Y.P.; Tsuboi, Akihiko

    1996-12-31

    Potential applications such as hardfacing of engine valve seat, turbine engine blade shroud interlock and leading edge of steam turbine blade, with controlled dilution have stimulated the use of laser cladding recently. In this paper, based on the laser beam interaction time, the microscopical study of dendrite structure formation in laser cladding of stellite {number_sign}6 on mild steel and Cr-Ni base materials is investigated. The effect of the various parameters such as heat input, beam interaction time, scanning frequency and transverse speed are considered. The applications in the rapid prototyping industries are also discussed.

  3. Cladding of the carbon fiber on the steel base using electron beam in the air atmosphere

    NASA Astrophysics Data System (ADS)

    Losinskaya, A.; Lozhkina, E.; Bardin, A.; Stepanova, N.

    2016-11-01

    The formation of the high-carbon layers on the low-carbon steel (0.18 % C) using the method of electron-beam partial melting of the carbon fibers is considered. A 1.4 MeV electron beam extracted into air was used. The features of the cladded layers formation using different binders for a reliable fixation of the cladding material are studied. It is revealed that the best results are obtained using the phenol-formaldehyde glue as the binder. A 3 mm thickness layers with 2.2 % C are shown to be formed.

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

    SciTech Connect

    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.

  5. Roll Casting of Aluminum Alloy Clad Strip

    SciTech Connect

    Nakamura, R.; Tsuge, H.; Haga, T.; Watari, H.; Kumai, S.

    2011-01-17

    Casting of aluminum alloy three layers of clad strip was tried using the two sets of twin roll casters, and effects of the casting parameters on the cladding conditions were investigated. One twin roll caster was mounted on the other twin roll caster. Base strip was 8079 aluminum alloy and overlay strips were 6022 aluminum alloy. Effects of roll-load of upper and lower casters and melt temperature of the lower caster were investigated. When the roll-load of the upper and lower caster was large enough, the overlay strip could be solidified and be connected. The overlay strip could be connected when the melt of the overlay strip cast by the lower caster was low enough. Sound three layers of clad strip could be cast by proper conditions.

  6. An empirical-statistical model for coaxial laser cladding of NiCrAlY powder on Inconel 738 superalloy

    NASA Astrophysics Data System (ADS)

    Ansari, M.; Shoja Razavi, R.; Barekat, M.

    2016-12-01

    In this study, coaxial laser cladding of NiCrAlY powder on a nickel-based superalloy is investigated from an experimental point of view so as to propose an empirical-statistical model for the process. The correlations between main processing parameters (i.e. scanning speed, powder feeding rate, and laser power) and geometrical characteristics (i.e. width, height, penetration depth, dilution and wetting angle) of single clad tracks have been predicted and are discussed using regression analysis (RA). The validity of the predictions is confirmed by providing correlation coefficient and analysis of the residuals. The correlations are established as a combined parameter (PαVβFγ) for each studied characteristic of single clad tracks. These correlations finally lead to the design of a processing map that can be practically used to select proper processing parameters for laser cladding of the particular material.

  7. Silicon micropowder negative electrode endures more than 1000 cycles when a surface-roughened clad current collector is used

    NASA Astrophysics Data System (ADS)

    Kataoka, Riki; Oda, Yoshimitsu; Inoue, Ryouji; Kawasaki, Norioki; Takeichi, Nobuhiko; Kiyobayashi, Tetsu

    2017-04-01

    A surface-roughened clad (S-clad) current collector significantly extends the cycle-life of the lithium ion negative electrode composed of a silicon micropowder and an aqueous binder. The high tensile strength of the S-clad is also proved to be important for improving the battery performance of the electrode by comparison to a surface-roughened pure Cu current collector. Moreover, adding 10 vol.% fluoroethylene carbonate to the electrolyte further extends the cycle-life of the Si electrode. The synergic effect of the high adhesive and tensile strength of the S-clad current collector as well as the electrolyte additive results in maintaining the reversible capacity of 1000 mA h g-1 for more than 1000 cycles, in which 1.0-1.2 mg cm-2 of the active material is loaded on the electrode.

  8. Effect of neutron irradiation on the mechanical properties of weld overlay cladding for reactor pressure vessel

    NASA Astrophysics Data System (ADS)

    Tobita, Tohru; Udagawa, Makoto; Chimi, Yasuhiro; Nishiyama, Yutaka; Onizawa, Kunio

    2014-09-01

    This study investigates the effects of high fluence neutron irradiation on the mechanical properties of two types of cladding materials fabricated using the submerged-arc welding and electroslag welding methods. The tensile tests, Charpy impact tests, and fracture toughness tests were conducted before and after the neutron irradiation with a fluence of 1 × 1024 n/m2 at 290 °C. With neutron irradiation, we could observe an increase in the yield strength and ultimate strength, and a decrease in the total elongation. All cladding materials exhibited ductile-to-brittle transition behavior during the Charpy impact tests. A reduction in the Charpy upper-shelf energy and an increase in the ductile-to-brittle transition temperature was observed with neutron irradiation. There was no obvious decrease in the elastic-plastic fracture toughness (JIc) of the cladding materials upon irradiation with high neutron fluence. The tearing modulus was found to decrease with neutron irradiation; the submerged-arc-welded cladding materials exhibited low JIc values at high temperatures.

  9. Phosphate-core silica-clad Er/Yb-doped optical fiber and cladding pumped laser.

    PubMed

    Egorova, O N; Semjonov, S L; Velmiskin, V V; Yatsenko, Yu P; Sverchkov, S E; Galagan, B I; Denker, B I; Dianov, E M

    2014-04-07

    We present a composite optical fiber with a Er/Yb co-doped phosphate-glass core in a silica glass cladding as well as cladding pumped laser. The fabrication process, optical properties, and lasing parameters are described. The slope efficiency under 980 nm cladding pumping reached 39% with respect to the absorbed pump power and 28% with respect to the coupled pump power. Due to high doping level of the phosphate core optimal length was several times shorter than that of silica core fibers.

  10. Investigation of the Effect of Fixed Absorbers on the Reactivity of PWR Spent Nuclear Fuel for Burnup Credit

    SciTech Connect

    Wagner, John C.; Sanders, Charlotta E.

    2002-08-15

    The effect of fixed absorbers on the reactivity of pressurized water reactor (PWR) spent nuclear fuel (SNF) in support of burnup-credit criticality safety analyses is examined. A fuel assembly burned in conjunction with fixed absorbers may have a higher reactivity for a given burnup than an assembly that has not used fixed absorbers. As a result, guidance on burnup credit, issued by the U.S. Nuclear Regulatory Commission's Spent Fuel Project Office, recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommendation eliminates a large portion of the currently discharged SNF from loading in burnup credit casks and thus severely limits the practical usefulness of burnup credit. Therefore, data are needed to support the extension of burnup credit to additional SNF. This research investigates the effect of various fixed absorbers, including integral burnable absorbers, burnable poison rods, control rods, and axial power shaping rods, on the reactivity of PWR SNF. Trends in reactivity with relevant parameters (e.g., initial fuel enrichment, burnup and absorber type, exposure, and design) are established, and anticipated reactivity effects are quantified. Where appropriate, recommendations are offered for addressing the reactivity effects of the fixed absorbers in burnup-credit safety analyses.

  11. Corrosion Resistant Cladding by YAG Laser Welding in Underwater Environment

    SciTech Connect

    Tsutomi Kochi; Toshio Kojima; Suemi Hirata; Ichiro Morita; Katsura Ohwaki

    2002-07-01

    It is known that stress-corrosion cracking (SCC) will occur in nickel-base alloys used in Reactor Pressure Vessel (RPV) and Internals of nuclear power plants. A SCC sensitivity has been evaluated by IHI in each part of RPV and Internals. There are several water level instrumentation nozzles installed in domestic BWR RPV. In water level instrumentation nozzles, 182 type nickel-base alloys were used for the welding joint to RPV. It is estimated the SCC potential is high in this joint because of a higher residual stress than the yield strength (about 400 MPa). This report will describe a preventive maintenance method to these nozzles Heat Affected Zone (HAZ) and welds by a corrosion resistant cladding (CRC) by YAG Laser in underwater environment (without draining a reactor water). There are many kinds of countermeasures for SCC, for example, Induction Heating Stress Improvement (IHSI), Mechanical Stress Improvement Process (MSIP) and so on. A YAG laser CRC is one of them. In this technology a laser beam is used for heat source and irradiated through an optical fiber to a base metal and SCC resistant material is used for welding wires. After cladding the HAZ and welds are coated by the corrosion resistant materials so their surfaces are improved. A CRC by gas tungsten arc welding (GTAW) in an air environment had been developed and already applied to a couple of operating plants (16 Nozzles). This method was of course good but it spent much time to perform because of an installation of some water-proof working boxes to make a TIG-weldability environment. CRC by YAG laser welding in underwater environment has superior features comparing to this conventional TIG method as follows. At the viewpoint of underwater environment, (1) an outage term reduction (no drainage water). (2) a radioactive exposure dose reduction for personnel. At that of YAG laser welding, (1) A narrower HAZ. (2) A smaller distortion. (3) A few cladding layers. A YAG laser CRC test in underwater

  12. Microindentation hardness evaluation of iridium alloy clad vent set cups

    NASA Astrophysics Data System (ADS)

    Ulrich, George B.; DeRoos, Larry F.; Stinnette, Samuel E.

    1993-01-01

    An iridium alloy, DOP-26, is used as cladding for 238PuO2 fuel in radioisotope heat sources for space power systems. Presently, DOP-26 iridium alloy clad vent sets (CVS) are being manufactured at the Oak Ridge Y-12 Plant for potential use in the National Aeronautics and Space Administration's Cassini mission to Saturn. Wrought/ground/stress relieved blanks are warm formed into CVS cups. These cups are then annealed to recrystallize the material for subsequent fabrication/assembly operations as well as for final use. One of the cup manufacturing certification requirements is to test for Vickers microindentation hardness. New microindentation hardness specification limits, 210 to 310 HV, have been established for a test load of 1000 grams-force (gf). The original specification limits, 250 to 350 HV, were for 200 gf testing. The primary reason for switching to a higher test load was to reduce variability in the test data. The DOP-26 alloy exhibits microindentation hardness load dependence, therefore, new limits were needed for 1000 gf testing. The new limits were established by testing material from 15 CVS cups using 200 gf and 1000 gf loads and then statistically analyzing the data. Additional work using a Knoop indenter and a 10 gf load indicated that the DOP-26 alloy grain boundaries have higher hardnesses than the grain interiors.

  13. Microindentation hardness evaluation of iridium alloy clad vent set cups

    SciTech Connect

    Ulrich, G.B.; DeRoos, L.F.; Stinnette, S.E. )

    1993-01-15

    An iridium alloy, DOP-26, is used as cladding for [sup 238]PuO[sub 2] fuel in radioisotope heat sources for space power systems. Presently, DOP-26 iridium alloy clad vent sets (CVS) are being manufactured at the Oak Ridge Y-12 Plant for potential use in the National Aeronautics and Space Administration's Cassini mission to Saturn. Wrought/ground/stress relieved blanks are warm formed into CVS cups. These cups are then annealed to recrystallize the material for subsequent fabrication/assembly operations as well as for final use. One of the cup manufacturing certification requirements is to test for Vickers microindentation hardness. New microindentation hardness specification limits, 210 to 310 HV, have been established for a test load of 1000 grams-force (gf). The original specification limits, 250 to 350 HV, were for 200 gf testing. The primary reason for switching to a higher test load was to reduce variability in the test data. The DOP-26 alloy exhibits microindentation hardness load dependence, therefore, new limits were needed for 1000 gf testing. The new limits were established by testing material from 15 CVS cups using 200 gf and 1000 gf loads and then statistically analyzing the data. Additional work using a Knoop indenter and a 10 gf load indicated that the DOP-26 alloy grain boundaries have higher hardnesses than the grain interiors.

  14. Microindentation hardness evaluation of iridium alloy clad vent set cups

    SciTech Connect

    Ulrich, G.B.; DeRoos, L.F.; Stinnette, S.E.

    1992-05-15

    An iridium alloy, DOP-26, is used as cladding for {sup 238}PuO{sup 2} fuel in radioisotope heat sources for space power systems. Presently, DOP-26 iridium alloy clad vent sets (CVS) are being manufactured at the Oak Ridge Y-12 Plant for potential use in the National Aeronautics and Space Administration`s Cassini mission to Saturn. Wrought/ground/stress relieved blanks are warm formed into CVS cups. These cups are then annealed to recrystallize the material for subsequent fabrication/assembly operations as well as for final use. One of the cup manufacturing certification requirements is to test for Vickers microindentation hardness. New microindentation hardness specification limits, 210 to 310 HV, have been established for a test load of 1000 grams-force (gf). The original specification limits, 250 to 350 HV, were for 200 gf testing. The primary reason for switching to a higher test load was to reduce variability in the test data. The DOP-26 alloy exhibits microindentation hardness load dependence, therefore, new limits were needed for 1000 gf testing. The new limits were established by testing material from 15 CVS cups using 200 gf and 1000 gf loads and then statistically analyzing the data. Additional work using a Knoop indenter and a 10 gf load indicated that the DOP-26 alloy grain boundaries have higher hardnesses than the grain interiors.

  15. Microindentation hardness evaluation of iridium alloy clad vent set cups

    SciTech Connect

    Ulrich, G.B.; DeRoos, L.F.; Stinnette, S.E.

    1992-05-15

    An iridium alloy, DOP-26, is used as cladding for {sup 238}PuO{sup 2} fuel in radioisotope heat sources for space power systems. Presently, DOP-26 iridium alloy clad vent sets (CVS) are being manufactured at the Oak Ridge Y-12 Plant for potential use in the National Aeronautics and Space Administration's Cassini mission to Saturn. Wrought/ground/stress relieved blanks are warm formed into CVS cups. These cups are then annealed to recrystallize the material for subsequent fabrication/assembly operations as well as for final use. One of the cup manufacturing certification requirements is to test for Vickers microindentation hardness. New microindentation hardness specification limits, 210 to 310 HV, have been established for a test load of 1000 grams-force (gf). The original specification limits, 250 to 350 HV, were for 200 gf testing. The primary reason for switching to a higher test load was to reduce variability in the test data. The DOP-26 alloy exhibits microindentation hardness load dependence, therefore, new limits were needed for 1000 gf testing. The new limits were established by testing material from 15 CVS cups using 200 gf and 1000 gf loads and then statistically analyzing the data. Additional work using a Knoop indenter and a 10 gf load indicated that the DOP-26 alloy grain boundaries have higher hardnesses than the grain interiors.

  16. Some recent studies on laser cladding and dissimilar welding

    NASA Astrophysics Data System (ADS)

    Kaul, Rakesh; Ganesh, P.; Paul, C. P.; Albert, S. K.; Mudali, U. Kamachi; Nath, A. K.

    2006-01-01

    Indigenous development of high power CO II laser technology and industrial application of lasers represent two important mandates of the laser program, being pursued at Centre for Advanced Technology (CAT), India. The present paper describes some of the important laser material processing studies, involving cladding and dissimilar welding, performed in authors' laboratory. The first case study describes how low heat input characteristics of laser cladding process has been successfully exploited for suppressing dilution in "Colmonoy6" (a nickel-base hardfacing alloy) deposits on austenitic stainless steel components. Crack free hardfaced deposits were obtained by controlling heating and cooling rates associated with laser treatment. The results show significant advantage over Colmonoy 6 deposits made by GTAW, where a 2.5 mm thick region of dilution (with reduced hardness) develops next to substrateiclad interface. The next work involves laser-assisted deposition of graded "Stellite6" (a Co-base hardfacing alloy) with smooth transition in chemical composition and hardness for enhanced resistance against cracking, esp. under thermal cycling conditions. The following two case studies demonstrate significant improvement in corrosion properties of type 304L stainless steel by laser surface alloying, achieved through cladding route. The following case study demonstrates engineering of fusion zone microstructure of end plug dissimilar weld (between alloy D9 and type 3 16M stainless steel) by controlled preferential displacement of focused laser beam, which, in-turn, enhanced its resistance against solidification cracking. Crater appearing at the termination point of laser weld is also eliminated by ramping of laser power towards the end of laser welding. The last case study involves engineering of fusion zone microstructure of dissimilar laser weld between type 304 austenitic stainless steel and stabilized 17%Cr ferritic stainless steel by controlling welding parameters.

  17. Improved LWR Cladding Performance by EPD Surface Modification Technique

    SciTech Connect

    Corradini, Michael; Sridharan, Kumar

    2012-11-26

    This project will utilize the electro-phoretic deposition technique (EPD) in conjunction with nanofluids to deposit oxide coatings on prototypic zirconium alloy cladding surfaces. After demonstrating that this surface modification is reproducible and robust, the team will subject the modified surface to boiling and corrosion tests to characterize the improved nucleate boiling behavior and superior corrosion performance. The scope of work consists of the following three tasks: The first task will employ the EPD surface modification technique to coat the surface of a prototypic set of zirconium alloy cladding tube materials (e.g. Zircaloy and advanced alloys such as M5) with a micron-thick layer of zirconium oxide nanoparticles. The team will characterize the modified surface for uniformity using optical microscopy and scanning-electron microscopy, and for robustness using standard hardness measurements. After zirconium alloy cladding samples have been prepared and characterized using the EPD technique, the team will begin a set of boiling experiments to measure the heat transfer coefficient and critical heat flux (CHF) limit for each prepared sample and its control sample. This work will provide a relative comparison of the heat transfer performance for each alloy and the surface modification technique employed. As the boiling heat transfer experiments begin, the team will also begin corrosion tests for these zirconium alloy samples using a water corrosion test loop that can mimic light water reactor (LWR) operational environments. They will perform extended corrosion tests on the surface-modified zirconium alloy samples and control samples to examine the robustness of the modified surface, as well as the effect on surface oxidation

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

    NASA Astrophysics Data System (ADS)

    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.

  19. Criticality reference benchmark calculations for burnup credit using spent fuel isotopics

    SciTech Connect

    Bowman, S.M.

    1991-04-01

    To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as burnup credit,'' the fuel loading capacity of these casks can be increased. One of the difficulties in implementing burnup credit in criticality analyses is that there have been no critical experiments performed with spent fuel which can be used for computer code validation. In lieu of that, a reference problem set of fresh fuel critical experiments which model various conditions typical of light water reactor (LWR) transportation and storage casks has been identified and used in the validation of SCALE-4. This report documents the use of this same problem set to perform spent fuel criticality benchmark calculations by replacing the actual fresh fuel isotopics from the experiments with six different sets of calculated spent fuel isotopics. The SCALE-4 modules SAS2H and CSAS4 were used to perform the analyses. These calculations do not model actual critical experiments. The calculated k-effectives are not supposed to equal unity and will vary depending on the initial enrichment and burnup of the calculated spent fuel isotopics. 12 refs., 11 tabs.

  20. Model for evolution of grain size in the rim region of high burnup UO2 fuel

    NASA Astrophysics Data System (ADS)

    Xiao, Hongxing; Long, Chongsheng; Chen, Hongsheng

    2016-04-01

    The restructuring process of the high burnup structure (HBS) formation in UO2 fuel results in sub-micron size grains that accelerate the fission gas swelling, which will raise some concern over the safety of extended the nuclear fuel operation life in the reactor. A mechanistic and engineering model for evolution of grain size in the rim region of high burnup UO2 fuel based on the experimental observations of the HBS in the literature is presented. The model takes into account dislocations evolution under irradiation and the grain subdivision occur successively at increasing local burnup. It is assumed that the original driving force for subdivision of grain in the HBS of UO2 fuel is the production and accumulation of dislocation loops during irradiation. The dislocation loops can also be annealed through thermal diffusion when the temperature is high enough. The capability of this model is validated by the comparison with the experimental data of temperature threshold of subdivision, dislocation density and sub-grain size as a function of local burnup. It is shown that the calculated results of the dislocation density and subdivided grain size as a function of local burnup are in good agreement with the experimental results.

  1. Prism coupling into clad uniform optical waveguides

    SciTech Connect

    Revelli, J.F.; Sarid, D.

    1980-07-01

    The theory of prism coupling into multilayered dielectric slab waveguides is presented. In addition to including the possibility of high index cladding, the present theory also extends the region of validity of previously reported work to cover the regime of ''strong coupling''. The limiting conditions for validity of the present theory are that both ..cap alpha../sub m//k and ..cap alpha../sub m/ +- p/k be much smaller than either unity or vertical-bar..beta../sub m/-..beta../sub m/ +- pvertical-bar, where m is the mode under consideration, ..cap alpha../sub m/ is the leakage of that mode, and vertical-bar..beta../sub m/-..beta../sub m/ +- pvertical-bar is the separation of the effective indices of adjacent modes. A numerical example is presented in which the coupling efficiency into a uniform or slab waveguide with ..delta..n=0.002 is calculated for various cladding thicknesses with a cladding index of 2.5. The introduction of cladding is found to reduce coupling efficiency in this example due to increased phase mismatch between the incident and ''ideal'' beams.

  2. CLAD/BCLAD Examinations. Test Specifications.

    ERIC Educational Resources Information Center

    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…

  3. Cladding For Transversely-Pumped Laser Rod

    NASA Technical Reports Server (NTRS)

    Byer, Robert L.; Fan, Tso Yee

    1989-01-01

    Combination of suitable dimensioning and cladding of neodymium:yttrium aluminum garnet of similar solid-state laser provides for more efficient utilization of transversely-incident pump light from diode lasers. New design overcomes some of limitations of longitudinal- and older transverse-pumping concepts and promotes operation at higher output powers in TEM00 mode.

  4. Sensitivity and Uncertainty Analysis to Burnup Estimates on ADS using the ACAB Code

    SciTech Connect

    Cabellos, O.; Sanz, J.; Rodriguez, A.; Gonzalez, E.; Embid, M.; Alvarez, F.; Reyes, S.

    2005-05-24

    Within the scope of the Accelerator Driven System (ADS) concept for nuclear waste management applications, the burnup uncertainty estimates due to uncertainty in the activation cross sections (XSs) are important regarding both the safety and the efficiency of the waste burning process. We have applied both sensitivity analysis and Monte Carlo methodology to actinides burnup calculations in a lead-bismuth cooled subcritical ADS. The sensitivity analysis is used to identify the reaction XSs and the dominant chains that contribute most significantly to the uncertainty. The Monte Carlo methodology gives the burnup uncertainty estimates due to the synergetic/global effect of the complete set of XS uncertainties. These uncertainty estimates are valuable to assess the need of any experimental or systematic re-evaluation of some uncertainty XSs for ADS.

  5. Spent fuel pool storage calculations using the ISOCRIT burnup credit tool

    SciTech Connect

    Kucukboyaci, Vefa; Marshall, William BJ J

    2012-01-01

    In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse's state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc.) with a quick turnaround.

  6. On the oxidation state of UO 2 nuclear fuel at a burn-up of around 100 MWd/kgHM

    NASA Astrophysics Data System (ADS)

    Walker, C. T.; Rondinella, V. V.; Papaioannou, D.; Winckel, S. Van; Goll, W.; Manzel, R.

    2005-10-01

    Results for the radial distribution of the oxygen potential and stoichiometry of a PWR fuel with an average pellet burn-up of 102 MWd/kgHM are presented. The local Δ G bar (O2) of the fuel was measured using a miniature solid state galvanic cell, the local O/U ratio was calculated from the lattice parameter measured by micro-X-ray diffraction and the local O/M ratio was derived from the fuel composition determined by ICP-MS. During irradiation the O/U ratio of the fuel decreased from 2.005 to 1.991 ± 0.008. The average fuel O/M ratio was 1.973 compared with the stoichiometric value of 1.949. The amount of free oxygen in the fuel, represented by the difference between these two quantities, increased from the centre to periphery of the pellet. Similarly, the Δ G bar (O2) of the fuel increased from -370 kJ mol-1 at r/r0 = 0.1 to -293 kJ mol-1 at r/r0 = 0.975. Thus, the Δ G bar (O2) of the fuel had not been buffered by the oxidation of fission product Mo. About one-quarter of the free oxygen accumulated during the irradiation had been gettered by the Zircaloy cladding.

  7. Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

    SciTech Connect

    Mueller, Don; Bowen, Douglas G; Marshall, William BJ J

    2015-01-01

    The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members accept the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δkeff (ISG-8, Rev. 3, Recommendation 4).

  8. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    SciTech Connect

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

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

    NASA Astrophysics Data System (ADS)

    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

  10. High critical currents in iron-clad superconducting MgB2 wires.

    PubMed

    Jin, S; Mavoori, H; Bower, C; van Dover, R B

    2001-05-31

    Technically useful bulk superconductors must have high transport critical current densities, Jc, at operating temperatures. They also require a normal metal cladding to provide parallel electrical conduction, thermal stabilization, and mechanical protection of the generally brittle superconductor cores. The recent discovery of superconductivity at 39 K in magnesium diboride (MgB2) presents a new possibility for significant bulk applications, but many critical issues relevant for practical wires remain unresolved. In particular, MgB2 is mechanically hard and brittle and therefore not amenable to drawing into the desired fine-wire geometry. Even the synthesis of moderately dense, bulk MgB2 attaining 39 K superconductivity is a challenge because of the volatility and reactivity of magnesium. Here we report the successful fabrication of dense, metal-clad superconducting MgB2 wires, and demonstrate a transport Jc in excess of 85,000 A cm-2 at 4.2 K. Our iron-clad fabrication technique takes place at ambient pressure, yet produces dense MgB2 with little loss of stoichiometry. While searching for a suitable cladding material, we found that other materials dramatically reduced the critical current, showing that although MgB2 itself does not show the 'weak-link' effect characteristic of the high-Tc superconductors, contamination does result in weak-link-like behaviour.

  11. Utilizing clad piping to improve process plant piping integrity, reliability, and operations

    SciTech Connect

    Chakravarti, B.

    1996-07-01

    During the past four years carbon steel piping clad with type 304L (UNS S30403) stainless steel has been used to solve the flow accelerated corrosion (FAC) problem in nuclear power plants with exceptional success. The product is designed to allow ``like for like`` replacement of damaged carbon steel components where the carbon steel remains the pressure boundary and type 304L (UNS S30403) stainless steel the corrosion allowance. More than 3000 feet of piping and 500 fittings in sizes from 6 to 36-in. NPS have been installed in the extraction steam and other lines of these power plants to improve reliability, eliminate inspection program, reduce O and M costs and provide operational benefits. This concept of utilizing clad piping in solving various corrosion problems in industrial and process plants by conservatively selecting a high alloy material as cladding can provide similar, significant benefits in controlling corrosion problems, minimizing maintenance cost, improving operation and reliability to control performance and risks in a highly cost effective manner. This paper will present various material combinations and applications that appear ideally suited for use of the clad piping components in process plants.

  12. 78 FR 67348 - Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-12

    ... Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research and... notice of request for public comment on its draft test plan for the High Burnup Dry Storage Cask Research... of dry storage cask systems and experiments would be beneficial. A draft test plan for the...

  13. Post Irradiation Examination Plan for High-Burnup Demonstration Project Sister Rods

    SciTech Connect

    Scaglione, John M; Montgomery, Rose; Bevard, Bruce Balkcom

    2016-04-01

    This test plan describes the experimental work to be implemented by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) to characterize high burnup (HBU) spent nuclear fuel (SNF) in conjunction with the High Burnup Dry Storage Cask Research and Development Project and serves to coordinate and integrate the multi-year experimental program to collect and develop data regarding the continued storage and eventual transport of HBU (i.e., >45 GWd/MTU) SNF. The work scope involves the development, performance, technical integration, and oversight of measurements and collection of relevant data, guided by analyses and demonstration of need.

  14. Progress of the RIA experiments with high burnup fuels and their evaluation in JAERI

    SciTech Connect

    Ishijima, Kiyomi; Fuketa, Toyoshi

    1997-01-01

    Recent results obtained in the NSRR power burst experiments with high burnup PWR fuel rods are described and discussed in this paper. Data concerning test condition, transient records during pulse irradiation and post irradiation examination are described. Another high burnup PWR fuel rod failed in the test HBO-5 at the slightly higher energy deposition than that in the test HBO-1. The failure mechanism of the test HBO-5 is the same as that of the test HBO-1, that is, hydride-assisted PCMI. Some influence of the thermocouples welding on the failure behavior of the HBO-5 rod was observed.

  15. Magnetically controllable silicon microring with ferrofluid cladding.

    PubMed

    El Amili, A; Souza, M C M M; Vallini, F; Frateschi, N C; Fainman, Y

    2016-12-01

    We experimentally investigate the application of magnetic fluids (MFs) on integrated silicon photonics. Using a ferrofluid-clad silicon microring resonator, we demonstrate active control of resonances by applying an external magnetic field. Relatively high loaded quality factors on the order of 6000 are achieved, despite the optical losses introduced by the magnetic nanoparticles. We demonstrate resonance shifts of 185 pm in response to a 110 Oe strong magnetic field, corresponding to an overall refractive index change of -3.2×10-3 for the cladding MF. The combination of MFs and integrated photonics could potentially lead to the development of magnetically controllable optical devices and ultra-compact cost-effective magnetic field sensors.

  16. Copper-Nickel Cladding on Stainless Steel

    DTIC Science & Technology

    2005-07-01

    steel,. Monel (65Ni/35Cu) alloy consumables should be used as they can tolerate more iron dilution from the steel than the 70-30 copper-nickel alloy ...Cooper Alloys , 400 , K-500 Stainless Steel - Tyles 302, 304, 321, 347 N ickel 200 Silver Braze Alloys Nickel-Chromium Alloy 600 Nickel-Aluminum Bronze 70...cladding of austenitic stainless steels may also offer some ballistic, non-magnetic, and electromagnetic signature advantages over current hull alloys and

  17. Methodology for Mechanical Property Testing on Fuel Cladding Using an Expanded Plug Wedge Test

    SciTech Connect

    Wang, Jy-An John; Jiang, Hao

    2013-08-01

    To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at 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 utilizing 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 specimen strain. It was also found that cladding strength could be determined from the test results. The basic approach of this test method is to apply an axial compressive load to a cylindrical plug of polyurethane (or other materials) fitted inside a short ring of the test material to achieve radial expansion of the specimen. The diameter increase of the specimen is used to calculate the circumferential strain accrued during the test. The other two basic measurements are total applied load and amount of plug compression (extension). A simple procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves. However, several deficiencies exist in this expanded-plug loading ring test, which will impact accuracy of test results and introduce potential shear failure of the specimen due to inherited large axial compressive stress from the expansion plug test. First of all, the highly non-uniform stress and strain distribution resulted in the gage section of the clad. To ensure reliable testing and test repeatability, the potential for highly non-uniform stress distribution or displacement/strain deformation has to be eliminated at the gage section of the specimen. Second, significant

  18. Microstructure and Mechanical Properties of Laser Clad and Post-cladding Tempered AISI H13 Tool Steel

    NASA Astrophysics Data System (ADS)

    Telasang, Gururaj; Dutta Majumdar, Jyotsna; Wasekar, Nitin; Padmanabham, G.; Manna, Indranil

    2015-05-01

    This study reports a detailed investigation of the microstructure and mechanical properties (wear resistance and tensile strength) of hardened and tempered AISI H13 tool steel substrate following laser cladding with AISI H13 tool steel powder in as-clad and after post-cladding conventional bulk isothermal tempering [at 823 K (550 °C) for 2 hours] heat treatment. Laser cladding was carried out on AISI H13 tool steel substrate using a 6 kW continuous wave diode laser coupled with fiber delivering an energy density of 133 J/mm2 and equipped with a co-axial powder feeding nozzle capable of feeding powder at the rate of 13.3 × 10-3 g/mm2. Laser clad zone comprises martensite, retained austenite, and carbides, and measures an average hardness of 600 to 650 VHN. Subsequent isothermal tempering converted the microstructure into one with tempered martensite and uniform dispersion of carbides with a hardness of 550 to 650 VHN. Interestingly, laser cladding introduced residual compressive stress of 670 ± 15 MPa, which reduces to 580 ± 20 MPa following isothermal tempering. Micro-tensile testing with specimens machined from the clad zone across or transverse to cladding direction showed high strength but failure in brittle mode. On the other hand, similar testing with samples sectioned from the clad zone parallel or longitudinal to the direction of laser cladding prior to and after post-cladding tempering recorded lower strength but ductile failure with 4.7 and 8 pct elongation, respectively. Wear resistance of the laser surface clad and post-cladding tempered samples (evaluated by fretting wear testing) registered superior performance as compared to that of conventional hardened and tempered AISI H13 tool steel.

  19. Surface protection of light metals by one-step laser cladding with oxide ceramics

    NASA Astrophysics Data System (ADS)

    Nowotny, S.; Richter, A.; Tangermann, K.

    1999-06-01

    Today, intricate problems of surface treatment can be solved through precision cladding using advanced laser technology. Metallic and carbide coatings have been produced with high-power lasers for years, and current investigations show that laser cladding is also a promising technique for the production of dense and precisely localized ceramic layers. In the present work, powders based on Al2O3 and ZrO2 were used to clad aluminum and titanium light alloys. The compact layers are up to 1 mm thick and show a nonporous cast structure as well as a homogeneous network of vertical cracks. The high adhesive strength is due to several chemical and mechanical bonding mechanisms and can exceed that of plasmasprayed coatings. Compared to thermal spray techniques, the material deposition is strictly focused onto small functional areas of the workpiece. Thus, being a precision technique, laser cladding is not recommended for large-area coatings. Examples of applications are turbine components and filigree parts of pump casings.

  20. 3D analysis of thermal and stress evolution during laser cladding of bioactive glass coatings.

    PubMed

    Krzyzanowski, Michal; Bajda, Szymon; Liu, Yijun; Triantaphyllou, Andrew; Mark Rainforth, W; Glendenning, Malcolm

    2016-06-01

    Thermal and strain-stress transient fields during laser cladding of bioactive glass coatings on the Ti6Al4V alloy basement were numerically calculated and analysed. Conditions leading to micro-cracking susceptibility of the coating have been investigated using the finite element based modelling supported by experimental results of microscopic investigation of the sample coatings. Consecutive temperature and stress peaks are developed within the cladded material as a result of the laser beam moving along the complex trajectory, which can lead to micro-cracking. The preheated to 500°C base plate allowed for decrease of the laser power and lowering of the cooling speed between the consecutive temperature peaks contributing in such way to achievement of lower cracking susceptibility. The cooling rate during cladding of the second and the third layer was lower than during cladding of the first one, in such way, contributing towards improvement of cracking resistance of the subsequent layers due to progressive accumulation of heat over the process.

  1. A deformation and thermodynamic model for hydride precipitation kinetics in spent fuel cladding

    SciTech Connect

    Stout, R.B.

    1989-10-01

    Hydrogen is contained in the Zircaloy cladding of spent fuel rods from nuclear reactors. All the spent fuel rods placed in a nuclear waste repository will have a temperature history that decreases toward ambient; and as a result, most all of the hydrogen in the Zircaloy will eventually precipitate as zirconium hydride platelets. A model for the density of hydride platelets is a necessary sub-part for predicting Zircaloy cladding failure rate in a nuclear waste repository. A model is developed to describe statistically the hydride platelet density, and the density function includes the orientation as a physical attribute. The model applies concepts from statistical mechanics to derive probable deformation and thermodynamic functionals for cladding material response that depend explicitly on the hydride platelet density function. From this model, hydride precipitation kinetics depend on a thermodynamic potential for hydride density change and on the inner product of a stress tensor and a tensor measure for the incremental volume change due to hydride platelets. The development of a failure response model for Zircaloy cladding exposed to the expected conditions in a nuclear waste repository is supported by the US DOE Yucca Mountain Project. 19 refs., 3 figs.

  2. Multiscale modeling of thermal conductivity of high burnup structures in UO2 fuels

    SciTech Connect

    Bai, Xian -Ming; Tonks, Michael R.; Zhang, Yongfeng; Hales, Jason D.

    2015-12-22

    The high burnup structure forming at the rim region in UO2 based nuclear fuel pellets has interesting physical properties such as improved thermal conductivity, even though it contains a high density of grain boundaries and micron-size gas bubbles. To understand this counterintuitive phenomenon, mesoscale heat conduction simulations with inputs from atomistic simulations and experiments were conducted to study the thermal conductivities of a small-grain high burnup microstructure and two large-grain unrestructured microstructures. We concluded that the phonon scattering effects caused by small point defects such as dispersed Xe atoms in the grain interior must be included in order to correctly predict the thermal transport properties of these microstructures. In extreme cases, even a small concentration of dispersed Xe atoms such as 10-5 can result in a lower thermal conductivity in the large-grain unrestructured microstructures than in the small-grain high burnup structure. The high-density grain boundaries in a high burnup structure act as defect sinks and can reduce the concentration of point defects in its grain interior and improve its thermal conductivity in comparison with its large-grain counterparts. Furthermore, an analytical model was developed to describe the thermal conductivity at different concentrations of dispersed Xe, bubble porosities, and grain sizes. Upon calibration, the model is robust and agrees well with independent heat conduction modeling over a wide range of microstructural parameters.

  3. Multiscale modeling of thermal conductivity of high burnup structures in UO2 fuels

    DOE PAGES

    Bai, Xian -Ming; Tonks, Michael R.; Zhang, Yongfeng; ...

    2015-12-22

    The high burnup structure forming at the rim region in UO2 based nuclear fuel pellets has interesting physical properties such as improved thermal conductivity, even though it contains a high density of grain boundaries and micron-size gas bubbles. To understand this counterintuitive phenomenon, mesoscale heat conduction simulations with inputs from atomistic simulations and experiments were conducted to study the thermal conductivities of a small-grain high burnup microstructure and two large-grain unrestructured microstructures. We concluded that the phonon scattering effects caused by small point defects such as dispersed Xe atoms in the grain interior must be included in order to correctlymore » predict the thermal transport properties of these microstructures. In extreme cases, even a small concentration of dispersed Xe atoms such as 10-5 can result in a lower thermal conductivity in the large-grain unrestructured microstructures than in the small-grain high burnup structure. The high-density grain boundaries in a high burnup structure act as defect sinks and can reduce the concentration of point defects in its grain interior and improve its thermal conductivity in comparison with its large-grain counterparts. Furthermore, an analytical model was developed to describe the thermal conductivity at different concentrations of dispersed Xe, bubble porosities, and grain sizes. Upon calibration, the model is robust and agrees well with independent heat conduction modeling over a wide range of microstructural parameters.« less

  4. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    SciTech Connect

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  5. Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

    SciTech Connect

    Gauld, I.C.

    2005-08-12

    U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k{sub eff}) to determine the net importance of cross sections to k{sub eff}. The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: {sup 151}Sm, {sup 103}Rh, {sup 155}Eu, {sup 150}Sm, {sup 152}Sm, {sup 153}Eu, {sup 154}Eu, and {sup 143}Nd.

  6. Cladded single crystal fibers for high power fiber lasers

    NASA Astrophysics Data System (ADS)

    Kim, W.; Shaw, B.; Bayya, S.; Askins, C.; Peele, J.; Rhonehouse, D.; Meyers, J.; Thapa, R.; Gibson, D.; Sanghera, J.

    2016-09-01

    We report on the recent progress in the development of cladded single crystal fibers for high power single frequency lasers. Various rare earth doped single crystal YAG fibers with diameters down to 17 μm with length > 1 m have been successfully drawn using a state-of-the-art Laser Heated Pedestal Growth system. Single and double cladding on rare earth doped YAG fibers have been developed using glasses where optical and physical properties were precisely matched to doped YAG core single crystal fiber. The double clad Yb:YAG fiber structures have dimensions analogous to large mode area (LMA) silica fiber. We also report successful fabrications of all crystalline core/clad fibers where thermal and optical properties are superior over glass cladded YAG fibers. Various fabrication methods, optical characterization and gain measurements on these cladded YAG fibers are reported.

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

    SciTech Connect

    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

  8. The Effect of Pitch, Burnup, and Absorbers on a TRIGA Spent-Fuel Pool Criticality Safety

    SciTech Connect

    Logar, Marjan; Jeraj, Robert; Glumac, Bogdan

    2003-02-15

    It has been shown that supercriticality might occur for some postulated accident conditions at the TRIGA spent-fuel pool. However, the effect of burnup was not accounted for in previous studies. In this work, the combined effect of fuel burnup, pitch among fuel elements, and number of uniformly mixed absorber rods for a square arrangement on the spent-fuel pool k{sub eff} is investigated.The Monte Carlo computer code MCNP4B with the ENDF-B/VI library and detailed three dimensional geometry was used. The WIMS-D code was used to model the isotopic composition of the standard TRIGA and FLIP fuel for 5, 10, 20 and 30% burnup level and 2- and 4-yr cooling time.The results show that out of the three studied effects, pitch from contact (3.75 cm) up to rack design pitch (8 cm), number of absorbers from zero to eight, and burnup up to 30%, the pitch has the greatest influence on the multiplication factor k{sub eff}. In the interval in which the pitch was changed, k{sub eff} decreased for up to {approx}0.4 for standard and {approx}0.3 for FLIP fuel. The number of absorber rods affects the multiplication factor much less. This effect is bigger for more compact arrangements, e.g., for contact of standard fuel elements with eight absorber rods among them, k{sub eff} values are smaller for {approx}0.2 ({approx}0.1 for FLIP) than for arrangements without absorber rods almost regardless of the burnup. The effect of burnup is the smallest. For standard fuel elements, it is {approx}0.1 for almost all pitches and numbers of absorbers. For FLIP fuel, it is smaller for a factor of 3, but increases with the burnup for compact arrangements. Cooling time of fuel has just a minor effect on the k{sub eff} of spent-fuel pool and can be neglected in spent-fuel pool design.

  9. Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

    SciTech Connect

    Wagner, J.C.

    2001-09-28

    The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs

  10. An interaction of austenitic Cr sbnd Ni steel cladding with Li sbnd Pb eutectic after In-reactor tests at temperatures 550 and 670 K

    NASA Astrophysics Data System (ADS)

    Kalachikov, V. E.; Kozlov, A. V.; Sinelnikov, L. P.; Zyrianov, A. P.; Abramov, V. Ya.; Kalinin, G. M.; Sidorenkov, A. V.

    1996-10-01

    A material science study of 0.04C sbnd 16Cr sbnd 11Ni sbnd 3Mo sbnd Ti steels cladding after static reactor testing in contact with a Li sbnd Pb eutectic melt at 550 and 670 K, was conducted. It was found that cladding plasticity was maintained at the level of 15.6-25.8% and corrosion did not exceed 0.16 mm/y under standard modes of testing. In several cases characteristics of the cladding material changed substantially. For example, plasticity reduced to 3.1-4.4% and the corrosion rate reached 1.6 mm/y. The results obtained led to the conclusion that non-standard parameters of these tests (plumbing of gas outlet by Li sbnd Pb eutectic, temperature splashes during sampling, etc.) had an essential influence on the observed cladding property degradation.

  11. Numerical simulation of heat transfer and fluid flow in coaxial laser cladding process for direct metal deposition

    NASA Astrophysics Data System (ADS)

    Qi, Huan; Mazumder, Jyotirmoy; Ki, Hyungson

    2006-07-01

    The coaxial laser cladding process is the heart of direct metal deposition (DMD). Rapid materials processing, such as DMD, is steadily becoming a tool for synthesis of materials, as well as rapid manufacturing. Mathematical models to develop the fundamental understanding of the physical phenomena associated with the coaxial laser cladding process are essential to further develop the science base. A three-dimensional transient model was developed for a coaxial powder injection laser cladding process. Physical phenomena including heat transfer, melting and solidification phase changes, mass addition, and fluid flow in the melt pool, were modeled in a self-consistent manner. Interactions between the laser beam and the coaxial powder flow, including the attenuation of beam intensity and temperature rise of powder particles before reaching the melt pool were modeled with a simple heat balance equation. The level-set method was implemented to track the free surface movement of the melt pool, in a continuous laser cladding process. The governing equations were discretized using the finite volume approach. Temperature and fluid velocity were solved for in a coupled manner. Simulation results such as the melt pool width and length, and the height of solidified cladding track were compared with experimental results and found to be reasonably matched.

  12. Transversely polarized source cladding for an optical fiber

    NASA Technical Reports Server (NTRS)

    Egalon, Claudio Oliveira (Inventor); Rogowski, Robert S. (Inventor)

    1994-01-01

    An optical fiber comprising a fiber core having a longitudinal symmetry axis is provided. An active cladding surrounds a portion of the fiber core and comprises light-producing sources which emit light in response to chemical or light excitation. The cladding sources are oriented transversely with respect to the longitudinal axis of the fiber core. This polarization results in a superior power efficiency compared to active cladding sources that are randomly polarized or longitudinally polarized parallel with the longitudinal symmetry axis.

  13. Clad fiber capacitor and method of making same

    DOEpatents

    Tuncer, Enis

    2012-12-11

    A clad capacitor and method of manufacture includes assembling a preform comprising a ductile, electrically conductive fiber; a ductile, electrically insulating cladding positioned on the fiber; and a ductile, electrically conductive sleeve positioned over the cladding. One or more preforms are then bundled, heated and drawn along a longitudinal axis to decrease the diameter of the ductile components of the preform and fuse the preform into a unitized strand.

  14. Development and characterisations of WC–12Co microwave clad

    SciTech Connect

    Zafar, Sunny Sharma, Apurbba Kumar

    2014-10-15

    In the present work, WC–12Co based cermet clad was developed on AISI 304 stainless steel using microwave hybrid heating technique. The experimental trials were carried out in a 1.4 kW industrial multimode microwave applicator. The paper explains the major events occurring during microwave irradiation and formation of clad. The developed clads were subsequently characterised through field emission scanning electron microscopy equipped with energy dispersive X-ray spectroscopy, X-ray diffraction, assessment of porosity and microhardness. The WC–12Co clads developed with an approximate thickness of 1 mm, illustrated excellent metallurgical bonding with substrate. The microstructure of the WC–12Co clad mainly consists of skeleton structured carbides embedded in tough metallic phase. The phase analysis of the developed clads indicate the presence of various stable and complex carbides like Co{sub 6}W{sub 6}C, Co{sub 3}W{sub 3}C and Fe{sub 6}W{sub 6}C. The uniform distribution of such carbides with skeleton-like morphology in the microstructure is indicative of high hardness of the clad. The developed clads were free from visible interfacial cracking and the clad porosity was found in the order of approximately 0.98%. The average microhardness of the WC–12Co microwave clads was observed to be 1135 ± 88 HV. - Highlights: • Microwave cladding of WC–12Co on AISI 304 stainless steel is carried out. • Skeleton-like structures of W–Co based carbides are embedded in metallic matrix. • Clad–substrate interface is free from un-melted and un-dissolved carbide particles. • Hardness of clad (1135 ± 88 HV) is 3.5 times that of the substrate (325 ± 49 HV)

  15. Benefits of the delta K of depletion benchmarks for burnup credit validation

    SciTech Connect

    Lancaster, D.; Machiels, A.

    2012-07-01

    Pressurized Water Reactor (PWR) burnup credit validation is demonstrated using the benchmarks for quantifying fuel reactivity decrements, published as 'Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty,' EPRI Report 1022909 (August 2011). This demonstration uses the depletion module TRITON available in the SCALE 6.1 code system followed by criticality calculations using KENO-Va. The difference between the predicted depletion reactivity and the benchmark's depletion reactivity is a bias for the criticality calculations. The uncertainty in the benchmarks is the depletion reactivity uncertainty. This depletion bias and uncertainty is used with the bias and uncertainty from fresh UO{sub 2} critical experiments to determine the criticality safety limits on the neutron multiplication factor, k{sub eff}. The analysis shows that SCALE 6.1 with the ENDF/B-VII 238-group cross section library supports the use of a depletion bias of only 0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertainty in the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the delta k of depletion ('Kopp memo') was shown to be conservative for fuel with more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertainty is not a function of burnup, the Kopp memo's recommended bias and uncertainty may be exceeded at low burnups, but its absolute magnitude is small. (authors)

  16. High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

    SciTech Connect

    2014-02-27

    EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

  17. Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs

    SciTech Connect

    Wagner, J.C.

    2002-12-17

    This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).

  18. OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

    SciTech Connect

    DeHart, M.D.; Parks, C.V.; Brady, M.C.

    1996-06-01

    In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155.

  19. OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

    SciTech Connect

    DeHart, M.D.

    1993-01-01

    Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are {sup 149}Sm, {sup 151}Sm, and {sup 155}Gd.

  20. Microstructural Characterization of Cermet Cladding Developed Through Microwave Irradiation

    NASA Astrophysics Data System (ADS)

    Gupta, Dheeraj; Sharma, Apurbba Kumar

    2012-10-01

    In the present work, cladding of hardfacing WC10Co2Ni powder on austenitic stainless steel has been developed through a novel processing technique. The clads were developed using microwave hybrid heating. The clad of average thickness ~2 mm has been developed through the exposure of microwave radiation at frequency 2.45 GHz and power 900 W for the duration of 360 s. The developed clads were characterized using field emission scanning electron microscope, X-ray elemental analysis, X-ray diffraction, and measurement of Vicker's microhardness. The microstructure study of the clad showed good metallurgical bonding with substrate and revealed that clads are free from any visible interface cracking. Clads were formed with partial dilution of a thin layer of the substrate. The cermet microstructure mainly consists of relatively soft metallic matrix phase and uniformly distributed hard carbide phase with skeleton-like structure. The developed clads exhibit an average microhardness of 1064 ± 99 Hv. The porosity of developed clad has been significantly less at approximately 0.89%.

  1. Hollow core anti-resonant fibres with split cladding

    NASA Astrophysics Data System (ADS)

    Huang, Xiaosheng; Qi, Wenliang; Ho, Daryl; Luan, Feng; Yong, Ken-Tye; Yoo, Seongwoo

    2016-03-01

    A split cladding fibers (SCF) is proposed as an additional design to the anti-resonant type fiber. The introduced split cladding helps to reduce the fabrication distortion. We use numerical simulations to compare the Kagome fibers (KFs) and the proposed split cladding fibers (SCFs) over two normalized transmission bands. It reveals that SCFs are able to maintain the desired round shape of silica cladding walls, hence improving the confinement loss (CL) compared to the KF. Fabrication of the SCF is demonstrated by the stack-and-draw technique. The near filed mode patterns are measured to prove the feasibility of this fiber design.

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

    NASA Astrophysics Data System (ADS)

    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.

  3. Mechanical behavior of aluminum-bearing ferritic alloys for accident-tolerant fuel cladding applications

    NASA Astrophysics Data System (ADS)

    Guria, Ankan

    Nuclear power currently provides about 13% of electrical power worldwide. Nuclear reactors generating this power traditionally use Zirconium (Zr) based alloys as the fuel cladding material. Exothermic reaction of Zr with steam under accident conditions may lead to production of hydrogen with the possibility of catastrophic consequences. Following the Fukushima-Daiichi incident, the exploration of accident-tolerant fuel cladding materials accelerated. Aluminum-rich (around 5 wt. %) ferritic steels such as Fecralloy, APMT(TM) and APM(TM) are considered as potential materials for accident-tolerant fuel cladding applications. These materials create an aluminum-based oxide scale protecting the alloy at elevated temperatures. Tensile deformation behavior of the above alloys was studied at different temperatures (25-500 °C) at a strain rate of 10-3 s-1 and correlated with microstructural characteristics. Higher strength and decent ductility of APMT(TM) led to further investigation of the alloy at various combination of strain rates and temperatures followed by fractography and detailed microscopic analyses. Serrations appeared in the stress-strain curves of APMT(TM) and Fecralloy steel tested in a limited temperature range (250-400 °C). The appearance of serrations is explained on the basis of dynamic strain aging (DSA) effect due to solute-dislocation interactions. The research in this study is being performed using the funds received from the US DOE Office of Nuclear Energy's Nuclear Energy University Programs (NEUP).

  4. Fracture behavior of shallow cracks in full-thickness clad beams from an RPV wall section

    SciTech Connect

    Keeney, J.A.; Bass, B.R.; McAfee, W.J.

    1995-04-01

    A testing program is described that utilizes full-thickness clad beam specimens to quantify fracture toughness for shallow cracks in weld material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPVs). The beam specimens are fabricated from an RPV shell segment that includes weld, plate and clad material. Metallurgical factors potentially influencing fracture toughness for shallow cracks in the beam specimens include material gradients and material inhomogeneities in welded regions. The shallow-crack clad beam specimens showed a significant loss of constraint similar to that of other shallow-crack single-edge notch bend (SENB) specimens. The stress-based Dodds-Anderson scaling model appears to be effective in adjusting the test data to account for in-plane loss of constraint for uniaxially tested beams, but cannot predict the observed effects of out-of-plane biaxial loading on shallow-crack fracture toughness. A strain-based dual-parameter fracture toughness correlation (based on plastic zone width) performed acceptably when applied to the uniaxial and biaxial shallow-crack fracture toughness data.

  5. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    SciTech Connect

    Marshall, William BJ J; Ade, Brian J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Mertyurek, Ugur; Radulescu, Georgeta

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  6. Thermomechanical loading applied on the cladding tube during the pellet cladding mechanical interaction phase of a rapid reactivity initiated accident

    NASA Astrophysics Data System (ADS)

    Hellouin de Menibus, Arthur; Sercombe, Jerome; Auzoux, Quentin; Poussard, Christophe

    2014-10-01

    Calculations of the CABRI REP-Na5 pulse were performed with the ALCYONE code in order to determine the evolution of the thermomechanical loading applied on the cladding tube during the Pellet-Cladding Mechanical Interaction (PCMI) phase of a rapid Reactivity Initiated Accident (RIA) initiated at 280 °C that lasted 8.8 ms. The evolution of the following parameters are reported: the cladding temperature, heating rate, strain rate and loading biaxiality. The impact of these parameters on the cladding mechanical behavior and fracture are then briefly reviewed.

  7. Optimization of Heat Treatments for Reversion of Strain-Induced Martensite in 304L SS Explosive Clad

    NASA Astrophysics Data System (ADS)

    Prasanthi, T. N.; Sudha, C.; Parida, P. K.; Dasgupta, Arup; Saroja, S.

    2016-02-01

    Explosive clad joints of 304L SS and Ti-5Ta-2Nb alloy, fabricated for an important application in the spent nuclear fuel reprocessing industry showed formation of deformation induced metastable α' martensite and fcc Ti phase in SS and TiTaNb alloy respectively. A biphasic structure consisting of metastable phases is not preferred for industrial applications due to degradation of corrosion and mechanical properties of the structural materials during service. Hence, it is essential to carry out post cladding heat treatments. The results reported in this paper provide evidence for the presence of α' phase in 304L SS in `as clad' joints and its reversion process during thermal exposure. The temperature window in the range of 400-700 °C and time was optimized based on complete transformation of the metastable phases to parent phases, and avoiding the formation of brittle Fe-Ti intermetallics at the interface. A systematic increase in the fraction of austenite phase associated with the reversion phenomena has been studied using electron back scattered diffraction and transmission electron microscopy. Orientation relationship between product fcc and parent bcc phases was found to obey the K-S relationship. The reverted γ phase was found to nucleate within the martensite laths. A temperature of 550 °C for duration of about 10 h was found to be optimum for the post cladding treatments of the explosive clad joints.

  8. Rheological evaluation of pretreated cladding removal waste

    SciTech Connect

    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.

  9. COMPARISON OF CLADDING CREEP RUPTURE MODELS

    SciTech Connect

    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.

  10. Air-leakage effects on stone cladding panels

    NASA Astrophysics Data System (ADS)

    Colantonio, Antonio

    1995-03-01

    This paper looks at the effects of air leakage on insulated stone clad precast panels used in present day construction of large commercial buildings. The building investigated was a newly built twenty story office building in a high density urban setting. Air leakage was suspected as a possible cause for thermal comfort complaints at isolated locations within the perimeter zones of the building. During the warrantee period the building owner asked for a quality control inspection of the air barrier assembly of the building envelope. Infrared thermography was used to locate areas of suspected air leakage within the building envelope. In order to differentiate thermal patterns produced by air leakage, conduction and convection as well as radiation from external sources, the building was inspected from the exterior; (1) after being pressurized for three hours, (2) one hour after the building was depressurized and (3) two and a half hours after total building depressurization was maintained by the building mechanical systems. Thermal images from similar locations were correlated for each time and pressure setting to verify air leakage locations within the building envelope. Areas exhibiting air leakage were identified and contractors were requested to carry out the necessary repairs. The pressure differential across the building envelope needs to be known in order to properly carry out an inspection to identify all locations of air leakage within a building envelope. As well the direction of the air movement and the density of the cladding material need to be accounted for in the proper inspection of these types of wall assemblies.

  11. SintClad: A New Approach for the Production of Wear-Resistant Tools

    NASA Astrophysics Data System (ADS)

    Blüm, M.; Hill, H.; Moll, H.; Weber, S.; Theisen, W.

    2012-05-01

    Tools used in the mineral processing industry are required to feature high wear resistance to facilitate an adequate cost efficiency. These kinds of tools are made of composite materials based on a low-alloyed substrate material and a high-alloyed coating. The coatings can be applied in different ways using production processes like HIP cladding, deposit welding, and composite casting. The article is concerned with the problem of a novel and cost-effective coating alternative: sinter cladding, using the principle of super-solidus liquid-phase sintering (SLPS). Usually SLPS represents a sintering technique, which is used for the compaction of high-alloyed metal powders. However, no recognizable efforts were made to use the SLPS-process for applying a PM-coating on a bulk substrate material. Sinter cladding for the first time uses SLPS to combine the process of powder compaction with the application of a coating to a solid steel substrate into one single step. Another advantage of the process is the possibility to produce massive bulk coatings with thicknesses exceeding 20 mm. This article is original in the scope of question and investigation methods in terms of microstructure, hardness profiles, EDX measurements, diffusion calculations, and computational thermodynamics.

  12. Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding

    SciTech Connect

    Besmann, T. M.; Yamamoto, Y.; Unocic, K. A.

    2016-06-21

    We assessed the thermochemical compatibility of potential replacement cladding materials for zirconium alloys in light water reactors. Considered were FeCrAl steel (similar to Kanthal APMT), Nb-1%Zr (similar to PWC-11), and a hybrid SiC-composite with a metallic barrier layer. The niobium alloy was also seen as requiring an oxidation protective layer, and a diffusion silicide was investigated. Metallic barrier layers for the SiC-composite reviewed included a FeCrAl alloy, Nb-1%Zr, and chromium. Thermochemical calculations were performed to determine oxidation behavior of the materials in steam, and for hybrid SiC-composites possible interactions between the metallic layer and SiC. Additionally, experimental exposures of SiC-alloy reaction couples at 673K, 1073K, and 1273K for 168 h in an inert atmosphere were made and microanalysis performed. Whereas all materials were determined to oxidize under higher oxygen partial pressures in the steam environment, these varied by material with expected protective oxides forming. Finally, the computed and experimental results indicate the formation of liquid phase eutectic in the FeCrAl-SiC system at the higher temperatures.

  13. Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding

    DOE PAGES

    Besmann, T. M.; Yamamoto, Y.; Unocic, K. A.

    2016-06-21

    We assessed the thermochemical compatibility of potential replacement cladding materials for zirconium alloys in light water reactors. Considered were FeCrAl steel (similar to Kanthal APMT), Nb-1%Zr (similar to PWC-11), and a hybrid SiC-composite with a metallic barrier layer. The niobium alloy was also seen as requiring an oxidation protective layer, and a diffusion silicide was investigated. Metallic barrier layers for the SiC-composite reviewed included a FeCrAl alloy, Nb-1%Zr, and chromium. Thermochemical calculations were performed to determine oxidation behavior of the materials in steam, and for hybrid SiC-composites possible interactions between the metallic layer and SiC. Additionally, experimental exposures of SiC-alloymore » reaction couples at 673K, 1073K, and 1273K for 168 h in an inert atmosphere were made and microanalysis performed. Whereas all materials were determined to oxidize under higher oxygen partial pressures in the steam environment, these varied by material with expected protective oxides forming. Finally, the computed and experimental results indicate the formation of liquid phase eutectic in the FeCrAl-SiC system at the higher temperatures.« less

  14. Hot Forging of a Cladded Component by Automated GMAW Process

    NASA Astrophysics Data System (ADS)

    Rafiq, Muhammad; Langlois, Laurent; Bigot, Régis

    2011-01-01

    Weld cladding is employed to improve the service life of engineering components by increasing corrosion and wear resistance and reducing the cost. The acceptable multi-bead cladding layer depends on single bead geometry. Hence, in first step, the relationship between input process parameters and the single bead geometry is studied and in second step a comprehensive study on multi bead clad layer deposition is carried out. This paper highlights an experimental study carried out to get single layer cladding deposited by automated Gas Metal Arc Welding (GMAW) process and to find the possibility of hot forming of the cladded work piece to get the final hot formed improved structure. GMAW is an arc welding process that uses an arc between a consumable electrode and the welding pool with an external shielding gas and the cladding is done by alongside deposition of weld beads. The experiments for single bead were conducted by varying the three main process parameters wire feed rate, arc voltage and welding speed while keeping other parameters like nozzle to work distance, shielding gas and its flow rate and torch angle constant. The effect of bead spacing and torch orientation on the cladding quality of single layer from the results of single bead deposition was studied. Effect of the dilution rate and nominal energy on the cladded layer hot bending quality was also performed at different temperatures.

  15. 77 FR 37439 - Clad Steel Plate From Japan; Notice of Commission Determination To Conduct a Full Five-Year Review

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-21

    ... From the Federal Register Online via the Government Publishing Office INTERNATIONAL TRADE COMMISSION Clad Steel Plate From Japan; Notice of Commission Determination To Conduct a Full Five-Year Review... plate from Japan would be likely to lead to continuation or recurrence of material injury within...

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

    SciTech Connect

    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.

  17. Influence of the tritium beta(-) decay on low-temperature thermonuclear burn-up in deuterium-tritium mixtures

    PubMed

    Frolov

    2000-09-01

    Low-temperature (Tburn-up in deuterium-tritium mixtures with various deuterium-tritium-helium-3 ratios is considered. The general dependence is studied for the critical burn-up parameter x(c)=rhor(c) upon the initial temperature T, density rho(0), and tritium molar concentration y for the [D]:y[T]:(1-y)[3He] mixture. In particular, it is shown that, if the tritium concentration y decreases, then the critical burn-up parameter x(c)(T,rho(0),y) grows very quickly (at fixed T and rho(0)). This means that tritium beta(-) decay significantly complicates thermonuclear burn-up in deuterium-tritium mixtures.

  18. The Challenges Associated with High Burnup and High Temperature for UO2 TRISO-Coated Particle Fuel

    SciTech Connect

    David Petti; John Maki

    2005-02-01

    The fuel service conditions for the DOE Next Generation Nuclear Plant (NGNP) will be challenging. All major fuel related design parameters (burnup, temperature, fast neutron fluence, power density, particle packing fraction) exceed the values that were qualified in the successful German UO2 TRISO-coated particle fuel development program in the 1980s. While TRISO-coated particle fuel has been irradiated at NGNP relevant levels for two or three of the design parameters, no data exist for TRISO-coated particle fuel for all five parameters simultaneously. Of particular concern are the high burnup and high temperatures expected in the NGNP. In this paper, where possible, we evaluate the challenges associated with high burnup and high temperature quantitatively by examining the performance of the fuel in terms of different known failure mechanisms. Potential design solutions to ameliorate the negative effects of high burnup and high temperature are also discussed.

  19. Bendability of aluminiumand steel-clad chromium plates

    NASA Astrophysics Data System (ADS)

    Yoshida, Fusahito; Okada, Tatsuo; Itoh, Misao; Harada, Yasunori; Ohmori, Masanobu

    1998-05-01

    The present paper describes how the cladding of chromium plate with dissimilar metals improves the plastic bendability of the chromium. Three-point bending tests at various temperatures were performed for three types of chromium specimens: a monolithic chromium plate, aluminium- and steel-clad chromium plate. The aluminium-clad chromium plate was bent at 343 K up to a bent angle of 90 degrees without failure, even when the chromium layer was located outside of the plate (tension side), while the monolithic chromium plate could be bent exclusively at temperatures above 403 K. When the chromium layer was located inside of the steel-clad chromium plate (compression side), the plate was successfully bent at 307 K. The FE stress analysis of bending proved that the cladding of chromium plates with proper metals of different kinds is effective to reduce the tensile stress in chromium induced during bending and also the residual stress existing after bending operation.

  20. Compact cladding-pumped planar waveguide amplifier and fabrication method

    DOEpatents

    Bayramian, Andy J.; Beach, Raymond J.; Honea, Eric; Murray, James E.; Payne, Stephen A.

    2003-10-28

    A low-cost, high performance cladding-pumped planar waveguide amplifier and fabrication method, for deployment in metro and access networks. The waveguide amplifier has a compact monolithic slab architecture preferably formed by first sandwich bonding an erbium-doped core glass slab between two cladding glass slabs to form a multi-layer planar construction, and then slicing the construction into multiple unit constructions. Using lithographic techniques, a silver stripe is deposited and formed at a top or bottom surface of each unit construction and over a cross section of the bonds. By heating the unit construction in an oven and applying an electric field, the silver stripe is then ion diffused to increase the refractive indices of the core and cladding regions, with the diffusion region of the core forming a single mode waveguide, and the silver diffusion cladding region forming a second larger waveguide amenable to cladding pumping with broad area diodes.

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

    SciTech Connect

    Mickalonis, J. I.

    2015-08-01

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

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

    SciTech Connect

    Mickalonis, J. I.

    2015-08-31

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

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

    SciTech Connect

    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.

  4. Microstructural modeling of thermal conductivity of high burn-up mixed oxide fuel

    NASA Astrophysics Data System (ADS)

    Teague, Melissa; Tonks, Michael; Novascone, Stephen; Hayes, Steven

    2014-01-01

    Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISON [1] fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez-Lucuta model was favorable.

  5. Microstructural Modeling of Thermal Conductivity of High Burn-up Mixed Oxide Fuel

    SciTech Connect

    Melissa Teague; Michael Tonks; Stephen Novascone; Steven Hayes

    2014-01-01

    Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISON fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez–Lucuta model was favorable.

  6. Experience with incomplete control rod insertion in fuel with burnup exceeding approximately 40 GWD/MTU

    SciTech Connect

    Kee, E.

    1997-01-01

    Analysis and measurement experience with fuel assemblies having incomplete control rod insertion at burnups of approximately 40 GWD/MTU is presented. Control rod motion dynamics and simplified structural analyses are presented and compared to measurement data. Fuel assembly growth measurements taken with the plant Refueling Machine Z-Tape are described and presented. Bow measurements (including plug gauging) are described and potential improvements are suggested. The measurements described and analysis performed show that sufficient guide tube bow (either from creep or yield buckling) is present in some high burnup assemblies to stop the control rods before they reach their full limit of travel. Recommendations are made that, if implemented, could improve cost performance related to testing and analysis activities.

  7. THE EFFECT OF BURNUP AND SEPARATION EFFICIENCY ON URANIUM UTILIZATION AND RADIOTOXICITY

    SciTech Connect

    Samuel Bays; Steven Piet

    2001-11-01

    This paper addresses two fundamental issues of fuel cycle sustainability. The two primary issues of interest are efficient use of the natural uranium resource (cradle), and management of nuclear waste radiotoxicity (grave). Both uranium utilization and radiotoxicity are directly influenced by the burnup achieved during irradiation (transmutation related) and where applicable the separation efficiency (partitioning related). Burnup influences the in-growth of transuranics by breeding them into the fuel cycle. Transuranic breeding is virtually essential to resource sustainability because it increases utilization of naturally abundant fertile U-238. However, the direct consequence of this build-up is the in-growth of transuranic isotopes which generally increase the source of future geologically committed radiotoxicity. For scenarios involving recycle, separation efficiency influences the degree to which this transuranic source term is removed from active service in the fuel stream and made a disposal legacy of human activity.

  8. Determination of deuterium–tritium critical burn-up parameter by four temperature theory

    SciTech Connect

    Nazirzadeh, M.; Ghasemizad, A.; Khanbabei, B.

    2015-12-15

    Conditions for thermonuclear burn-up of an equimolar mixture of deuterium-tritium in non-equilibrium plasma have been investigated by four temperature theory. The photon distribution shape significantly affects the nature of thermonuclear burn. In three temperature model, the photon distribution is Planckian but in four temperature theory the photon distribution has a pure Planck form below a certain cut-off energy and then for photon energy above this cut-off energy makes a transition to Bose-Einstein distribution with a finite chemical potential. The objective was to develop four temperature theory in a plasma to calculate the critical burn up parameter which depends upon initial density, the plasma components initial temperatures, and hot spot size. All the obtained results from four temperature theory model are compared with 3 temperature model. It is shown that the values of critical burn-up parameter calculated by four temperature theory are smaller than those of three temperature model.

  9. S∧4 Reactor: Operating Lifetime and Estimates of Temperature and Burnup Reactivity Coefficients

    NASA Astrophysics Data System (ADS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The S∧4 reactor has a sectored, Mo-14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor is loaded with UN fuel, cooled with a He-Xe gas mixture at ~1200 K and operates at steady thermal power of 550 kW. Following a launch abort accident, the axial and radial BeO reflectors easily disassemble upon impact so that the bare reactor is subcriticial when submerged in wet sand or seawater and the core voids are filled with seawater. Spectral Shift Absorber (SSA) additives have been shown to increase the UN fuel enrichment and significantly reduce the total mass of the reactor. This paper investigates the effects of SSA additions on the temperature and burnup reactivity coefficients and the operational lifetime of the S∧4 reactor. SSAs slightly decrease the temperature reactivity feedback coefficient, but significantly increase the operating lifetime by decreasing the burnup reactivity coefficient. With no SSAs, fuel enrichment is only 58.5 wt% and the estimated operating lifetime is the shortest (7.6 years) with the highest temperature and burnup reactivity feedback coefficients (-0.2709 ¢/K and -1.3470 $/atom%). With europium-151 and gadolinium-155 additions, the enrichment (91.5 and 94 wt%) and operating lifetime (9.9 and 9.8 years) of the S∧4 reactor are the highest while the temperature and burnup reactivity coefficients (-0.2382 and -0.2447 ¢/K -0.9073 and 0.8502 $/atom%) are the lowest.

  10. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

    SciTech Connect

    Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

    2006-10-31

    The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty,cycles (high burnup, boiling, aggressive chemistry) andto investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment.

  11. Evaluation of Fission Product Critical Experiments and Associated Biases for Burnup Credit Validation

    SciTech Connect

    Mueller, Don; Rearden, Bradley T; Reed, Davis Allan

    2010-01-01

    One of the challenges associated with implementation of burnup credit is the validation of criticality calculations used in the safety evaluation; in particular the availability and use of applicable critical experiment data. The purpose of the validation is to quantify the relationship between reality and calculated results. Validation and determination of bias and bias uncertainty require the identification of sets of critical experiments that are similar to the criticality safety models. A principal challenge for crediting fission products (FP) in a burnup credit safety evaluation is the limited availability of relevant FP critical experiments for bias and bias uncertainty determination. This paper provides an evaluation of the available critical experiments that include FPs, along with bounding, burnup-dependent estimates of FP biases generated by combining energy dependent sensitivity data for a typical burnup credit application with the nuclear data uncertainty information distributed with SCALE 6. A method for determining separate bias and bias uncertainty values for individual FPs and illustrative results is presented. Finally, a FP bias calculation method based on data adjustment techniques and reactivity sensitivity coefficients calculated with the SCALE sensitivity/uncertainty tools and some typical results is presented. Using the methods described in this paper, the cross-section bias for a representative high-capacity spent fuel cask associated with the ENDF/B-VII nuclear data for 16 most important stable or near stable FPs is predicted to be no greater than 2% of the total worth of the 16 FPs, or less than 0.13 % k/k.

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

    DOEpatents

    Johnson, Gerald D.; Lobsinger, Ralph J.; Hamilton, Margaret L.; Gelles, David S.

    1990-01-01

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

  13. Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations

    SciTech Connect

    Fensin, M. L.; Galloway, J. D.; James, M. R.

    2015-04-11

    The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and new predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.

  14. Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations

    DOE PAGES

    Fensin, M. L.; Galloway, J. D.; James, M. R.

    2015-04-11

    The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and newmore » predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.« less

  15. Modeling depletion simulations for a high-burnup, highly heterogeneous BWR fuel assembly with scale

    SciTech Connect

    Smith, H. J.

    2012-07-01

    Extensive SCALE isotopic validation studies have been performed for various PWR fuel assembly designs and operating conditions, and to a lesser extent for BWR fuel assembly designs. However, no SCALE validation work has been documented for newer, highly heterogeneous BWR fuel assembly designs at high burnup. Isotopic benchmark calculations of the earlier, more geometrically uniform BWR fuel assemblies are less sensitive to simplification of the operating history details and certain modeling assumptions than heterogeneous fuel assemblies, particularly at high burnup. This analysis shows the capability of SCALE to simulate a complex highly heterogeneous SVEA96 Optima fuel assembly and illustrates the importance of the need for the highest possible accuracy and precision in isotope measurements intended to be used as benchmark-quality results. In addition, this analysis quantifies the impact of various modeling assumptions on the results. The sample for which the simulation results are reported here achieved a burnup 62 GWd/MTU and was analyzed as part of the MALIBU Extension program. (authors)

  16. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    SciTech Connect

    Ilas, Germina; Gauld, Ian C

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  17. Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel

    SciTech Connect

    Parks, C V; DeHart, M D; Wagner, John C

    2000-03-13

    This report has been prepared to review relevant background information and provide technical discussion that will help initiate a PIRT (Phenomena Identification and Ranking Tables) process for use of burnup credit in light-water reactor (LWR) spent fuel storage and transport cask applications. The PIRT process will be used by the NRC Office of Nuclear Regulatory Research to help prioritize and guide a coordinated program of research and as a means to obtain input/feedback from industry and other interested parties. The review and discussion in this report are based on knowledge and experience gained from work performed in the United States and other countries. Current regulatory practice and perceived industry needs are also reviewed as a background for prioritizing technical needs that will facilitate safe practice in the use of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation is given. Finally, phenomena that need to be better understood for effective licensing, together with technical issues that require resolution, are presented and discussed in the form of a prioritization ranking and initial draft program plan.

  18. Grain and burnup dependence of spent fuel oxidation: geological repository impact

    SciTech Connect

    Hanson, B. D.; Kansa, E. J.; Stoot, R.B.

    1998-10-15

    Further refinements to the oxidation model of Stout et al. have been made. The present model incorporates the burnup dependence of the oxidation rate in addition to an allowance for a distribution of grain sizes. The model was tested by comparing the model results with the oxidation histories of spent fuel samples oxidized in Thermogravimetric Analysis (TGA) or Oven Dry-Bath (ODB) experiments. The comparison between the experimental and model results are remarkably close and confirm the assumption that grain-size distributions and activation energies are the important parameters to predicting oxidation behavior. The burnup dependence of the activation energy was shown to have a greater effect than decreasing the effective grain size in suppressing the rate of the reaction U{sub 4}O{sub 9}(rightwards arrow)U{sub 3} O{sub 4}. Model results predict that U{sub 3}O{sub 8} formation of spent fuels exposed to oxygen will be suppressed even for high burnup fuels that have undergone restructuring in the rim region, provided the repository temperature is kept sufficient.

  19. Uncertainties in the effects of burnup and their impact on criticality safety licensing criteria

    SciTech Connect

    Carlson, R.W.; Fisher, L.E.

    1990-07-13

    Current criteria for criticality safety for spent fuel shipping and storage casks are conservative because no credit is permitted for the effects of burnup of the fuel inside the cask. Cask designs that will transport and store large numbers of fuel assemblies (20 or more) must devote a substantial part of their payload to criticality control measures if they are to meet this criteria. The Department of Energy is developing the data necessary to support safety analyses that incorporate the effects of burnup for the next generation of spent fuel shipping casks. The efforts described here are devoted to the development of acceptance criteria that will be the basis for accepting safety analyses. Preliminary estimates of the uncertainties of the effects of burnup have been developed to provide a basis for the consideration of critically safety criteria. The criticality safety margins in a spent fuel shipping or storage cask are dominated by the portions of a fuel assembly that are in low power regions of a reactor core, and the reactor operating conditions are very different from spent fuel storage or transport cask conditions. Consequently, the experience that has been gathered during years of reactor operation does not apply directly to the prediction of criticality safety margins for spent fuel shipping or storage casks. The preliminary estimates of the uncertainties presented in this paper must be refined by both analytical and empirical studies that address both the magnitude of the uncertainties and their interdependence. 9 refs., 5 figs.

  20. Preliminary Study of Burnup Characteristics for a Simplified Small Pebble Bed Reactor

    SciTech Connect

    Irwanto, Dwi; Kato, Yukikata; Obara, Toru; Yamanaka, Ichiro

    2010-06-22

    Simplification of the pebble bed reactor by removing the unloading device from the system was peformed. For this reactor design, a suitable fuel-loading scheme is the Peu a Peu (little by little) fueling scheme. In the Peu a Peu modus, there is no unloading device; as such, the fuels are never discharged and remain at the bottom of the core during reactor operation. This means that the burnup cycle and reactivity is controlled by the addition of fuel. The objectives of the the present study were to find a means of carrying out the exact calculations needed to analyze the Peu a Peu fuel-loading scheme and to optimize the fuel composition, and fuel-loading scheme to achieve better burnup characteristics. The Monte Carlo method is used to perform calculations with high accuracy. Before the calculation of the whole core, the analysis for the infinite geometry was performed. The power generated per mass consumed for each combination of the uranium enrichment and packing fraction was analyzed from the parametric survey. By using the optimal value obtained, a whole-core calculation for the small 20 MWth reactor was performed and the criticality and burnup of this design was analyzed.

  1. Quantitative study of the effect of cladding thickness on modal confinement loss in photonic waveguides.

    PubMed

    Jiang, Shidong; Lai, Jun

    2016-10-31

    There has been increasing interest in making photonic devices more and more compact in the integrated photonics industry, and one of the important questions for manufacturers and design engineers is how to quantify the effect of the finite cladding thickness on the modal confinement loss of photonic waveguides. This requires at least six to seven digits of accuracy for the computation of propagation constant β since the modal confinement loss is proportional to the imaginary part of β that is six to seven orders of magnitude smaller than its real part by the industrial standard. In this paper, we present an accurate and efficient method to compute the propagation constant of the electromagnetic modes of photonic waveguides with an arbitrary number of (nonsmooth) inclusions in a layered media. The method combines a well-conditioned boundary integral equation formulation for photonic waveguides which requires the discretization of the material interface only, and efficient Sommerfeld integral representations to treat the effect of the layered medium. Our scheme is capable of calculating the propagation loss of the electromagnetic modes with high fidelity, even for waveguides with corners embedded in a cladding material of finite thickness. The numerical results, with a more than 10-digit accuracy, show quantitatively that the modal confinement loss of the rectangular waveguide increases exponentially fast as the cladding thickness decreases.

  2. Seamless metal-clad fiber-reinforced organic matrix composite structures and process for their manufacture

    NASA Technical Reports Server (NTRS)

    Bluck, Raymond M. (Inventor); Bush, Harold G. (Inventor); Johnson, Robert R. (Inventor)

    1990-01-01

    A metallic outer sleeve is provided which is capable of enveloping a hollow metallic inner member having continuous reinforcing fibers attached to the distal end thereof. The inner member is then introduced into outer sleeve until inner member is completely enveloped by outer sleeve. A liquid matrix member is then injected into space between inner member and outer sleeve. A pressurized heat transfer medium is flowed through the inside of inner member, thereby forming a fiber reinforced matrix composite material. The wall thicknesses of both inner member and outer sleeve are then reduced to the appropriate size by chemical etching, to adjust the thermal expansion coefficient of the metal-clad composite structure to the desired value. thereby forming a fiber reinforced matrix composite material. The wall thicknesses of both inner member and outer sleeve are then reduced to the appropriate size by chemical etching, to adjust the thermal expansion coefficient of the metal-clad composite structure to the desired value. The novelty of this invention resides in the development of a efficient method of producing seamless metal clad fiber reinforced organic matrix composite structures.

  3. Intrinsic optical fiber sensor for sensing organophosphate nerve agent using the modified cladding approach

    NASA Astrophysics Data System (ADS)

    Bansal, Lalitkumar; El-Sherif, Mahmoud

    2004-03-01

    The concept of modified cladding based sensors represents the largest class of intrinsic fiber optic chemical sensors. In this design, the passive cladding of the optical fiber is replaced by an active coating, called modified cladding. The analyte in this case diffuses into the coating and induces changes in the absorbance, fluorescence, or some other spectroscopic property of the modified cladding, the coating acts as a chemo-chromic transducer and sensing takes place by intensity modulation. This design i.e. of the coating based sensors, has found enormous applicability in the realm of chemical and biochemical sensing which also includes environmental monitoring and detection of chemical warfare agents. In this paper, the development of an intrinsic fiber optic sensor for detection of organophosphate dimethyl-methyl phoshopnate (DMMP) is presented. DMMP is a chemical precursor to the nerve agent sarin. The chemo-chromic transducer material used as a modified coating on the fiber core is NDSA (Naphthalene disulphonic acid) doped polypyrrole. This coating material shows conductivity and absorbance change when exposed to DMMP. The fabrication of the sensor device is a three step process which involves (a) etching a small section of the optical fiber to expose the core, (b) coating the etched section of the optical fiber with the polymer, (c) integration of sensor components and testing. Thin film characterization is done using the UV-Vis spectrophotometer on in-situ coated films of polypyrrole on a glass substrate to check for absorbance change upon exposure to DMMP. The development procedure is presented next and encouraging results are discussed.

  4. Hollow core anti-resonant fiber with split cladding.

    PubMed

    Huang, Xiaosheng; Qi, Wenliang; Ho, Daryl; Yong, Ken-Tye; Luan, Feng; Yoo, Seongwoo

    2016-04-04

    An improved design for hollow core anti-resonant fibers (HAFs) is presented. A split cladding structure is introduced to reduce the fabrication distortion within design tolerance. We use numerical simulations to compare the Kagome fibers (KFs) and the proposed split cladding fibers (SCFs) over two normalized transmission bands. It reveals that SCFs are able to maintain the desired round shape of silica cladding walls, hence improving the confinement loss (CL) compared to the KF and is comparable to that of the nested antiresonant nodeless fiber (NANF) with the same core size. In addition, the SCF allows stacking multiple layers of cladding rings to control the CL. The influences of the number of cladding layers and the cladding gap width on the CL of the SCFs have been studied. SCF with three cladding rings is fabricated by the stack-and-draw technique. A measured attenuation spectrum matches well with the calculation prediction. The measured near field mode patterns also prove the feasibility of our fiber design.

  5. Scanning Optics Enabled Possibilities and Challenges in Laser Cladding

    NASA Astrophysics Data System (ADS)

    Pekkarinen, Joonas

    Laser cladding using a scanned beam is quite a similar process than laser cladding using static optics (e.g. lens or mirror optics). The main difference comes from the manipulation of the laser beam. In laser cladding with scanning optics the laser beam is manipulated with a scanner so that the laser's area of influence can be shaped numerically. This increases cladding process flexibility. Scanning optics enable laser beam modification considerably versatile way than normal static optics can. This is due to possibility of numerical adjustment of scanning amplitude, laser power and scanning frequency. By modifying these parameters clad beads geometry can by modified quite freely. However scanned laser beam in surface modification process creates some restricting factors to the process. Mainly limitations for the process parameter values come from the dual characteristics of the energy input. This paper treats usability of scanning optics in laser cladding process in general level. In this paper is discussed how scanned beam can be used to increase the flexibility but also maters that limit the usage of scanned beam in cladding process. Process possibilities and limitations are presented in trough experimental data and examples.

  6. Interfacial Characterization of Dissimilar Joints Between Al/Mg/Al-Trilayered Clad Sheet to High-Strength Low-Alloy Steel

    NASA Astrophysics Data System (ADS)

    Macwan, A.; Jiang, X. Q.; Chen, D. L.

    2015-07-01

    Magnesium (Mg) alloys are increasingly used in the automotive and aerospace sectors to reduce vehicle weight. Al/Mg/Al tri-layered clad sheets are deemed as a promising alternative to improve the corrosion resistance and formability of Mg alloys. The structural application of Al/Mg/Al tri-layered clad sheets inevitably involves welding and joining in the multi-material vehicle body manufacturing. This study aimed to characterize the bonding interface microstructure of the Al/Mg/Al-clad sheet to high-strength low-alloy steel with and without Zn coating using ultrasonic spot welding at different levels of welding energy. It was observed that the presence of Zn coating improved the bonding at the interface due to the formation of Al-Zn eutectic structure via enhanced diffusion. At a higher level of welding energy, characteristic flow patterns of Zn into Al-clad layer were observed with an extensive penetration mainly along some high angle grain boundaries. The dissimilar joints without Zn coating made at a high welding energy of 800 J failed partially from the Al/Fe weld interface and partially from the Al/Mg clad interface, while the joints with Zn coating failed from the Al/Mg clad interface due to the presence of brittle Al12Mg17 phase.

  7. Silver-clad Bi-2223 processing

    NASA Astrophysics Data System (ADS)

    Lay, K. W.; Arendt, R. H.; Tkaczyk, J. E.; Garbauskas, M. F.

    1992-04-01

    One of the most promising techniques for the production of long lengths of high Tc Superconductors utilizes themechanical deformation and subsequent heat treatment of silver-clad bismuth cuprates. Some recent experiments on Bi 1.7Pb0.3Sr2Ca2+x Cu3+xOy, with x=0.2, 0.4, and 0.75, processed by packing powder ina silver tube followed by swaging, drawing, and rolling are described. The effect of the initial powder composition, both cation ratios and phase composition, on the critical current of the composite tapes are explored. Different final deformation methods and heat treatment schedules also are discussed. A set of guidelines for the production of high Jc 2223 tapes is given.

  8. Fabrication of oxide dispersion strengthened ferritic clad fuel pins

    SciTech Connect

    Zirker, L.R. ); Bottcher, J.H. ); Shikakura, S. ); Tsai, C.L. . Dept. of Welding Engineering); Hamilton, M.L. )

    1991-01-01

    A resistance butt welding procedure was developed and qualified for joining ferritic fuel pin cladding to end caps. The cladding are INCO MA957 and PNC ODS lots 63DSA and 1DK1, ferritic stainless steels strengthened by oxide dispersion, while the end caps are HT9 a martensitic stainless steel. With adequate parameter control the weld is formed without a residual melt phase and its strength approaches that of the cladding. This welding process required a new design for fuel pin end cap and weld joint. Summaries of the development, characterization, and fabrication processes are given for these fuel pins. 13 refs., 6 figs., 1 tab.

  9. Oxidation performance of platinum-clad Mo-47Re alloy

    NASA Technical Reports Server (NTRS)

    Clark, Ronald K.; Wallace, Terryl A.

    1994-01-01

    The alloy Mo-47Re has favorable mechanical properties at temperatures above 1400 C, but it undergoes severe oxidation when used in air with no protective coating. To shield the alloy from oxidation, platinum cladding has been evaluated. The unprotected alloy undergoes catastrophic oxidation under static and dynamic oxidation conditions. The platinum cladding provides good protection from static and dynamic oxidation for moderate times at 1260 C. Samples tested for longer times under static oxidation conditions experienced severe oxidation. The data suggest that oxidation results from the transport of oxygen through the grain boundaries and through the pinhole defects of the platinum cladding.

  10. Linear inner cladding fiber amplifier suppressing mode instability

    NASA Astrophysics Data System (ADS)

    Li, Zebiao; Huang, Zhihua; Lin, Honghuan; Xu, Shanhui; Yang, Zhongmin; Wang, Jianjun; Jing, Feng

    2016-11-01

    We use a semi-analytical model considering pump power saturation in high power fiber laser systems of multi-kW-class to calculate mode instability threshold. A novel designed fiber, linear inner-cladding fiber, can mitigate mode instability effect by decreasing nonlinear coupling coefficient and smoothing heat profile along the fiber. We investigate strong pump absorption of linear inner-cladding fiber, leading to shorter fiber length. With 915 nm pumping, linear inner-cladding fiber can reach 10 kW output power without mode instability in theory.

  11. Inhibitors, cladded trees protect sour gas wells in Abu Dhabi

    SciTech Connect

    Morsi, K.M. )

    1994-06-13

    Continuous chemical inhibition has prevented corrosion downhole, and tests indicate that Inconel 625 cladding will protect the christmas trees on wells producing sour gas from the Thamama C reservoir. Metallic corrosion is a costly problem. Estimates indicate that corrosion costs the oil industry several billion dollars per year. In addition, oil companies spend over $100 million/year on corrosion inhibitors for combating downhole tubular and casing corrosion. Abu Dhabi National Oil Co. (Adnoc) has successfully completed wells in extremely harsh operating conditions with high temperatures, pressures, and high concentrations of H[sub 2]S, CO[sub 2], and brine. Such environments require special materials for downhole and surface equipment. The Thamama C reservoir, in an onshore gas field, produces gas containing H[sub 2]S and CO[sub 2] in the range of 0.7--8.0 mole % and 4.0--8.0 mole %, respectively. The Thamama C gas-gathering system comprises 19 wells connected to four trunk lines that transport produced gas and associated condensate to a central processing plant. The paper discusses material and inhibitor selection.

  12. Boron-copper neutron absorbing material and method of preparation

    DOEpatents

    Wiencek, Thomas C.; Domagala, Robert F.; Thresh, Henry

    1991-01-01

    A composite, copper clad neutron absorbing material is comprised of copper powder and boron powder enriched with boron 10. The boron 10 content can reach over 30 percent by volume, permitting a very high level of neutron absorption. The copper clad product is also capable of being reduced to a thickness of 0.05 to 0.06 inches and curved to a radius of 2 to 3 inches, and can resist temperatures of 900.degree. C. A method of preparing the material includes the steps of compacting a boron-copper powder mixture and placing it in a copper cladding, restraining the clad assembly in a steel frame while it is hot rolled at 900.degree. C. with cross rolling, and removing the steel frame and further rolling the clad assembly at 650.degree. C. An additional sheet of copper can be soldered onto the clad assembly so that the finished sheet can be cold formed into curved shapes.

  13. Microstructural characteristics and embrittlement phenomena in neutron irradiated 309L stainless steel RPV clad

    NASA Astrophysics Data System (ADS)

    Lee, J. S.; Kim, I. S.; Kasada, R.; Kimura, A.

    2004-03-01

    The effects of neutron irradiation on the microstructural features and mechanical properties of 309L stainless steel RPV clad were investigated using TEM, SEM, small tensile, microhardness and small punch (SP) tests. The neutron irradiations were performed at 290 °C up to the fluences of 5.1 × 10 18 and 1.02 × 10 19 n/cm 2 (>1 MeV) in Japan Materials Testing Reactor (JMTR). The microstructure of the clad before and after irradiation was composed of main part of fcc austenite, a few percent of bcc δ-ferrite and small amount of brittle σ phase. After irradiation, not only the yield stress and microhardness, but SP ductile to brittle transition temperature (SP-DBTT) were increased. However, the increase in SP-DBTT is almost saturated, independent of the neutron fluence. Based on the TEM observation, the origin of irradiation hardening was accounted for by the irradiation-produced defect clusters of invisible fine size (<1-2 nm), and the shift of SP-DBTT was primary due to the higher hardening and the preferential failure of δ-ferrite. The embrittlement of the clad was strongly affected by the initial microstructural factors, such as the amount of brittle σ phase, which caused a cracking even in an early stage of deformation.

  14. Laser cladding of tungsten carbides (Spherotene ®) hardfacing alloys for the mining and mineral industry

    NASA Astrophysics Data System (ADS)

    Amado, J. M.; Tobar, M. J.; Alvarez, J. C.; Lamas, J.; Yáñez, A.

    2009-03-01

    The abrasive nature of the mechanical processes involved in mining and mineral industry often causes significant wear to the associated equipment and derives non-negligible economic costs. One of the possible strategies to improve the wear resistance of the various components is the deposition of hardfacing layers on the bulk parts. The use of high power lasers for hardfacing (laser cladding) has attracted a great attention in the last decade as an alternative to other more standard methods (arc welding, oxy-fuel gas welding, thermal spraying). In laser cladding the hardfacing material is used in powder form. For high hardness applications Ni-, Co- or Fe-based alloys containing hard phase carbides at different ratios are commonly used. Tungsten carbides (WC) can provide coating hardness well above 1000 HV (Vickers). In this respect, commercially available WC powders normally contain spherical micro-particles consisting of crushed WC agglomerates. Some years ago, Spherotene ® powders consisting of spherical-fused monocrystaline WC particles, being extremely hard, between 1800 and 3000 HV, were patented. Very recently, mixtures of Ni-based alloy with Spherotene powders optimized for laser processing were presented (Technolase ®). These mixtures have been used in our study. Laser cladding tests with these powders were performed on low carbon steel (C25) substrates, and results in terms of microstructure and hardness will be discussed.

  15. Surface modification of AISI H13 tool steel by laser cladding with NiTi powder

    NASA Astrophysics Data System (ADS)

    Norhafzan, B.; Aqida, S. N.; Chikarakara, E.; Brabazon, D.

    2016-04-01

    This paper presents laser cladding of NiTi powder on AISI H13 tool steel surface for surface properties enhancement. The cladding process was conducted using Rofin DC-015 diffusion-cooled CO2 laser system with wavelength of 10.6 µm. NiTi powder was pre-placed on H13 tool steel surface. The laser beam was focused with a spot size of 90 µm on the sample surface. Laser parameters were set to 1515 and 1138 W peak power, 18 and 24 % duty cycle and 2300-3500 Hz laser pulse repetition frequency. Hardness properties of the modified layer were characterized by Wilson Hardness tester. Metallographic study and chemical composition were conducted using field emission scanning electron microscope and energy-dispersive X-ray spectrometer (EDXS) analysis. Results showed that hardness of NiTi clad layer increased three times that of the substrate material. The EDXS analysis detected NiTi phase presence in the modified layer up to 9.8 wt%. The metallographic study shows high metallurgical bonding between substrate and modified layer. These findings are significant to both increased hardness and erosion resistance of high-wear-resistant components and elongating their lifetime.

  16. Conservation of Stone Cladding on the FAÇADE of Royal Palace in Caserta

    NASA Astrophysics Data System (ADS)

    Titomanlio, I.

    2013-07-01

    The beauty of cultural heritage and monumental architecture, is often linked to their non-structural elements and decorative stones façades cladding. The collapse of these elements causes significant consequences that interest the social, the economic, the historical and the technical fields. Several regulatory documents and literature studies contain methods to address the question of relief and of the risk analysis and due to the non - structural stones security. Among the references are widespread international regulatory documents prepared by the Federal Emergency Management Agency of the United States by Applied Technology Council and California. In Italy there are some indications contained in the Norme Tecniche per le Costruzioni and the Direttiva del Presidente del Consiglio dei Ministri in 2007, finalize to the reduction of seismic risk assessment of cultural heritage. The paper, using normative references and scientific researches, allows to analyze on Royal Palace of Caserta the safety and the preservation of cultural heritage and the vulnerability of non-structural stones façade cladding. Using sophisticated equipments of Laboratory ARS of the Second University of Naples, it was possible to analyze the collapse of stone elements due to degradation caused by natural phenomena of deterioration (age of the building, type of materials, geometries , mode of fixing of the elements themselves). The paper explains the collapse mechanisms of stones façade cladding of Luigi Vanvitelli Palace.

  17. Initial and Long-Term Movement of Cladding Installed Over Exterior Rigid Insulation

    SciTech Connect

    Baker, P.

    2014-09-01

    Changes in the International Energy Conservation Code (IECC) from 2009 to 2012 have resulted in the use of exterior rigid insulation becoming part of the prescriptive code requirements. With more jurisdictions adopting the 2012 IECC builders are going to finding themselves required to incorporate exterior insulation in the construction of their exterior wall assemblies. For thick layers of exterior insulation (levels greater than 1.5 inches), the use 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. However, there has been a significant resistance to its widespread implementation due to a lack of research and understanding of the mechanisms involved and potential creep effects of the assembly under the sustained dead load of a cladding. This research was an extension on previous research conducted by BSC in 2011, and 2012. Each year the understanding of the system discrete load component interactions, as well as impacts of environmental loading has increased. The focus of the research was to examine more closely the impacts of screw fastener bending on the total system capacity, effects of thermal expansion and contraction of materials on the compressive forces in the assembly, as well as to analyze a full years worth of cladding movement data from assemblies constructed in an exposed outdoor environment.

  18. Results of irradiation of (U0.55Pu0.45)N and (U0.4Pu0.6)N fuels in BOR-60 up to ˜12 at.% burn-up

    NASA Astrophysics Data System (ADS)

    Rogozkin, B. D.; Stepennova, N. M.; Fedorov, Yu. Ye.; Shishkov, M. G.; Kryukov, F. N.; Kuzmin, S. V.; Nikitin, O. N.; Belyaeva, A. V.; Zabudko, L. M.

    2013-09-01

    In the article presented are the results of post-irradiation tests of helium bonded fuel pins with mixed mononitride fuel (U0.55Pu0.45)N and (U0.4Pu0.6)N having 85% density irradiated in BOR-60 reactor. Achieved maximum burn-up was, respectively, equal to 9.4 and 12.1 at.% with max linear heat rates 41.9 and 54.5 kW/m. Maximum irradiation dose was 43 dpa. No damage of claddings made of ChS-68 steel (20% cold worked) was observed, and ductility margin existed. Maximum depth of cladding corrosion was within 15 μm. Swelling rates of (U0.4Pu0.6)N and (U0.55Pu0.45)N were, respectively, ˜1.1% and ˜0.68% per 1 at.%. Gas release rate did not exceed 19.3% and 19%. Pattern of porosity distribution in the fuel influenced fuel swelling and gas release rates. Plutonium and uranium are uniformly distributed in the fuel, local minimum values of their content being caused by pores and cracks in the pellets. The observable peaks in content distribution are probably connected with the local formation of isolated phases (e.g. Mo, Pd) while the minimum values refer to fuel pores and cracks. Xenon and cesium tend to migrate from the hot sections of fuel, and therefore their min content is observed in the central section of the fuel pellets. Phase composition of the fuel was determined with X-ray diffractometer. The X-ray patterns of metallographic specimens were obtained by the scanning method (the step was 0.02°, the step exposition was equal to 2 s). From the X-ray diffraction analysis data, it follows that the nitrides of both fuel types have the single-phase structure with an FCC lattice (see Table 6).

  19. Neutron-induced helium implantation in GCFR cladding

    SciTech Connect

    Yamada, H.; Poeppel, R. B.; Sevy, R. H.

    1980-10-01

    The neutron-induced implantation of helium atoms on the exterior surfaces of the cladding of a prototypic gas-cooled fast reactor (GCFR) has been investigated analytically. A flux of recoil helium particles as high as 4.2 x 10/sup 10/ He/cm/sup 2/.s at the cladding surface has been calculated at the peak power location in the core of a 300-MWe GCFR. The calculated profile of the helium implantation rates indicates that although some helium is implanted as deep as 20 ..mu..m, more than 99% of helium particles are implanted in the first 2-..mu..m-deep layer below the cladding surface. Therefore, the implanted helium particles should mainly affect surface properties of the GCFR cladding.

  20. Composite polymer-glass edge cladding for laser disks

    DOEpatents

    Powell, Howard T.; Riley, Michael O.; Wolfe, Charles R.; Lyon, Richard E.; Campbell, John H.; Jessop, Edward S.; Murray, James E.

    1989-01-01

    Large neodymium glass laser disks for disk amplifiers such as those used in the Nova laser require an edge cladding which absorbs at 1 micrometer. This cladding prevents edge reflections from causing parasitic oscillations which would otherwise deplete the gain. Nova now utilizes volume-absorbing monolithic-glass claddings which are fused at high temperature to the disks. These perform quite well but are expensive to produce. Absorbing glass strips are adhesively bonded to the edges of polygonal disks using a bonding agent whose index of refraction matches that of both the laser and absorbing glass. Optical finishing occurs after the strips are attached. Laser disks constructed with such claddings have shown identical gain performance to the previous Nova disks and have been tested for hundreds of shots without significant degradation.

  1. Composite polymer: Glass edge cladding for laser disks

    DOEpatents

    Powell, H.T.; Wolfe, C.A.; Campbell, J.H.; Murray, J.E.; Riley, M.O.; Lyon, R.E.; Jessop, E.S.

    1987-11-02

    Large neodymium glass laser disks for disk amplifiers such as those used in the Nova laser require an edge cladding which absorbs at 1 micrometer. This cladding prevents edge reflections from causing parasitic oscillations which would otherwise deplete the gain. Nova now utilizes volume-absorbing monolithic-glass claddings which are fused at high temperature to the disks. These perform quite well but are expensive to produce. Absorbing glass strips are adhesively bonded to the edges of polygonal disks using a bonding agent whose index of refraction matches that of both the laser and absorbing glass. Optical finishing occurs after the strips are attached. Laser disks constructed with such claddings have shown identical gain performance to the previous Nova disks and have been tested for hundreds of shots without significant degradation. 18 figs.

  2. View of building 11050, showing metal clad addition on east ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    View of building 11050, showing metal clad addition on east elevation, looking southwest. - Naval Ordnance Test Station Inyokern, China Lake Pilot Plant, Machine Shop, C Street, China Lake, Kern County, CA

  3. Jet slurry erosion performance of composite clad and its characterization

    NASA Astrophysics Data System (ADS)

    B, Lohit R.; Horakeri, Gururaj S.; Bhovi, Prabakhar M.

    2016-09-01

    In the present work, development of composite cladding consists of Cr23C6 (chromium carbide) as reinforcement particles 20 wt. % in Ni-based matrix 80 wt. % on austenitic stainless steel through exposure of microwave radiation has been carried out. The jet slurry erosion test was performed on microwave composite clad. The functional performance of composite clad has been evaluated for different parametric conditions like varying impingement velocity and impact angle. The increasing weight loss trend was observed with time for the first 30 min. after that the individual trend decreased; at high impingement velocity and maximum impact angle. SEM micrographs of eroded clad samples at various impact angle and impingement velocity were discussed. The maximum weight loss occurred at 90° angle and velocity of 60 m/s, and minimum at 30° angle and velocity of 20 m/s.

  4. Young's modulus anisotropy in reactor pressure vessel cladding

    NASA Astrophysics Data System (ADS)

    Vandermeulen, W.; Mertens, M.; Scibetta, M.

    2012-02-01

    In a previous study it was shown that the anisotropy of Young's modulus in the stainless steel cladding of a reactor pressure vessel could be attributed to the solidification texture of the cladding. Further it was found that annealing the samples to remove the delta phase caused a modulus change but only in some directions. Since the texture was only estimated from X-ray diffraction patterns the moduli, calculated for some principal directions, differed considerably from the measured ones. In the present study, executed on a practically identical cladding, the texture was determined by actual texture measurements. It was found to be close to a fibre texture with <0 0 1> perpendicular to the cladding plane and the values calculated from it agreed much better with the experimental ones. The annealing effect found in the previous study was shown to be due to surface recrystallization induced by milling damage.

  5. Double-clad nuclear-fuel safety rod

    DOEpatents

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

    1981-12-30

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

  6. Cladding for transverse-pumped solid-state laser

    NASA Technical Reports Server (NTRS)

    Byer, Robert L. (Inventor); Fan, Tso Y. (Inventor)

    1989-01-01

    In a transverse pumped, solid state laser, a nonabsorptive cladding surrounds a gain medium. A single tranverse mode, namely the Transverse Electromagnetic (TEM) sub 00 mode, is provided. The TEM sub 00 model has a cross sectional diameter greater than a transverse dimension of the gain medium but less than a transverse dimension of the cladding. The required size of the gain medium is minimized while a threshold for laser output is lowered.

  7. Cladding-pumped erbium-doped multicore fiber amplifier.

    PubMed

    Abedin, K S; Taunay, T F; Fishteyn, M; DiGiovanni, D J; Supradeepa, V R; Fini, J M; Yan, M F; Zhu, B; Monberg, E M; Dimarcello, F V

    2012-08-27

    A cladding pumped multicore erbium-doped fiber amplifier for simultaneous amplification of 6 channels is demonstrated. Peak gain over 32 dB has been obtained at a wavelength of 1560 nm and the bandwidth measured at 20-dB gain was about 35 nm. Numerical modeling of cladding pumped multicore erbium-doped amplifier was also performed to study the properties of the amplifier. The results of experiment and simulation are found to be in good agreement.

  8. Review of cladding-coolant interactions during LWR accident transients

    SciTech Connect

    Hobson, D.O.

    1980-01-01

    Some of the coolant-cladding interactions that can take place during the design basis loss-of-coolant accident and the Three Mile Island loss-of-coolant accident are analyzed. The physical manifestations of the interactions are quite similar, but the time sequences involved can cause very different end results. These results are described and a listing is given of the main research programs that are involved in coolant-cladding interaction research.

  9. Hollow cylindrical plasma filament waveguide with discontinuous finite thickness cladding

    SciTech Connect

    Alshershby, Mostafa; Hao Zuoqiang; Lin Jingquan

    2013-01-15

    We have explored here a hollow cylindrical laser plasma multifilament waveguide with discontinuous finite thickness cladding, in which the separation between individual filaments is in the range of several millimeters and the waveguide cladding thickness is in the order of the microwave penetration depth. Such parameters give a closer representation of a realistic laser filament waveguide sustained by a long stable propagation of femtosecond (fs) laser pulses. We report how the waveguide losses depend on structural parameters like normalized plasma filament spacing, filament to filament distance or pitch, normal spatial frequency, and radius of the plasma filament. We found that for typical plasma parameters, the proposed waveguide can support guided modes of microwaves in extremely high frequency even with a cladding consisting of only one ring of plasma filaments. The loss of the microwave radiation is mainly caused by tunneling through the discontinuous finite cladding, i.e., confinement loss, and is weakly dependent on the plasma absorption. In addition, the analysis indicates that the propagation loss is fairly large compared with the loss of a plasma waveguide with a continuous infinite thickness cladding, while they are comparable when using a cladding contains more than one ring. Compared to free space propagation, this waveguide still presents a superior microwave transmission to some distance in the order of the filamentation length; thus, the laser plasma filaments waveguide may be a potential channel for transporting pulsed-modulated microwaves if ensuring a long and stable propagation of fs laser pulses.

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

    SciTech Connect

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

    2014-12-15

    The results of reactor tests of a carbonitride fuel in a single-crystal cladding from a molybdenum-based alloy can be used in substantiating the operational reliability of fuels in developing a project of a megawatt space nuclear power plant. The results of experimental studies of uranium and cesium penetration into the single-crystal cladding of fuel elements with a carbonitride fuel are interpreted. Those fuel elements passed nuclear power tests in the Ya-82 pilot plant for 8300 h at a temperature of about 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.

  11. Method and apparatus for measuring reactivity of fissile material

    DOEpatents

    Lee, D.M.; Lindquist, L.O.

    1982-09-07

    Given are a method and apparatus for measuring nondestructively and noninvasively (i.e., using no internal probing) the burnup, reactivity, or fissile content of any material which emits neutrons and which has fissionable components. The assay is accomplished by altering the return flux of neutrons into the fuel assembly by means of changing the reflecting material. The existing passive neutron emissions in the material being assayed are used as the source of interrogating neutrons. Two measurements of either emitted neutron or emitted gamma-ray count rates are made and are then correlated to either reactivity, burnup, or fissionable content of the material being assayed, thus providing a measurement of either reactivity, burnup, or fissionable content of the material being assayed. Spent fuel which has been freshly discharged from a reactor can be assayed using this method and apparatus. Precisions of 1000 MWd/tU appear to be feasible.

  12. Low Burnup Inert Matrix Fuels Performance: TRANSURANUS Analysis of the Halden IFA-652 First Irradiation Cycle

    SciTech Connect

    Calabrese, R.; Vettraino, F.; Tverberg, T.

    2006-07-01

    The inert matrix fuels are a promising option to reduce-eliminate worldwide plutonium stockpiles by burning it in LWRs. These fuels, where plutonium is hosted in a U-free inert matrix phase, may reach high burning efficiency while preventing new plutonium build-up under irradiation. A specific investigation on CSZ and thoria inert matrices has been developed by ENEA since several years. In-pile testing on the ENEA-conceived innovative fuels is ongoing in the OECD Halden HBWR since June 2000 (IFA-652 experiment). The registered burnup at the end of 2005 is about 38 MWd.kgU{sub eq}{sup -1} vs. 45 MWd.kgU{sub eq}{sup -1} (40 MWd.kgUOX{sub eq}{sup -1}) target. Fuel pins are equipped with fuel temperature thermocouples, internal pressure transducers and fuel stack elongation sensors, with the task of studying thermal conductivity and its degradation with burnup, densification-swelling behaviour and the FGR. In this paper, the response at low burnup (< 7 MWd.kgU{sub eq}{sup -1}) of CSZ-based fuels loaded in IFA-652, is analysed by means of the TRANSURANUS code. To this purpose, a comprehensive modelling of the above mentioned un-irradiated fuels, mainly relying on the thermophysical characterisation performed at the JRC/ITU-Karlsruhe, has been implemented in a custom TRANSURANUS version (TU-IMF). A comparison of the code predictions vs. the experimental data, aimed at evaluating the early-stage under irradiation phenomena, particularly densification and relocation, has been performed. (authors)

  13. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    SciTech Connect

    Wagner, J. C.

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

  14. An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

    SciTech Connect

    Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; Sterbentz, James W.

    2014-09-03

    AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methods for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma spectrometry

  15. An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

    DOE PAGES

    Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; ...

    2014-09-03

    AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methodsmore » for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma

  16. Development of Cladding Materials for Evacuated Panel Superinsulation

    SciTech Connect

    Wilkes, K.E.; Weaver, F.J.; Cumberbatch, G.M.; Begnoche, B.; Brodie, V.; Lamb, W.; Reitz, R.; Caldwell, P.; Meyer, C.

    1999-11-01

    This Cooperative Research and Development Agreement (CRADA) was among E.I. DuPont de Nemours and Company, VacuPanel, Inc., and Lockheed Martin Energy Research Corp. Evacuated panel superinsulations have thermal resistivities (R) substantially above that of conventional existing insulation without the environmental problems of some insulations such as foam insulations blown with Chlorofluorocarbons (CFCs) or hydrochlorofluorocarbons (HCFCs).

  17. Laser Cladding of Composite Bioceramic Coatings on Titanium Alloy

    NASA Astrophysics Data System (ADS)

    Xu, Xiang; Han, Jiege; Wang, Chunming; Huang, Anguo

    2016-02-01

    In this study, silicon nitride (Si3N4) and calcium phosphate tribasic (TCP) composite bioceramic coatings were fabricated on a Ti6Al4V (TC4) alloy using Nd:YAG pulsed laser, CO2 CW laser, and Semiconductor CW laser. The surface morphology, cross-sectional microstructure, mechanical properties, and biological behavior were carefully investigated. These investigations were conducted employing scanning electron microscope, energy-dispersive x-ray spectroscopy, and other methodologies. The results showed that both Si3N4 and Si3N4/TCP composite coatings were able to form a compact bonding interface between the coating and the substrate by using appropriate laser parameters. The coating layers were dense, demonstrating a good surface appearance. The bioceramic coatings produced by laser cladding have good mechanical properties. Compared with that of the bulk material, microhardness of composite ceramic coatings on the surface significantly increased. In addition, good biological activity could be obtained by adding TCP into the composite coating.

  18. Advanced Fuel Cycle Initiative - Projected Linear Heat Generation Rate and Burnup Calculations

    SciTech Connect

    Richard G. Ambrosek; Gray S. Chang; Debbie J. Utterbeck

    2005-02-01

    This report provides documentation of the physics analysis performed to determine the linear heat generation rate (LHGR) and burnup calculations for the Advanced Fuel Cycle Initiative (AFCI) tests, AFC-1D, AFC-1H, and AFC-1G. The AFC-1D and AFC-1H tests consists of low-fertile metallic fuel compositions and the AFC-1G test consists of non-fertile and low-fertile nitride compositions. These tests will be irradiated in the East Flux Trap (EFT) positions E1, E2, and E3, respectively, during Advanced Test Reactor (ATR) Cycle 135B.

  19. Antineutrinos for Reactor Safeguards: Effect of Fuel Loading and Burnup on the Signal

    NASA Astrophysics Data System (ADS)

    Erickson, Anna; Bernstein, Adam; Bowden, Nathaniel

    2014-02-01

    Various types of nuclear reactor related information, including relative power level and fuel evolution parameters, can be inferred remotely using antineutrino detectors. We show that it is possible to verify assembly-level burnup using information derived from an antineutrino detector if the nominal reactor fuel loading is known. Alternatively, if the core power is measured using an independent method, for example, a thermal hydraulic element, and the nominal core behavior is known, the antineutrino detector has a capability to determine previously unknown MOX loading in the core.

  20. Comparison of fiber lasers based on distributed side-coupled cladding-pumped fibers and double-cladding fibers.

    PubMed

    Huang, Zhihe; Cao, Jianqiu; Guo, Shaofeng; Chen, Jinbao; Xu, Xiaojun

    2014-04-01

    We compare both analytically and numerically the distributed side-coupled cladding-pumped (DSCCP) fiber lasers and double cladding fiber (DCF) lasers. We show that, through optimization of the coupling and absorbing coefficients, the optical-to-optical efficiency of DSCCP fiber lasers can be made as high as that of DCF lasers. At the same time, DSCCP fiber lasers are better than the DCF lasers in terms of thermal management.

  1. Bioactivity of fluorapatite/alumina composite coatings deposited on Ti6Al4V substrates by laser cladding

    NASA Astrophysics Data System (ADS)

    Chien, C. S.; Liu, C. W.; Kuo, T. Y.; Wu, C. C.; Hong, T. F.

    2016-04-01

    Hydroxyapatite (HA) is one of the most commonly used coating materials for metal implants. However, following high-temperature deposition, HA easily decomposes into an unstable phase or forms an amorphous phase, and hence, the long-term stability of the implant is reduced. Accordingly, the present study investigates the use of fluorapatite (FA) fortified with 20 wt% alumina (α-Al2O3) as an alternative biomedical coating material. The coatings are deposited on Ti6Al4V substrates using a Nd:YAG laser cladding process performed with laser powers and travel speeds of 400 W/200 mm/min, 800 W/400 mm/min and 1200 W/600 mm/min, respectively. The results show that for all of the specimens, a strong metallurgical bond is formed at the interface between the coating layer and the transition layer due to melting and diffusion. The XRD analysis results reveal that the cladding layers in all of the specimens consist mainly of FA, β-TCP, CaF2, Ti and θ-Al2O3 phases. In addition, the cladding layers of the specimens prepared using laser powers of 400 and 800 W also contain CaTiO3 and CaAl2O4, while that of the specimen clad using a power of 1200 W contains TTCP and CaO. Following immersion in simulated body fluid for 14 days, all of the specimens precipitate dense bone-like apatite and exhibit excellent bioactivity. However, among all of the specimens, the specimen that is prepared with a laser power of 800 W shows the best biological activity due to the presence of residual FA, apatite-generating CaTiO3 and a rough cladding layer surface.

  2. Online monitoring of thermo-cycles and its correlation with microstructure in laser cladding of nickel based super alloy

    NASA Astrophysics Data System (ADS)

    Muvvala, Gopinath; Patra Karmakar, Debapriya; Nath, Ashish Kumar

    2017-01-01

    Laser cladding, basically a weld deposition technique, is finding applications in many areas including surface coatings, refurbishment of worn out components and generation of functionally graded components owing to its various advantages over conventional methods like TIG, PTA etc. One of the essential requirements to adopt this technique in industrial manufacturing is to fulfil the increasing demand on product quality which could be controlled through online process monitoring and correlating the signals with the mechanical and metallurgical properties. Rapid thermo-cycle i.e. the fast heating and cooling rates involved in this process affect above properties of the deposited layer to a great extent. Therefore, the current study aims to monitor the thermo-cycles online, understand its variation with process parameters and its effect on different quality aspects of the clad layer, like microstructure, elemental segregations and mechanical properties. The effect of process parameters on clad track geometry is also studied which helps in their judicious selection to deposit a predefined thickness of coating. In this study Inconel 718, a nickel based super alloy is used as a clad material and AISI 304 austenitic steel as a substrate material. The thermo-cycles during the cladding process were recorded using a single spot monochromatic pyrometer. The heating and cooling rates were estimated from the recorded thermo-cycles and its effects on microstructures were characterised using SEM and XRD analyses. Slow thermo-cycles resulted in severe elemental segregations favouring Laves phase formation and increased γ matrix size which is found to be detrimental to the mechanical properties. Slow cooling also resulted in termination of epitaxial growth, forming equiaxed grains near the surface, which is not preferred for single crystal growth. Heat treatment is carried out and the effect of slow cooling and the increased γ matrix size on dissolution of segregated elements in

  3. Critical assessment of the pore size distribution in the rim region of high burnup UO2 fuels

    NASA Astrophysics Data System (ADS)

    Cappia, F.; Pizzocri, D.; Schubert, A.; Van Uffelen, P.; Paperini, G.; Pellottiero, D.; Macián-Juan, R.; Rondinella, V. V.

    2016-11-01

    A new methodology is introduced to analyse porosity data in the high burnup structure. Image analysis is coupled with the adaptive kernel density estimator to obtain a detailed characterisation of the pore size distribution, without a-priori assumption on the functional form of the distribution. Subsequently, stereological analysis is carried out. The method shows advantages compared to the classical approach based on the histogram in terms of detail in the description and accuracy within the experimental limits. Results are compared to the approximation of a log-normal distribution. In the investigated local burnup range (80-200 GWd/tHM), the agreement of the two approaches is satisfactory. From the obtained total pore density and mean pore diameter as a function of local burnup, pore coarsening is observed starting from ≈100 GWd/tHM, in agreement with a previous investigation.

  4. A semi-empirical model for the formation and depletion of the high burnup structure in UO2

    NASA Astrophysics Data System (ADS)

    Pizzocri, D.; Cappia, F.; Luzzi, L.; Pastore, G.; Rondinella, V. V.; Van Uffelen, P.

    2017-04-01

    In the rim zone of UO2 nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. For this purpose, we performed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Based on these new experimental data, we infer an exponential reduction of the average grain size with local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes.

  5. Cladding Attachment Over Thick Exterior Insulating Sheathing

    SciTech Connect

    Baker, P.; Eng, P.; Lepage, R.

    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?

  6. Cladding Attachment Over Thick Exterior Insulating Sheathing

    SciTech Connect

    Baker, P.; Eng, P.; Lepage, R.

    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?

  7. Weld overlay cladding with iron aluminides

    SciTech Connect

    Goodwin, G.M.

    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.

  8. Fabrication, properties, and tritium recovery from solid breeder materials

    SciTech Connect

    Johnson, C.E. ); Kondo, T. ); Roux, N. ); Tanaka, S. ); Vollath, D. )

    1991-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 133 refs., 1 fig.

  9. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

    DOE PAGES

    Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; ...

    2014-11-01

    This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application ofmore » the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.« less

  10. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    SciTech Connect

    Widiawati, Nina Su’ud, Zaki

    2015-09-30

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from −0.6695443 % at BOC to −0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

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

    SciTech Connect

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

    1993-01-01

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

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

    SciTech Connect

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

    1993-03-01

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

  13. A multi-platform linking code for fuel burnup and radiotoxicity analysis

    NASA Astrophysics Data System (ADS)

    Cunha, R.; Pereira, C.; Veloso, M. A. F.; Cardoso, F.; Costa, A. L.

    2014-02-01

    A linking code between ORIGEN2.1 and MCNP has been developed at the Departamento de Engenharia Nuclear/UFMG to calculate coupled neutronic/isotopic results for nuclear systems and to produce a large number of criticality, burnup and radiotoxicity results. In its previous version, it evaluated the isotopic composition evolution in a Heat Pipe Power System model as well as the radiotoxicity and radioactivity during lifetime cycles. In the new version, the code presents features such as multi-platform execution and automatic results analysis. Improvements made in the code allow it to perform simulations in a simpler and faster way without compromising accuracy. Initially, the code generates a new input for MCNP based on the decisions of the user. After that, MCNP is run and data, such as recoverable energy per prompt fission neutron, reaction rates and keff, are automatically extracted from the output and used to calculate neutron flux and cross sections. These data are then used to construct new ORIGEN inputs, one for each cell in the core. Each new input is run on ORIGEN and generates outputs that represent the complete isotopic composition of the core on that time step. The results show good agreement between GB (Coupled Neutronic/Isotopic code) and Monteburns (Automated, Multi-Step Monte Carlo Burnup Code System), developed by the Los Alamos National Laboratory.

  14. Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

    NASA Astrophysics Data System (ADS)

    Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

    2017-02-01

    International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

  15. EBSD and TEM Characterization of High Burn-up Mixed Oxide Fuel

    SciTech Connect

    Teague, Melissa C.; Gorman, Brian P.; Miller, Brandon D.; King, Jeffrey

    2014-01-01

    Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5 mRem/h compared to approximately 1.1 R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had approximately 2.5x higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice approximately 25 um cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel.

  16. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

    SciTech Connect

    Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; Wagner, John C.

    2014-11-01

    This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application of the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.

  17. Instant release fraction and matrix release of high burn-up UO2 spent nuclear fuel: Effect of high burn-up structure and leaching solution composition

    NASA Astrophysics Data System (ADS)

    Serrano-Purroy, D.; Clarens, F.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; de Pablo, J.; Casas, I.; Giménez, J.; Martínez-Esparza, A.

    2012-08-01

    Two weak points in Performance Assessment (PA) exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of the so-called Instant Release Fraction (IRF) and the effect of High Burn-Up Structure (HBS). This manuscript focuses on the effect of HBS in matrix (long term) and instant release of a Pressurised Water Reactor (PWR) SNF irradiated in a commercial reactor with a mean Burn-Up (BU) of 60 GWd/tU. In order to study the HBS contribution, two samples from different radial positions have been prepared. One from the centre of the SNF, labelled CORE, and one from the periphery, enriched with HBS and labelled OUT. Static leaching experiments have been carried out with two synthetic leaching solutions: bicarbonate (BIC) and Bentonitic Granitic Groundwater (BGW), and in all cases under oxidising conditions. IRF values have been calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP). In all studied cases, some radionuclides (RN): Rb, Sr and Cs, have shown higher release rates than uranium, especially at the beginning of the experiment, and have been considered as IRF. Redox sensitive RN like Mo and Tc have been found to dissolve slightly faster than uranium and further studies might be needed to confirm if they can also be considered part of the IRF. Most of the remaining studied RN, mainly actinides and lanthanides, have been found to dissolve congruently with the uranium matrix. Finally, Zr, Ru and Rh presented lower release rates than the matrix. Higher matrix release has been determined for CORE than for OUT samples showing that the formation of HBS might have a protective effect against the oxidative corrosion of the SNF. On the contrary, no significant differences have been observed between the two studied leaching solutions (BIC and BGW). Two different IRF contributions have been determined. One corresponding to the fraction of inventory segregated in the external open grain boundaries, directly available to water and

  18. Assessment of Fission Product Cross-Section Data for Burnup Credit Applications

    SciTech Connect

    Leal, Luiz C; Derrien, Herve; Dunn, Michael E; Mueller, Don

    2007-12-01

    Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the

  19. Process development and fabrication of space station type aluminum-clad graphite epoxy struts

    NASA Technical Reports Server (NTRS)

    Ring, L. R.

    1990-01-01

    The manufacture of aluminum-clad graphite epoxy struts, designed for application to the Space Station truss structure, is described. The strut requirements are identified, and the strut material selection rationale is discussed. The manufacturing procedure is described, and shop documents describing the details are included. Dry graphite fiber, Pitch-75, is pulled between two concentric aluminum tubes. Epoxy resin is then injected and cured. After reduction of the aluminum wall thickness by chemical milling the end fittings are bonded on the tubes. A discussion of the characteristics of the manufactured struts, i.e., geometry, weight, and any anomalies of the individual struts is included.

  20. Stressed waveguides with tubular depressed-cladding inscribed in phosphate glasses by femtosecond hollow laser beams.

    PubMed

    Long, Xuewen; Bai, Jing; Zhao, Wei; Stoian, Razvan; Hui, Rongqing; Cheng, Guanghua

    2012-08-01

    We report on the single-step fabrication of stressed optical waveguides with tubular depressed-refractive-index cladding in phosphate glasses by the use of focused femtosecond hollow laser beams. Tubelike low index regions appear under direct exposure due to material rarefaction following expansion. Strained compacted zones emerged in domains neighboring the tubular track of lower refractive index, and waveguiding occurs mainly within the tube core fabricated by the engineered femtosecond laser beam. The refractive index profile of the optical waveguide was reconstructed from the measured transmitted near-field intensity.

  1. Optical and mechanical properties of hollow-core fibers with cobweb cladding structure

    NASA Astrophysics Data System (ADS)

    Chen, Mingyang; Yu, Rongjin; Tian, Zhenguo; Bai, Xiangzhong

    2006-02-01

    This paper explores the possibility of realizing a hollow-core optical fiber, whose cladding is composed of cylindrical alternating layers of air and high-index material with supporting structure. The optical properties and the design criteria of the proposed fiber are evaluated by the compact two-dimensional (2D) finite-difference time-domain (FDTD) method. In particular, the influence of the number and width of supporting strips on the leakage loss of the fiber is investigated. Furthermore, the mechanical performances of the fiber are estimated by finite-element method, confirming that hollow-core fibers with a reasonable size and number of supporting strips are reliable.

  2. Formation of core-shell structure in high entropy alloy coating by laser cladding

    NASA Astrophysics Data System (ADS)

    Zhang, Hui; Wu, Wanfei; He, Yizhu; Li, Mingxi; Guo, Sheng

    2016-02-01

    The formation of core-shell structure is an interesting phenomenon occurring during the solidification process, due to the liquid phase separation. The formation of core-shell structure in high-entropy alloys, a new class of advanced metallic materials, has not been reported previously, and thus constitutes an intriguing scientific question. Here, we firstly report the formation of core-shell structure in one laser cladded high-entropy alloy, where we show the nanosized-Y2O3 powder addition, serves as the catalyst for the liquid phase separation.

  3. Optical monitoring of high power direct diode laser cladding

    NASA Astrophysics Data System (ADS)

    Liu, Shuang; Farahmand, Parisa; Kovacevic, Radovan

    2014-12-01

    Laser cladding is one of the most advanced surface modification techniques which can be used to build and repair high-value components. High power direct diode laser (HPDDL) offers unique quality and cost advantages over other lasers (CO2, Nd:YAG). Especially its rectangular laser beam with top-hat intensity distribution makes HPDDL an ideal tool for large area cladding. In order to utilize this technique successfully, the development of on-line monitoring and process control is necessary. In this study, an optical monitoring system consisting of a high-speed CCD camera, a pyrometer, and an infrared camera was used to analyze the mass- and heat-transfer in the cladding process. The particle transport in flight was viewed by a high-speed CCD camera; the interaction between powder flow and laser beam was observed by an infrared camera; and the thermal behavior of the molten pool was recorded by the pyrometer and the infrared camera. The effects of the processing parameters on the laser attenuation, particle heating and clad properties were investigated based on the obtained signals. The optical monitoring method improved the understanding about mutual interrelated phenomena in the cladding process.

  4. Cladding Alloys for Fluoride Salt Compatibility Final Report

    SciTech Connect

    Muralidharan, Govindarajan; Wilson, Dane F; Santella, Michael L; Holcomb, David Eugene

    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.

  5. Raman and XPS characterization of fuel-cladding interactions using miniature specimens

    SciTech Connect

    Windisch, Charles F.; Henager, Charles H.; Engelhard, Mark H.; Bennett, Wendy D.

    2009-01-01

    Laser Raman spectroscopy was evaluated as a tool for studying fuel-cladding chemical interactions at elevated temperatures. Materials and conditions were selected to simulate the interface of oxide fuels and fission products with high-temperature cladding materials for TRU-MOX fueled reactors. Both ex-situ and in-situ spectroscopy measurements were performed using polished HT-9 disks, uncoated and coated with yttria-stabilized zirconia, that were exposed to air oxidation at temperatures between 873-973K. Ex-situ measurements (under ambient conditions) were conducted to identify oxide phases, determine oxidation mechanisms and approximate film growth rates with an optimal signal-to-noise for the equipment used. Subsequently performed in-situ measurements were used to evaluate the sensitivity of the technique for measurements at elevated temperature in a hot-stage. Raman spectra were supported with x-ray photoelectron spectroscopy depth profiling. The results, which are for non-fueled materials in this study, illustrated a method for fast screening of candidate alloys with actinide-based MOX fuel mixtures.

  6. Ceramics reinforced metal base composite coatings produced by CO II laser cladding

    NASA Astrophysics Data System (ADS)

    Yang, Xichen; Wang, Yu; Yang, Nan

    2008-03-01

    Due to the excellent performance in high strength, anti-temperature and anti-wear, ceramics reinforced metal base composite material was used in some important fields of aircraft, aerospace, automobile and defense. The traditional bulk metal base composite materials are the expensive cost, which is limited in its industrial application. Development of laser coating of ceramics reinforced metal base composite is very interesting in economy. This paper is focused on three laser cladding ceramics coatings of SiC particle /Al matrix , Al IIO 3 powder/ Al matrix and WC + Co/mild steel matrix. Powder particle sizes are of 10-60μm. Chemical contents of aluminum matrix are of 3.8-4.0% Cu, 1.2-1.8% Mg, 0.3-0.99% Mn and balance Al. 5KW CO II laser, 5 axes CNC table, JKF-6 type powder feeder and co-axis feeder nozzle are used in laser cladding. Microstructure and performance of laser composite coatings have been respectively examined with OM,SEM and X-ray diffraction. Its results are as follows : Microstructures of 3C-,6H- and 5H- SiC particles + Al + Al 4SiC 4 + Si in SiC/Al composite, hexagonal α-Al IIO 3 + cubic γ-Al IIO 3 + f.c.c Al in Al IIO 3 powder/ Al composite and original WC particles + separated WC particles + eutectic WC + γ-Co solid solution + W IIC particles in WC + Co/steel coatings are respectively recognized. New microstructures of 5H-SiC in SiC/Al composite, cubic γ-Al IIO 3 in Al IIO 3 composite and W IIC in WC + Co/ steel composite by laser cladding have been respectively observed.

  7. Fuel/cladding compatibility in irradiated metallic fuel pins at elevated temperatures

    SciTech Connect

    Tsai, Hanchung.

    1990-04-01

    Over fifty fuel/cladding compatibility tests on irradiated metallic fuel specimens have been conducted in an in-cell facility at elevated temperatures. At temperatures below 700--725{degree}C, no fuel/cladding interaction was noted in tests up to 7 h. Liquid-phase cladding penetration occurred in some of the tests at temperatures greater than 725--750{degree}C. The effective rates of liquid- phase cladding penetration of six different fuel/cladding combinations during 1-h testing are reported. After the initial liquefaction at the fuel/cladding interface, which may be affected by the solid-state diffusional interaction during the steady-state irradiation, the rate of further cladding penetration stays constant or decreases with time. There was no runaway cladding penetration in the latter part of a heating cycle.

  8. Influence Of The Laser Cladding Strategies On The Mechanical Properties Of Inconel 718

    SciTech Connect

    Lamikiz, A.; Tabernero, I.; Ukar, E.; Lopez de Lacalle, L. N.

    2011-01-17

    This work presents different experimental results of the mechanical properties of Inconel registered 718 test parts built-up by laser cladding. Recently, turbine manufacturers for aeronautical sector have presented high interest on laser cladding processes. This process allows building fully functional structures on superalloys, such as Inconel registered 718, with high flexibility on complex shapes. However, there is limited data on mechanical properties of the laser cladding structures. Moreover, the available data do not include the influence of process parameters and laser cladding strategies. Therefore, a complete study of the influence of the laser cladding parameters and mainly, the variation of the tensile strength with the laser cladding strategy is presented. The results show that there is a high directionality of mechanical properties, depending on the strategies of laser cladding process. In other words, the test parts show a fiber -like structure that should be considered on the laser cladding strategy selection.

  9. Influence Of The Laser Cladding Strategies On The Mechanical Properties Of Inconel 718

    NASA Astrophysics Data System (ADS)

    Lamikiz, A.; Tabernero, I.; Ukar, E.; López de Lacalle, L. N.; Delgado, J.

    2011-01-01

    This work presents different experimental results of the mechanical properties of Inconel® 718 test parts built-up by laser cladding. Recently, turbine manufacturers for aeronautical sector have presented high interest on laser cladding processes. This process allows building fully functional structures on superalloys, such as Inconel® 718, with high flexibility on complex shapes. However, there is limited data on mechanical properties of the laser cladding structures. Moreover, the available data do not include the influence of process parameters and laser cladding strategies. Therefore, a complete study of the influence of the laser cladding parameters and mainly, the variation of the tensile strength with the laser cladding strategy is presented. The results show that there is a high directionality of mechanical properties, depending on the strategies of laser cladding process. In other words, the test parts show a fiber -like structure that should be considered on the laser cladding strategy selection.

  10. Nanoscale light–matter interactions in atomic cladding waveguides

    PubMed Central

    Stern, Liron; Desiatov, Boris; Goykhman, Ilya; Levy, Uriel

    2013-01-01

    Alkali vapours, such as rubidium, are being used extensively in several important fields of research such as slow and stored light nonlinear optics quantum computation, atomic clocks and magnetometers. Recently, there is a growing effort towards miniaturizing traditional centimetre-size vapour cells. Owing to the significant reduction in device dimensions, light–matter interactions are greatly enhanced, enabling new functionalities due to the low power threshold needed for nonlinear interactions. Here, taking advantage of the mature platform of silicon photonics, we construct an efficient and flexible platform for tailored light–vapour interactions on a chip. Specifically, we demonstrate light–matter interactions in an atomic cladding waveguide, consisting of a silicon nitride nano-waveguide core with a rubidium vapour cladding. We observe the efficient interaction of the electromagnetic guided mode with the rubidium cladding and show that due to the high confinement of the optical mode, the rubidium absorption saturates at powers in the nanowatt regime. PMID:23462991

  11. Nanoscale light-matter interactions in atomic cladding waveguides.

    PubMed

    Stern, Liron; Desiatov, Boris; Goykhman, Ilya; Levy, Uriel

    2013-01-01

    Alkali vapours, such as rubidium, are being used extensively in several important fields of research such as slow and stored light nonlinear optics quantum computation, atomic clocks and magnetometers. Recently, there is a growing effort towards miniaturizing traditional centimetre-size vapour cells. Owing to the significant reduction in device dimensions, light-matter interactions are greatly enhanced, enabling new functionalities due to the low power threshold needed for nonlinear interactions. Here, taking advantage of the mature platform of silicon photonics, we construct an efficient and flexible platform for tailored light-vapour interactions on a chip. Specifically, we demonstrate light-matter interactions in an atomic cladding waveguide, consisting of a silicon nitride nano-waveguide core with a rubidium vapour cladding. We observe the efficient interaction of the electromagnetic guided mode with the rubidium cladding and show that due to the high confinement of the optical mode, the rubidium absorption saturates at powers in the nanowatt regime.

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

    SciTech Connect

    Collins, Emory D; Johnson, Jared A.; Hylton, Tom D.; Brunson, Ronald Ray; Hunt, Rodney Dale; DelCul, Guillermo Daniel; Bradley, Eric Craig; Spencer, Barry B.

    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.

  13. In situ synthesis of hydroxyapatite coating by laser cladding.

    PubMed

    Wang, D G; Chen, C Z; Ma, J; Zhang, G

    2008-10-15

    HA bioceramic coatings were synthesized on titanium substrate by laser cladding using cheap calcium carbonate and calcium hydrogen phosphate. The thermodynamic condition for synthesizing HA was calculated by software Matlab 5.0, the microstructure and phase analysis of laser clad HA bioceramic coatings were studied by electron probe microanalyser (EPMA), X-ray diffractometer (XRD) and transmission electron microscopy (TEM). The theoretical results show that the Gibbs free enthalpy for the synthesis of HA phase is satisfied, and the presence of HA phase in the clad coatings was then further verified by XRD and the selected area diffraction patterns. When the laser power is 600W and the scanning speed is 3.5mm/s, the compact HA bioceramic coatings were obtained, which have cellular dendritic structure and consist of the phases of HA, alpha-Ca(2)P(2)O(7), CaO and CaTiO(3).

  14. Clad Nd:YAG fibers for laser applications

    SciTech Connect

    Digonnet, M.J.F.; Shaw, H.J.; Gaeta, C.J.; O'meara, D.

    1987-05-01

    The implementation of an extrusion method to clad Nd:YAG single crystal fibers with index-matched glasses (Delta-n = 0.048) is reported. A propagation-loss coefficient of 0.08 dB/cm was measured for the fundamental mode of a 41-micron-diameter glass-clad fiber laser, an improvement of about one order of magnitude over unclad fibers. Guided clad fiber lasers operated at 1.064 microns with thresholds as low as 0.3-0.5 mW and up to 65-mW CW output power are also reported. The origins of and means of reducing the residual loss are discussed. 14 references.

  15. Fabrication of stainless steel clad tubing. [gas pressure bonding

    NASA Technical Reports Server (NTRS)

    Kovach, C. W.

    1978-01-01

    The feasibility of producing stainless steel clad carbon steel tubing by a gas pressure bonding process was evaluated. Such a tube product could provide substantial chromium savings over monolithic stainless tubing in the event of a serious chromium shortage. The process consists of the initial assembly of three component tubesets from conventionally produced tubing, the formation of a strong metallurgical bond between the three components by gas pressure bonding, and conventional cold draw and anneal processing to final size. The quality of the tubes produced was excellent from the standpoint of bond strength, mechanical, and forming properties. The only significant quality problem encountered was carburization of the stainless clad by the carbon steel core which can be overcome by further refinement through at least three different approaches. The estimated cost of clad tubing produced by this process is greater than that for monolithic stainless tubing, but not so high as to make the process impractical as a chromium conservation method.

  16. Analysis of burnup and isotopic compositions of BWR 9 x 9 UO{sub 2} fuel assemblies

    SciTech Connect

    Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T.

    2012-07-01

    In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO{sub 2} fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for {sup 238}Pu, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148}Nd, {sup 134}Cs, {sup 154}Eu, {sup 152}Sm, {sup 154}Gd, and {sup 157}Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

  17. Metallic inert matrix fuel concept for minor actinides incineration to achieve ultra-high burn-up

    NASA Astrophysics Data System (ADS)

    Lipkina, K.; Savchenko, A.; Skupov, M.; Glushenkov, A.; Vatulin, A.; Uferov, O.; Ivanov, Y.; Kulakov, G.; Ershov, S.; Maranchak, S.; Kozlov, A.; Maynikov, E.; Konova, K.

    2014-09-01

    The advantages of using Inert Matrix Fuel (IMF) in a design of an isolated arrangement of fuel are considered, with emphasis on, low temperatures in the fuel center, achievement of high burn-ups, and an environment friendly process for the fuel element fabrication. Changes in the currently existing concept of IMF usage are suggested, involving novel IMF design in the nuclear fuel cycle.

  18. 77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-02

    ... criticality safety analyses of pressurized water reactor spent nuclear fuel (SNF) in transportation packages... Doc No: 2012-10618] NUCLEAR REGULATORY COMMISSION [NRC-2012-0100] Burnup Credit in the Criticality... the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.'' This...

  19. The structure and corrosion resistance of Ti-Nb layers formed on the surface of CP-titanium by electron beam cladding

    NASA Astrophysics Data System (ADS)

    Polyakov, Igor A.; Kolesnikova, Kseniya A.; Lozhkina, Elena A.; Samoylenko, Vitaliy V.

    2016-11-01

    The structure and corrosion resistance of Ti-Nb layers formed on the surface of CP-titanium plates by electron beam cladding in vacuum has been investigated. It is found that the cladded material is characterized by a defect-free structure. Optical and scanning electron microscopy showed that the martensitic structure is observed in the cladded layers. The concentration of niobium in the layers was 7.7 wt %. Surface alloying of niobium provides a 2.6-fold increase in the corrosion resistance of titanium in boiling 65% solution of nitric acid. The corrosion resistance of Ti-Nb layers in boiling 5% solution of hydrochloric acid is almost 10.6 times higher in comparison with commercially pure titanium.

  20. Process for Making Optical Fibers from Core and Cladding Glass Rods.

    DTIC Science & Technology

    1996-08-12

    77,577 PATENT APPLICATION Inventor’s Name: Sanghera et al. ORIGINAL 2 3 PROCESS FOR MAKING OPTICAL FIBERS FROM CORE AND CLADDING GLASS 4 RODS S...core and cladding glass rods and to the fibers 10 made by the process. More particularly, the invention relates to 11 separately melting core and...cladding glass rods and combining the 12 melts proximate a fiber drawing orifice so that the core glass is 13 surrounded by the cladding glass and

  1. Evaluation of stainless steel cladding for use in current design LWRs. Final report

    SciTech Connect

    Strasser, A.; Santucci, J.; Lindquist, K.; Yario, W.; Stern, G.; Goldstein, L.; Joseph, L.

    1982-12-01

    The design of stainless steel-clad LWR fuel and its performance at steady-state, transient, and accident conditions were reviewed. The objective was to evaluate the potential benefits and disadvantages of substituting stainless steel-clad fuel for the currently used Zircaloy-clad fuel. For a large, modern PWR, the technology and the fuel-cycle costs of stainless steel- and Zircaloy-clad fuels were compared.

  2. Results of NDE Technique Evaluation of Clad Hydrides

    SciTech Connect

    Kunerth, Dennis C.

    2014-09-01

    This report fulfills the M4 milestone, M4FT-14IN0805023, Results of NDE Technique Evaluation of Clad Hydrides, under Work Package Number FT-14IN080502. During service, zirconium alloy fuel cladding will degrade via corrosion/oxidation. Hydrogen, a byproduct of the oxidation process, will be absorbed into the cladding and eventually form hydrides due to low hydrogen solubility limits. The hydride phase is detrimental to the mechanical properties of the cladding and therefore it is important to be able to detect and characterize the presence of this constituent within the cladding. Presently, hydrides are evaluated using destructive examination. If nondestructive evaluation techniques can be used to detect and characterize the hydrides, the potential exists to significantly increase test sample coverage while reducing evaluation time and cost. To demonstrate the viability this approach, an initial evaluation of eddy current and ultrasonic techniques were performed to demonstrate the basic ability to these techniques to detect hydrides or their effects on the microstructure. Conventional continuous wave eddy current techniques were applied to zirconium based cladding test samples thermally processed with hydrogen gas to promote the absorption of hydrogen and subsequent formation of hydrides. The results of the evaluation demonstrate that eddy current inspection approaches have the potential to detect both the physical damage induced by hydrides, e.g. blisters and cracking, as well as the combined effects of absorbed hydrogen and hydride precipitates on the electrical properties of the zirconium alloy. Similarly, measurements of ultrasonic wave velocities indicate changes in the elastic properties resulting from the combined effects of absorbed hydrogen and hydride precipitates as well as changes in geometry in regions of severe degradation. However, for both approaches, the signal responses intended to make the desired measurement incorporate a number of contributing

  3. Welding fixture for nuclear fuel pin cladding assemblies

    DOEpatents

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

    1984-02-22

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

  4. Welding fixture for nuclear fuel pin cladding assemblies

    DOEpatents

    Oakley, David J.; Feld, Sam H.

    1986-01-01

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

  5. 78 FR 9676 - Clad Steel Plate From Japan: Continuation of Antidumping Duty Order

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-11

    ... International Trade Administration Clad Steel Plate From Japan: Continuation of Antidumping Duty Order AGENCY... International Trade Commission (ITC) that revocation of the antidumping duty order on clad steel plate from... sunset review of the antidumping duty order on clad steel plate from Japan pursuant to section 751(c)...

  6. Cladding glass ceramic for use in high powered lasers

    DOEpatents

    Marker, Alexander J.; Campbell, John H.

    1998-01-01

    A Cu-doped/Fe-doped low expansion glass ceramic composition comprising in Wt. %: SiO{sub 2} 50--65; Al{sub 2}O{sub 3} 18--27; P{sub 2}O{sub 5} 0--10; Li{sub 2}O 2--6; Na{sub 2}O 0--2; K{sub 2}O 0--2; B{sub 2}O{sub 3} 0--1; MgO 0--4; ZnO 0--5; CaO 0--4; BaO 0--5; TiO{sub 2} 1--3; ZrO{sub 3} 1--3; As{sub 2}O{sub 3} 0--1.5; Sb{sub 2}O{sub 3} 0--1.5; CuO 0--3; and Fe{sub 2}O{sub 3} 0--1 wherein the total amount of SiO{sub 2}, Al{sub 2}O{sub 3} and P{sub 2}O{sub 5} is 80--89 wt. %, and said glass ceramic contains as a dopant 0.1--3 wt. % CuO, 0.1--1 wt. % Fe{sub 2}O{sub 3} or a combined CuO+Fe{sub 2}O{sub 3} amount of 0.1--4 wt. %. The glass ceramic composition is suitable for use as a cladding material for solid laser energy storage mediums as well as for use in beam attenuators for measuring laser energy level and beam blocks or beam dumps used for absorbing excess or unused laser energy.

  7. Cladding glass ceramic for use in high powered lasers

    DOEpatents

    Marker, A.J.; Campbell, J.H.

    1998-02-17

    A Cu-doped/Fe-doped low expansion glass ceramic composition comprising in Wt. %: SiO{sub 2} 50--65; Al{sub 2}O{sub 3} 18--27; P{sub 2}O{sub 5} 0--10; Li{sub 2}O 2--6; Na{sub 2}O 0--2; K{sub 2}O 0--2; B{sub 2}O{sub 3} 0--1; MgO 0--4; ZnO 0--5; CaO 0--4; BaO 0--5; TiO{sub 2} 1--3; ZrO{sub 3} 1--3; As{sub 2}O{sub 3} 0--1.5; Sb{sub 2}O{sub 3} 0--1.5; CuO 0--3; and Fe{sub 2}O{sub 3} 0--1 wherein the total amount of SiO{sub 2}, Al{sub 2}O{sub 3} and P{sub 2}O{sub 5} is 80--89 wt. %, and said glass ceramic contains as a dopant 0.1--3 wt. % CuO, 0.1--1 wt. % Fe{sub 2}O{sub 3} or a combined CuO+Fe{sub 2}O{sub 3} amount of 0.1--4 wt. %. The glass ceramic composition is suitable for use as a cladding material for solid laser energy storage mediums as well as for use in beam attenuators for measuring laser energy level and beam blocks or beam dumps used for absorbing excess or unused laser energy.

  8. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    SciTech Connect

    Trellue, Holly Renee; Fugate, Michael Lynn; Tobin, Stephen Joesph

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  9. Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

    SciTech Connect

    Ilas, Germina; Betzler, Benjamin R; Ade, Brian J

    2017-01-01

    This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay, and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.

  10. Accident source terms for boiling water reactors with high burnup cores.

    SciTech Connect

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  11. Validation of a method for prediction of isotopic concentrations in burnup credit applications

    SciTech Connect

    DeHart, M.D.; Hermann, O.W.; Parks, C.V.

    1995-09-01

    Unlike fresh fuel assumptions typically employed in the criticality safety analysis of spent fuel configurations, burnup credit applications rely on depletion and decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in subsequent criticality calculations to assess the reduced worth of the spent fuel. To validate the codes and data used in depletion approaches, experimental measurements are compared with numerical predictions for relevant spent fuel samples. This paper describes a set of experimentally characterized pressurized-water-reactor (PWR) fuel samples and provides a comparison to results of SCALE-4 depletion calculations. An approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations.

  12. Effects of microstructural constraints on the transport of fission products in uranium dioxide at low burnups

    NASA Astrophysics Data System (ADS)

    Lim, Harn Chyi; Rudman, Karin; Krishnan, Kapil; McDonald, Robert; Dickerson, Patricia; Gong, Bowen; Peralta, Pedro

    2016-08-01

    Diffusion of fission gases in UO2 is studied at low burnups, before bubble growth and coalescence along grain boundaries (GBs) become dominant, using a 3-D finite element model that incorporates actual UO2 microstructures. Grain boundary diffusivities are assigned based on crystallography with lattice and GB diffusion coupled with temperature to account for temperature gradients. Heterogeneity of GB properties and connectivity can induce regions where concentration is locally higher than without GB diffusion. These regions are produced by "bottlenecks" in the GB network because of lack of connectivity among high diffusivity GBs due to crystallographic constraints, and they can lead to localized swelling. Effective diffusivities were calculated assuming a uniform distribution of high diffusivity among GBs. Results indicate an increase over the bulk diffusivity with a clear grain size effect and that connectivity and properties of different GBs become important factors on the variability of fission product concentration at the microscale.

  13. Development of burnup dependent fuel rod model in COBRA-TF

    NASA Astrophysics Data System (ADS)

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  14. Determination of nuclear fuel burn-up axial profile by neutron emission measurement

    NASA Astrophysics Data System (ADS)

    Prokopowicz, Rafal; Pytel, Krzysztof

    2016-12-01

    Burning-up of nuclear fuel is usually not a space-isotropic phenomenon. It depends on both the neutron flux density and energy spectrum distribution during fuel operation in a nuclear reactor. This paper presents the method of measurement of burn-up spatial distribution of spent nuclear fuel element. The method is based on recording of the neutron emission from investigated fuel element. Based on performed analyses and calculations, a suitable measuring setup has been designed and constructed. The subjects of investigation were fuel elements used in the MARIA research reactor, operated by National Centre for Nuclear Research in Świerk, Poland. The results of measurements made over a period of several years by means of the described method are presented in the paper.

  15. [Method and apparatus for applying metal cladding

    SciTech Connect

    Bosna, A.A.

    1989-09-27

    This progress report contains information related to the development of durable corrosion resistant coating materials for the hulls of ships and offshore platforms. Also contained are the details of an apparatus used to apply this material. (JEF)

  16. Process for the manufacture of seamless metal-clad fiber-reinforced organic matrix composite structures

    NASA Technical Reports Server (NTRS)

    Bluck, Raymond M. (Inventor); Bush, Harold G. (Inventor); Johnson, Robert R. (Inventor)

    1991-01-01

    A process for producing seamless metal-clad composite structures includes providing a hollow, metallic inner member and an outer sleeve to surround the inner member and define an inner space therebetween. A plurality of continuous reinforcing fibers is attached to the distal end of the outside diameter of the inner member, and the inner member is then introduced, distal end first, into one end of the outer sleeve. The inner member is then moved, distal end first, into the outer sleeve until the inner member is completely enveloped by the outer sleeve. A liquid matrix material is then injected into the space containing the reinforcing fibers between the inner member and the outer sleeve. Next a pressurized heat transfer medium is passed through the inner member to cure the liquid matrix material. Finally, the wall thickness of both the inner member and the outer sleeve are reduced to desired dimensions by chemical etching, which adjusts the thermal expansion coefficient of the metal-clad composite structure to a desired value.

  17. Laser Cladding of γ-TiAl Intermetallic Alloy on Titanium Alloy Substrates

    NASA Astrophysics Data System (ADS)

    Maliutina, Iuliia Nikolaevna; Si-Mohand, Hocine; Piolet, Romain; Missemer, Florent; Popelyukh, Albert Igorevich; Belousova, Natalya Sergeevna; Bertrand, Philippe

    2016-01-01

    The enhancement of titanium and titanium alloy's tribological properties is of major interest in many applications such as the aerospace and automotive industry. Therefore, the current research paper investigates the laser cladding of Ti48Al2Cr2Nb powder onto Ti6242 titanium alloy substrates. The work was carried out in two steps. First, the optimal deposition parameters were defined using the so-called "combined parameters," i.e., the specific energy E specific and powder density G. Thus, the results show that those combined parameters have a significant influence on the geometry, microstructure, and microhardness of titanium aluminide-formed tracks. Then, the formation of dense, homogeneous, and defect-free coatings based on optimal parameters has been investigated. Optical and scanning electron microscopy techniques as well as energy-dispersive spectroscopy and X-ray diffraction analyses have shown that a duplex structure consisting of γ-TiAl and α 2-Ti3Al phases was obtained in the coatings during laser cladding. Moreover, it was shown that produced coatings exhibit higher values of microhardness (477 ± 9 Hv0.3) and wear resistance (average friction coefficient is 0.31 and volume of worn material is 5 mm3 after 400 m) compared to those obtained with bare titanium alloy substrates (353 Hv0.3, average friction coefficient is 0.57 and a volume of worn material after 400 m is 35 mm3).

  18. Processing of pure Ti by rapid prototyping based on laser cladding

    NASA Astrophysics Data System (ADS)

    Arias-González, F.; del Val, J.; Comesaña, R.; Lusquiños, F.; Quintero, F.; Riveiro, A.; Boutinguiza, M.; Pou, J.

    2013-11-01

    Rapid prototyping based on laser cladding is an additive manufacturing (AM) process based on the overlapping of cladding tracks to produce functional components. Powder or wire are fed into a melting pool created using laser radiation as a heat source and the relative movement between the beam and the work piece makes possible to generate pieces layer-by-layer. This technique can be applied for any material which can be melted and the components can be manufactured directly according to a computer aided design (CAD) model. Additive manufacturing is particularly interesting to produce titanium components because, in this case, the loss of material produced by subtractive manufacturing methods is highly costly. Moreover, titanium and its alloys are widely used in biomedical, aircraft, chemical and marine industries due to their biocompatibility, excellent corrosion resistance and superior strength-to-weight ratio. In this research work, a near-infrared laser delivering a maximum power of 500W is used to produce pure titanium thin parts. Dimensions and surface morphology are characterized using Optical Microscopy (OM) and Scanning Electron Microscopy (SEM), the hardness by nanoindentation and the composition by X-Ray Diffraction (XRD) and Energy Dispersive X-Ray Spectroscopy (EDS). The aim of this work is to establish the conditions under which satisfactory properties are obtained and to understand the relationship between microstructure/properties and deposition parameters.

  19. RF Plasma Torch System for Metal Matrix Composite Production in Nuclear Fuel Cladding

    NASA Astrophysics Data System (ADS)

    Holik, Eddie, III

    2007-10-01

    For the first time in 30 years, plans are afoot to build new fission power plants in the US. It is timely to develop technology that could improve the safety and efficiency of new reactors. A program of development for advanced fuel cycles and Generation IV reactors is underway. The path to greater efficiency is to increase the core operating temperature. That places particular challenges to the cladding tubes that contain the fission fuel. A promising material for this purpose is a metal matrix composite (MMC) in which ceramic fibers are bonded within a high-strength steel matrix, much like fiberglass. Current MMC technology lacks the ability to effectively bond traditional high-temperature alloys to ceramic strands. The purpose of this project is to design an rf plasma torch system to use titanium as a buffer between the ceramic fibers and the refractory outer material. The design and methods of using an rf plasma torch to produce a non-equilibrium phase reaction to bond together the MMC will be discussed. The effects of having a long lived fuel cladding in the design of future reactors will also be discussed.

  20. Material Property Correlations: Comparisons between FRAPCON-3.4, FRAPTRAN 1.4, and MATPRO

    SciTech Connect

    Luscher, Walter G.; Geelhood, Kenneth J.

    2010-08-01

    The U.S. Nuclear Regulatory Commission (NRC) uses the computer codes FRAPCON-3 and FRAPTRAN to model steady state and transient fuel behavior, respectively, in regulatory analysis. In order to effectively model fuel behavior, material property correlations must be used for a wide range of operating conditions (e.g. temperature and burnup). In this sense, a 'material property' is a physical characteristic of the material whose quantitative value is necessary in the analysis process. Further, the property may be used to compare the benefits of one material versus another. Generally speaking, the material properties of interest in regulatory analysis of nuclear fuel behavior are mechanical or thermodynamic in nature. The issue of what is and is not a 'material property' will never be universally resolved. In this report, properties such as thermal conductivity are included. Other characteristics of the material (e.g. fission gas release) are considered 'models' rather than properties, and are discussed elsewhere. Still others (e.g., neutron absorption cross-section) are simply not required in this specific analysis. The material property correlations for the FRAPCON-3 and FRAPTRAN computer codes were documented in NUREG/CR-6534 and NUREG/CR-6739, respectively. Some of these have been modified or updated since the original code documentation was published. The primary purpose of this report is to consolidate the current material property correlations used in FRAPCON-3 and FRAPTRAN into a single document. Material property correlations for oxide fuels, including uranium dioxide (UO2) and mixed oxide (MOX) fuels, are described in Section 2. Throughout this document, the term MOX will be used to describe fuels that are blends of uranium and plutonium oxides, (U,Pu)O2. The properties for uranium dioxide with other additives (e.g., gadolinia) are also discussed. Material property correlations for cladding materials and gases are described in Sections 3 and 4, respectively