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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. 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

  3. 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.

  4. Cladding stress during extended storage of high burnup spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Raynaud, Patrick A. C.; Einziger, Robert E.

    2015-09-01

    In an effort to assess the potential for low temperature creep and delayed hydride cracking failures in high burnup spent fuel cladding during extended dry storage, the U.S. NRC analytical fuel performance tools were used to predict cladding stress during a 300 year dry storage period for UO2 fuel burned up to 65 GWd/MTU. Fuel swelling correlations were developed and used along with decay gas production and release fractions to produce circumferential average cladding stress predictions with the FRAPCON-3.5 fuel performance code. The resulting stresses did not result in cladding creep failures. The maximum creep strains accumulated were on the order of 0.54-1.04%, but creep failures are not expected below at least 2% strain. The potential for delayed hydride cracking was assessed by calculating the critical flaw size required to trigger this failure mechanism. The critical flaw size far exceeded any realistic flaw expected in spent fuel at end of reactor life.

  5. 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

  6. 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.

  7. Fuel rod and core materials investigations related to LWR extended burnup operation

    NASA Astrophysics Data System (ADS)

    Kolstad, Erik; Vitanza, Carlo

    1992-06-01

    The paper deals with tests and recent measurements related to extended burnup fuel performance and describes test facilities and results in the areas of waterside cladding corrosion and irradiation-assisted stress corrosion cracking (IASCC). Fuel temperature data suggest a gradual degradation of UO 2 thermal conductivity with exposure in the range 6-8% per 10 MWd/kgUO 2 at temperatures below 700°C. The effect on the fuel microstructure of interlinkage and resintering phenomena is shown by measuring the surface-to-volume ( S/ V) ratio of the fuel. Changes in S/V with burnup are correlated to power rating and fuel operating temperature. No evidence was found of enhanced fission gas release during load-follow operation in the burnup range 25-45 MWd/kgUO 2. The effect of high lithium concentration (high pH) on the corrosion behaviour of pre-irradiated high burnup Zircaloy-4 fuel rods subjected either to nucleate boiling or to one-phase cooling conditions was studied. The oxide thickness growth rates measured at an average burnup up to 40 MWd/kgUO 2 are consistent with literature data and show no evidence of corrosion enhancement due to the high lithium content and little effect of cooling regime. A test facility for exploring the effects of environmental variables on IASCC behaviour of in-core structural materials is described.

  8. Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

    NASA Astrophysics Data System (ADS)

    Uwaba, Tomoyuki; Ito, Masahiro; Maeda, Koji

    2011-09-01

    The C3M irradiation test, which was conducted in the experimental fast reactor, "Joyo", demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, "Monju". The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and 137Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

  9. 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

  10. Clad and duct materials property needs for sodium fast reactors: US perspective

    SciTech Connect

    Maloy, Stuart A

    2009-01-01

    In the Advanced Fuel Cycle Initiative, the clad and duct materials property needs are driven by the program direction of using a fast reactor to transmute minor actinides. To economically perform transmutation requires fuels to be irradiated to at least 200/0 burnup before discharge and recycle. The cladding materials used in sllch high dose irradiations will experience at least 200 dpa of exposure. The requirements are that the cladding must contain the fuel until discharge and the duct must maintain ductility and mechanical strength to direct the sodium coolant and withdraw the fuel at discharge. Presently, achieving a bumup of 20% in fast reactor fuels is a significant challenge. The materials property needs for clad materials are the following in order of importance: (1) Radiation resistance: the material must be tolerant of radiation damage up to 200 dpa or greater depending on the final burnup desired at temperatures ranging from 350-600 C (temperature range for maximum dose is 400-500 C). Thus, the material must be tolerant of swelling, irradiation creep, and embrittlement effects. (2) Mechanical properties: tensile properties (strength, ductility), creep resistance, fracture toughness and fatigue resistance must all sufficient to contain the fuel during irradiation and discharge. (3) Fuel Clad Chemical Interaction: the cladding must be resistant to chemical interaction with the fuel during irradiation to prevent clad failure. (4) Fabrication and joining: the material must be able to be fabricated into component parts and joined to contain the fuel during the irradiation and during discharge. (5) Corrosion resistance: the material must be tolerant of the sodium coolant, with minimal corrosion rate. (6) Reprocessing: the cladding must be dissolvable to enable fuel reprocessing. (7) Long-term stability: the clad and duct materials must be stable through 6 or more years of irradiation depending on the final fuel burnup desired. (8) Thermal properties: thermal

  11. 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.

  12. 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.

  13. 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.

  14. Improved cladding nano-structured materials with self-repairing capabilities

    SciTech Connect

    Popa-Simil, L.

    2012-07-01

    When designing nuclear reactors or the materials that go into them, one of the key challenges is finding materials that can withstand an outrageously extreme environment. In addition to constant bombardment by radiation, reactor materials may be subjected to extremes in temperature, physical stress, and corrosive conditions. A limitation in fuel burnup is and usage of the nuclear fuel material is related to the structural material radiation damage, that makes the fuel be removed with low-burnup and immobilized in the waste storage pools. The advanced burnup brings cladding material embitterment due to radiation damage effects corroborated with corrosion effects makes the fuel pellet life shorter. The novel nano-clustered structured sintered material may mitigate simultaneously the radiation damage and corrosion effects driving to more robust structural materials that may make the nuclear reactor safer and more reliable. The development of nano-clustered sinter alloys provides new avenues for further examination of the role of grain boundaries and engineered material interfaces in self-healing of radiation-induced defects driving to the design of highly radiation-tolerant materials for the next generation of nuclear energy applications. (authors)

  15. 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.

  16. Material selection for accident tolerant fuel cladding

    DOE PAGESBeta

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

    2015-09-14

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

  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

    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.

  19. Polymer materials as modified optical fiber cladding for chemical sensors

    NASA Astrophysics Data System (ADS)

    Yuan, Jianming

    An intrinsic fiber optic chemical sensor has been designed and developed by using a polymer material as a modified fiber cladding. The sensor is constructed by replacing a certain portion of the original cladding with a chemically sensitive material, specifically, polyaniline or polypyrrole. Both the light absorption coefficient and the refractive index of the polymers change upon the exposure to different chemical vapors. These changes induce the optical intensity modulation of the fiber optic sensor. Polyaniline or polypyrrole is coated as the modified cladding by either spin-cast or in-situ deposition method for sensing HCl, NH3, H 2O2, and H4N2 vapors. All sensors show rapid and strong response to the chemical vapors. Thus, these sensors demonstrate that polyaniline and polypyrrole are viable candidate materials for the detection of volatile toxic gases. Sensors exhibit better performance when correct parameters, such as modification area, in-situ deposition time, and spin-rate, are used in the cladding modification process. The reversibility of the sensor depends on the reaction between the modified cladding material and the chemical vapors. Polyaniline cladding has better reversibility than polypyrrole. The optimized sensor response and sensitivity can be achieved by selecting an incident light with suitable wavelength, power, and incident angle.

  20. Accident Performance of Light Water Reactor Cladding Materials

    SciTech Connect

    Nelson, Andrew T.

    2012-07-24

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

  1. 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)

  2. 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

  3. Cladding and Duct Materials for Advanced Nuclear Recycle Reactors

    SciTech Connect

    Allen, Todd R.; Busby, Jeremy T; Klueh, Ronald L; Maloy, S; Toloczko, M

    2008-01-01

    The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP s advanced nuclear recycle reactors program.

  4. Cladding and duct materials for advanced nuclear recycle reactors

    NASA Astrophysics Data System (ADS)

    Allen, T. R.; Busby, J. T.; Klueh, R. L.; Maloy, S. A.; Toloczko, M. B.

    2008-01-01

    The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.

  5. 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.

  6. 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.

  7. Analysis of transmission characteristics of doubly clad fibers with an inner cladding made of uniaxial crystal materials

    NASA Astrophysics Data System (ADS)

    Xiaoping, Zhang; Zhihong, Tan

    2002-04-01

    A doubly clad optical fiber with an inner cladding made of a uniaxial crystal material whose optical axis is parallel to the fiber axis was proposed, and exact characteristic equations of vector modes were derived. The influence of the ratio ( kcl) of the extraordinary to the ordinary ray indexes upon the waveguide dispersion was examined in detail. In view of the impossibility to deduce the expression of waveguide dispersion directly due to the complexity of the characteristic equations, a feasible approach to calculate waveguide dispersion was established. The calculated results indicate that the values of waveguide dispersion can be effectively changed through variation of kcl without changing the geometrical and optical parameters ( S and R). The influences of kcl, S and R on the propagation and cutoff characteristics of the low order modes are also analyzed.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  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. Corrosion Assessment of Candidate Materials for Fuel Cladding in Canadian SCWR

    NASA Astrophysics Data System (ADS)

    Zeng, Yimin; Guzonas, David

    2016-02-01

    The supercritical water-cooled reactor (SCWR) is an innovative next generation reactor that offers many promising features, but the high-temperature high-pressure coolant introduces unique challenges to the long-term safe and reliable operation of in-core components, in particular the fuel cladding. To achieve high thermal efficiency, the Canadian SCWR concept has a coolant core outlet temperature of 625°C at 25 MPa with a peak cladding temperature as high as 800°C. International and Canadian research programs on corrosion issues in supercritical water have been conducted to support the SCWR concept. This paper provides a brief review of corrosion in supercritical water and summarizes the Canadian corrosion assessment work on potential fuel cladding materials. Five alloys, SS 347H, SS310S, Alloy 800H, Alloy 625 and Alloy 214, have been shown to have sufficient corrosion resistance to be used as the fuel cladding. Additional work, including tests in an in-reactor loop, is needed to confirm that these alloys would work as the fuel cladding in the Canadian SCWR.

  14. Oxidation of Fuel Cladding Candidate Materials in Steam Environments at High Temperature and Pressure

    SciTech Connect

    Cheng, Ting; Keiser, James R; Brady, Michael P; Terrani, Kurt A; Pint, Bruce A

    2012-01-01

    Under certain severe accident conditions, the fuel rods of nuclear power plants are exposed to high temperature/pressure steam environments in which the Zr alloy cladding is rapidly oxidized. As alternative claddings, the oxidation resistances of SiC-based materials and stainless steels with high Cr and/or Al additions have been examined from 800-1200 C in high-pressure steam environments. Very low reaction kinetics were observed with alumina-forming FeCrAl alloys at 1200 C while Fe-Cr alloys with only 15-20% Cr were rapidly attacked.

  15. Oxidation of fuel cladding candidate materials in steam environments at high temperature and pressure

    NASA Astrophysics Data System (ADS)

    Cheng, Ting; Keiser, James R.; Brady, Michael P.; Terrani, Kurt A.; Pint, Bruce A.

    2012-08-01

    Under certain severe accident conditions, the fuel rods of nuclear power plants are exposed to high temperature/pressure steam environments in which the Zr alloy cladding is rapidly oxidized. As alternative claddings, the oxidation resistances of SiC-based materials and stainless steels with high Cr and/or Al additions have been examined from 800-1200 °C in high-pressure steam environments. Very low reaction kinetics were observed with alumina-forming FeCrAl alloys at 1200 °C while Fe-Cr alloys with only 15-20% Cr were rapidly attacked.

  16. Structural transformations in hull material clad by nitrogen stainless steel using various methods

    NASA Astrophysics Data System (ADS)

    Sagaradze, V. V.; Kataeva, N. V.; Mushnikova, S. Yu.; Khar'kov, O. A.; Kalinin, G. Yu.; Yampol'skii, V. D.

    2014-02-01

    Specimens of a 10N3KhDMBF shipbuilding hull steel were clad by a 04Kh20N6G11M2AFB nitrogen austenitic steel using various treatment conditions, which included hot rolling, austenitic facing, and explosive welding followed by hot rolling and heat treatment. Between the base and cladding materials, an intermediate layer with variable concentrations of chromium, manganese, and nickel was found, in which a martensitic structure was formed. In all the cases, the strength of bonding of the cladding layer to the hull steel (determined in tests for shear to fracture) was fairly high (σsh = 437-520 MPa). The only exception was the specimen produced by unidirectional facing without subsequent hot rolling (σsh = 308 MPa), in which nonfusions between the faced beads of stainless steel were detected.

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

    DOE PAGESBeta

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

    2016-09-28

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

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. Burnup estimation of fuel sourcing radioactive material based on monitored Cs and Pu isotopic activity ratios in Fukushima N. P. S. accident

    SciTech Connect

    Yamamoto, T.; Suzuki, M.; Ando, Y.

    2012-07-01

    After the severe core damage of Fukushima Dai-Ichi Nuclear Power Station, radioactive material leaked from the reactor buildings. As part of monitoring of radioactivity in the site, measurements of radioactivity in soils at three fixed points have been performed for {sup 134}Cs and {sup 137}Cs with gamma-ray spectrometry and for Pu, Pu, and {sup 240}Pu with {alpha}-ray spectrometry. Correlations of radioactivity ratios of {sup 134}Cs to {sup 137}Cs, and {sup 238}Pu to the sum of {sup 239}Pu and {sup 240}Pu with fuel burnup were studied by using theoretical burnup calculations and measurements on isotopic inventories, and compared with the Cs and Pu radioactivity rations in the soils. The comparison indicated that the burnup of the fuel sourcing the radioactivity was from 18 to 38 GWd/t, which corresponded to that of the fuel in the highest power and, therefore, the highest decay heat in operating high-burnup fueled BWR cores. (authors)

  3. 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.

  4. 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

  5. Testing of uranium nitride fuel in T-111 cladding at 1200 K cladding temperature

    NASA Technical Reports Server (NTRS)

    Rohal, R. G.; Tambling, T. N.; Smith, R. L.

    1973-01-01

    Two groups of six fuel pins each were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a tantalum alloy clad. The first group of fuel pins was irradiated for 1500 hours to a maximum burnup of 0.7-atom-percent uranium. The second group of fuel pins was irradiated for about 3000 hours to a maximum burnup of 1.0-atom-percent uranium. The average clad surface temperature during irradiation of both groups of fuel pins was approximately 1200 K. The postirradiation examination revealed the following: no clad failures or fuel swelling occurred; less than 1 percent of the fission gases escaped from the fuel; and the clad of the first group of fuel pins experienced clad embrittlement whereas the second group, which had modified assembly and fabrication procedures to minimize contamination, had a ductile clad after irradiation.

  6. 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.

  7. 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

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. Initial Cladding Condition

    SciTech Connect

    E. Siegmann

    2000-08-22

    The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M&O 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis will be used in

  14. 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

  15. Fabrication of a Porous Fiber Cladding Material Using Microsphere Templating for Improved Response Time with Fiber Optic Sensor Arrays

    PubMed Central

    Henning, Paul E.; Rigo, M. Veronica; Geissinger, Peter

    2012-01-01

    A highly porous optical-fiber cladding was developed for evanescent-wave fiber sensors, which contains sensor molecules, maintains guiding conditions in the optical fiber, and is suitable for sensing in aqueous environments. To make the cladding material (a poly(ethylene) glycol diacrylate (PEGDA) polymer) highly porous, a microsphere templating strategy was employed. The resulting pore network increases transport of the target analyte to the sensor molecules located in the cladding, which improves the sensor response time. This was demonstrated using fluorescein-based pH sensor molecules, which were covalently attached to the cladding material. Scanning electron microscopy was used to examine the structure of the templated polymer and the large network of interconnected pores. Fluorescence measurements showed a tenfold improvement in the response time for the templated polymer and a reliable pH response over a pH range of five to nine with an estimated accuracy of 0.08 pH units. PMID:22654644

  16. 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.

  17. 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).

  18. PWR cores with silicon carbide cladding

    SciTech Connect

    Dobisesky, J. P.; Carpenter, D.; Pilat, E.; Kazimi, M. S.

    2012-07-01

    The feasibility of using silicon carbide rather than Zircaloy cladding, to reach higher power levels and higher discharge burnups in PWRs has been evaluated. A preliminary fuel design using fuel rods with the same dimensions as in the Westinghouse Robust Fuel Assembly but with fuel pellets having 10 vol% central void has been adopted to mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide and to the persistence of the open clad-pellet gap over most of the fuel life. With this modified fuel design, it is possible to achieve 18 month cycles that meet present-day operating constraints on peaking factor, boron concentration, reactivity coefficients and shutdown margin, while allowing batch average discharge burnups up to 80 MWD/kgU and peak rod burnups up to 100 MWD/kgU. Power uprates of 10% and possibly 20% also appear feasible. For non-uprated cores, the silicon carbide-clad fuel has a clear advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a savings in enriched uranium. Control rod configuration modifications may be required to meet the shutdown margin criterion for the 20% up-rate. Silicon carbide's ability to sustain higher burnups than Zircaloy also allows the design of a licensable two year cycle with only 96 fresh assemblies, avoiding the enriched uranium penalty incurred with use of larger batch sizes due to their excessive leakage. (authors)

  19. 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.

  20. Evaluation of Corrosion of Aluminum Based Reactor Fuel Cladding Materials During Dry Storage

    SciTech Connect

    Peacock, H.B. Jr.

    1999-10-21

    This report provides an evaluation of the corrosion behavior of aluminum cladding alloys and aluminum-uranium alloys at conditions relevant to dry storage. The details of the corrosion program are described and the results to date are discussed.

  1. Development of new ferritic steels as cladding material for metallic fuel fast breeder reactor

    NASA Astrophysics Data System (ADS)

    Tokiwai, Moriyasu; Horie, Masaaki; Kako, Kenji; Fujiwara, Masayuki

    1993-09-01

    The excellent thermal, chemical and neutronic properties of metallic fuel (U-Pu-Zr alloy) will lead to drastic improvements in fast reactor safety and the related fuel cycle economy. Some new high molybdenum 12Cr ferritic stainless steel candidate cladding alloys have been designed to achieve the mechanical properties required for high performance metallic fuel elements. These candidate claddings were irradiated by ion bombardment and tested to determine their strength and creep rupture properties. A 12Cr-8Mo and a 12Cr-8Mo-0.1Y 2O 3 steel were fabricated into cladding via a powder metallurgy process and by a mechanical alloying process, respectively. These claddings had two and three times the creep rupture strength (pressurized at 650°C for 10000 h) of a conventional 12Cr ferritic steel (HT-9). These two steels also showed no void formation up to 350 dpa by Ni 3+ irradiation. A zircaloy-2 lined steel cladding tube has also been fabricated for the purpose of reducing fuel-cladding interdiffusion and chemical interaction.

  2. High temperature oxidation of fuel cladding candidate materials in steam-hydrogen environments

    NASA Astrophysics Data System (ADS)

    Pint, B. A.; Terrani, K. A.; Brady, M. P.; Cheng, T.; Keiser, J. R.

    2013-09-01

    Alternative fuel cladding materials to Zr alloys are being investigated for enhanced accident tolerance, which specifically involves oxidation resistance to steam or steam-H2 environments at ⩾1200 °C for short times. Based on a comparison of a range of commercial and model alloys, conventional austenitic steels do not have sufficient oxidation resistance with only ˜18Cr-10Ni. Higher alloyed type 310 stainless steel is protective but Ni is not a desirable alloy addition for this application. Results at 1350 °C indicated that FeCrAl alloys and CVD SiC remain oxidation resistant in steam. At 1200 °C, high (⩾25% Cr) ferritic alloys appear to be good candidates for this application. Higher pressures (up to 20.7 bar) and H2 additions appeared to have a limited effect on the oxidation behavior of the most oxidation resistant alloys, but higher pressures accelerated the maximum metal loss for less oxidation resistant steels and less metal loss was observed for type 317 L tubing in a H2-50%H2O environment at 10.3 bar compared to 100% H2O.

  3. 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.

  4. 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

  5. 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.

  6. 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.

  7. 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.

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

    NASA Astrophysics Data System (ADS)

    Gussev, M. N.; Byun, T. S.; Yamamoto, Y.; Maloy, S. A.; Terrani, K. A.

    2015-11-01

    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.

  9. 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)

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

    DOE PAGESBeta

    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 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

  11. 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.

  12. 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

  13. Repairing rabbit radial defects by combining bone marrow stroma stem cells with bone scaffold material comprising a core-cladding structure.

    PubMed

    Wu, H; Liu, G H; Wu, Q; Yu, B

    2015-10-05

    We prepared a bone scaffold material comprising a PLGA/β-TCP core and a Type I collagen cladding, and recombined it with bone marrow stroma stem cells (BMSCs) to evaluate its potential for use in bone tissue engineering by in vivo and in vitro experiments. PLGA/β-TCP without a cladding was used for comparison. The adherence rate of the BMSCs to the scaffold was determined by cell counting. Cell proliferation rate was determined by the 3-(4,5-dimethylthiazol-2-yl)-2,5-diphenyltetrazolium bromide method. The osteogenic capability was evaluated by alkaline phosphatase activity. The scaffold materials were recombined with the BMSCs and implanted into a large segmental rabbit radial defect model to evaluate defect repair. Osteogenesis was assessed in the scaffold materials by histological and double immunofluorescence labeling, etc. The adherence number, proliferation number, and alkaline phosphatase expression of the cells on the bone scaffold material with core-cladding structure were significantly higher than the corresponding values in the PLGA/β-TCP composite scaffold material (P < 0.05). An in vivo test indicated that the bone scaffold material with core-cladding structure completely degraded at the bone defect site and bone formation was completed. The rabbit large sentimental radial defect was successfully repaired. The degradation and osteogenesis rates matched well. The bone scaffold with core-cladding structure exhibited better osteogenic activity and capacity to repair a large segmental bone defect compared to the PLGA/β-TCP composite scaffold. The bone scaffold with core-cladding structure has excellent physical properties and biocompatibility. It is an ideal scaffold material for bone tissue engineering.

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

    DOE PAGESBeta

    Allen, Todd R.; Kaoumi, Djamel; Wharry, Janelle P.; Jiao, Zhijie; Topbasi, Cem; Kohnert, Aaron; Barnard, Leland; Certain, Alicia; Field, Kevin G.; Was, Gary S.; et al

    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

  15. 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.

  16. Appropriate burnup measurements for transportation burnup credit

    SciTech Connect

    Lancaster, D.; Fuentes, E.

    1997-04-01

    This paper addresses two of the measurement specifications used in analyzing spent fuel packages to gain burnup credit. The philosophy and calculation of rejection criteria and measurement accuracy are discussed. Any assembly for which the declared measured value and reactor record value deviate by more than 10% will be rejected. Measurement accuracy requirements are established for dependent and independent systems. The requirements have been tested and are achievable, ensuring safe operation without extra cost. 6 refs.

  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. 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.

  19. Conservative axial burnup distributions for actinide-only burnup credit

    SciTech Connect

    Kang, C.; Lancaster, D.

    1997-11-01

    Unlike the fresh fuel approach, which assumes the initial isotopic compositions for criticality analyses, any burnup credit methodology must address the proper treatment of axial burnup distributions. A straightforward way of treating a given axial burnup distribution is to segment the fuel assembly into multiple meshes and to model each burnup mesh with the corresponding isotopic compositions. Although this approach represents a significant increase in modeling efforts compared to the uniform average burnup approach, it can adequately determine the reactivity effect of the axial burnup distribution. A major consideration is what axial burnup distributions are appropriate for use in light of many possible distributions depending on core operating conditions and histories. This paper summarizes criticality analyses performed to determine conservative axial burnup distributions. The conservative axial burnup distributions presented in this paper are included in the Topical Report on Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel Packages, Revision 1 submitted in May 1997 by the US Department of Energy (DOE) to the US Nuclear Regulatory Commission (NRC). When approved by NRC, the conservative axial burnup distributions may be used to model PWR spent nuclear fuel for the purpose of gaining actinide only burnup credit.

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

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

    SciTech Connect

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

    2007-07-01

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

  11. Dry Storage Demonstration for High-Burnup Spent Nuclear Fuel-Feasibility Study

    SciTech Connect

    McKinnon, Mikal A. ); Cunningham, Mitchel E. )

    2003-09-09

    Initially, casks for dry storage of spent fuel were licensed for assembly-average burnup of about 35 GWd/MTU. Over the last two decades, the discharge burnup of fuel has increased steadily and now exceeds 45 GWd/MTU. With spent fuel burnups approaching the licensing limits (peak rod burnup of 62 GWd/MTU for pressurized water reactor fuel) and some lead test assemblies being burned beyond this limit, a need for a confirmatory dry storage demonstration program was first identified after the publication in May 1999 of the U.S. Nuclear Regulatory Commissions (NRC) Interim Staff Guidance 11 (ISG-11). With the publication in July 2002 of the second revision of ISG-11, the desirability for such a program further increased to obtain confirmatory data about the potential changes in cladding mechanical properties induced by dry storage, which would have implications to the transportation, handling, and disposal of high-burnup spent fuel. While dry storage licenses have kept pace with reactor discharge burnups, transportation licenses have not and are considered on a case by case basis. Therefore, this feasibility study was performed to examine the options available for conducting a confirmatory experimental program supporting the dry storage, transportation, and disposal of spent nuclear fuel with burnups well in excess of 45 GWd/MTU.

  12. A Probabilistic-Micro-mechanical Methodology for Assessing Zirconium Alloy Cladding Failure

    SciTech Connect

    Pan, Y.M.; Chan, K.S.; Riha, D.S.

    2007-07-01

    Cladding failure of fuel rods caused by hydride-induced embrittlement is a reliability concern for spent nuclear fuel after extended burnup. Uncertainties in the cladding temperature, cladding stress, oxide layer thickness, and the critical stress value for hydride reorientation preclude an assessment of the cladding failure risk. A set of micro-mechanical models for treating oxide cracking, blister cracking, delayed hydride cracking, and cladding fracture was developed and incorporated in a computer model. Results obtained from the preliminary model calculations indicate that at temperatures below a critical temperature of 318.5 deg. C [605.3 deg. F], the time to failure by delayed hydride cracking in Zr-2.5%Nb decreased with increasing cladding temperature. The overall goal of this project is to develop a probabilistic-micro-mechanical methodology for assessing the probability of hydride-induced failure in Zircaloy cladding and thereby establish performance criteria. (authors)

  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. 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.

  15. Results of a multi-year study aimed at the resolution of regulatory issues related to the storage and transportation of high-burnup spent fuel

    SciTech Connect

    Rashid, Joseph; Machiels, Albert

    2007-07-01

    Available in abstract form only. Full text of publication follows: Finding timely resolutions of current regulatory issues related to spent fuel storage and transportation is one of the most important priorities for both industry and regulators. Spent fuel pools at many US power plants have either reached or are rapidly approaching full capacity, a condition made worse by the longer cooling time required for high burnup (>45 GWd/MTU) spent fuel compared to lower burnup fuel for which most spent fuel pools were designed to accommodate. Consequently, the need for the transfer of spent fuel to dry storage, with eventual transportation to off-site interim storage facilities or to a permanent repository, has brought with it the need to cope with a number of regulatory issues that require significant lead time to resolve. In anticipation of this need, EPRI has, over the past several years, implemented a number of research programs, which include: (a) assessing the criticality risks during transportation; (b) evaluating the option of moderator exclusion; (c) participating in data gathering for implementation of full burnup credit; (d) evaluating the potential for fuel reconfiguration during transportation accidents; and (e) assessing the impact of fuel reconfiguration on spent fuel reactivity levels. The criteria by which the results of this program may be evaluated are the regulations contained in 10 CFR Parts 71 and 72 as well as in Standard Review Plans and Interim Staff Guidance (ISG) documents such as ISG-11, ISG-8 and ISG-19. Of these research programs, the fuel reconfiguration issue is the most complex because it requires long lead-time to develop the necessary material behavior models and analysis methods. To this end, the paper describes the results of EPRI's multi-year research program, with emphasis on the various phenomena that govern cladding thermo-mechanical behavior from the onset of placing spent fuel in dry storage casks to the consequences of

  16. Development of modified MDA (M-MDA), PWR fuel cladding tube for high duty operation in future

    SciTech Connect

    Watanabe, Seiichi; Kido, Toshiya; Arakawa, Yasushi

    2007-07-01

    A new cladding material of M-MDA has been developed in order to prepare for a strong growing demand for advanced fuel which can maintain its integrity even under high duties due to more efficient operation such as higher burnup, higher LHR, and longer operation cycle which will contribute the suppression of environmental burdens like CO{sub 2} emission. The main aim of M-MDA is to have excellent corrosion resistance while the other properties are inherited from MDA, which has been adopted to the step 2 fuel, instead of Zry-4, of Japanese PWR plant whose upper limit of assembly discharged burnup is 55 MWd/kgU. And we could confirm that the main aim of M-MDA was achieved by means of out-of-pile tests. In order to confirm improvement of corrosion resistance of M-MDA in the actual operation, irradiation test of M-MDA in the commercial reactor of Vandellos II is ongoing. The latest results of on-site examination after every end of cycle showed that oxide thickness of M-MDA-SR was much smaller than that of MDA at rod discharged burnup of approximately 60 MWd/kgU. The final irradiation cycle was completed on April 2007 and then we will obtain corrosion data of M-MDA over 70 MWd/kgU. M-MDA is a candidate alloy for advanced fuel under higher duty usage. (authors)

  17. 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.

  18. 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

    2013-05-01

    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. PMID:23669955

  19. 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]).

  20. 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)

  1. 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).

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Fracture behavior and microstructural characteristics of irradiated Zircaloy cladding

    SciTech Connect

    Chung, H.M.; Yaggee, F.L.; Kassner, T.F.

    1985-06-01

    Zircaloy cladding tube specimens from commercial power reactor fuel assemblies (burnup >22 MWd/kgU) have been deformed to fracture at 325/sup 0/C by either the internal gas-pressurization or the expanding-mandrel technique in a helium or argon environment containing no fission product species (e.g., I, Cs, or Cd). The fracture surfaces of 11 irradiated specimens fractured by internal gas pressurization were examined by scanning electron microscopy, and 7 specimens were found to contain various degrees of the pseudocleavage feature that is characteristic of pellet-cladding interaction failures. Out of 10 test specimens fractured by expanding-mandrel loading, 5 were found to contain regions of pseudocleavage on the fracture surfaces. The specimens exhibited ''X-marks'' on the outer surface and brittle incipient cracks distributed on the inner surface, which are also characteristic of pellet-cladding interaction failures.

  7. 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

  8. 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.

  9. 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.

  10. 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.

  11. Assessment of different fuel design options with SiC cladding for light water reactors

    SciTech Connect

    Xu, S.; Kazimi, M. S.

    2012-07-01

    It is very important to avoid high tensile stresses in SiC if it is to be used safely as the cladding material in LWRs. To achieve this goal, three fuel design options are examined here: fuel pellets with a 10% vol central void, adding 10% vol BeO in UO{sub 2} fuel and replacing the gas gap with Lead-Bismuth Eutectic (LBE) bond. For each case, a neutronic analysis is first applied using CASMO-4E to obtain the required enrichment to achieve a fuel average burnup of 50 MWd/kgU. The fuel performance code FRAPCON, specifically FRAPCON-3.3 and FRAPCON-EP, is modified and then used to simulate the effects of steady-state irradiation in each case for typical PWR peak rod conditions. Key parameters like fuel average temperature and plenum pressure are examined for the relative performance of each approach. The results show that all three options can improve the fuel performance by lowering the fuel temperature as well as the plenum pressure. Fuel pellets with a central void design offers the most advantages since the larger void volume in the fuel rod results in the lowest plenum pressure, which is the limiting condition for safe operation of the SiC cladding. (authors)

  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. The processing and evaluation of clad metals

    NASA Astrophysics Data System (ADS)

    Forster, James A.; Jha, Sunil; Amatruda, Andrew

    1993-06-01

    Clad metals are a specific form of composites in which the materials are arranged in a layered structure. Cold-roll bonding techniques are employed to produce more than 20,000 tonnes of clad metal laminates each year in the United States. This article is an introductory description of the processing steps in cold-roll bonding, the nature of the bond created in this process, and the methods used to evaluate the bond's strength.

  14. 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)

  15. A high-throughput investigation of Fe-Cr-Al as a novel high-temperature coating for nuclear cladding materials

    NASA Astrophysics Data System (ADS)

    Bunn, Jonathan Kenneth; Fang, Randy L.; Albing, Mark R.; Mehta, Apurva; Kramer, Matthew J.; Besser, Matthew F.; Hattrick-Simpers, Jason R.

    2015-07-01

    High-temperature alloy coatings that can resist oxidation are urgently needed as nuclear cladding materials to mitigate the danger of hydrogen explosions during meltdown. Here we apply a combination of computationally guided materials synthesis, high-throughput structural characterization and data analysis tools to investigate the feasibility of coatings from the Fe-Cr-Al alloy system. Composition-spread samples were synthesized to cover the region of the phase diagram previous bulk studies have identified as forming protective oxides. The metallurgical and oxide phase evolution were studied via in situ synchrotron glancing incidence x-ray diffraction at temperatures up to 690 K. A composition region with an Al concentration greater than 3.08 at%, and between 20.0 at% and 32.9 at% Cr showed the least overall oxide growth. Subsequently, a series of samples were deposited on stubs and their oxidation behavior at 1373 K was observed. The continued presence of a passivating oxide was confirmed in this region over a period of 6 h.

  16. A high-throughput investigation of Fe-Cr-Al as a novel high-temperature coating for nuclear cladding materials.

    PubMed

    Bunn, Jonathan Kenneth; Fang, Randy L; Albing, Mark R; Mehta, Apurva; Kramer, Matthew J; Besser, Matthew F; Hattrick-Simpers, Jason R

    2015-07-10

    High-temperature alloy coatings that can resist oxidation are urgently needed as nuclear cladding materials to mitigate the danger of hydrogen explosions during meltdown. Here we apply a combination of computationally guided materials synthesis, high-throughput structural characterization and data analysis tools to investigate the feasibility of coatings from the Fe–Cr–Al alloy system. Composition-spread samples were synthesized to cover the region of the phase diagram previous bulk studies have identified as forming protective oxides. The metallurgical and oxide phase evolution were studied via in situ synchrotron glancing incidence x-ray diffraction at temperatures up to 690 K. A composition region with an Al concentration greater than 3.08 at%, and between 20.0 at% and 32.9 at% Cr showed the least overall oxide growth. Subsequently, a series of samples were deposited on stubs and their oxidation behavior at 1373 K was observed. The continued presence of a passivating oxide was confirmed in this region over a period of 6 h. PMID:26086841

  17. 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.

  18. 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.

  19. Phenomena and Parameters Important to Burnup Credit

    SciTech Connect

    Parks, C.V.

    2001-01-10

    Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water-reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the US and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given.

  20. Overview of spent fuel burnup measurements

    SciTech Connect

    McLeod, N.B. Associates, Inc., Oakton, VA )

    1991-01-01

    The purpose of this paper is to provide a summary of the principal factors that influence burnup measurement accuracy, and the manner in which different combinations of these factors in any particular application influence the achievable accuracy in that application. The principal focus of the paper is on the passive measurement of various burnup indicators. This paper also provides a general background for four subsequent papers which discuss burnup measurement in two particular applications: for burnup credit in spent fuel shipping casks; and for safeguards purposes. This paper provides a basis for comparing these two applications in terms that directly relate to the measurement process.

  1. 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.

  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. 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.

  4. 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.

  5. 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.

  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. 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.

  8. 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.

  9. Hydride-related degradation of spent-fuel cladding under repository conditions

    SciTech Connect

    Chung, H. M.

    2000-04-03

    This report summarizes results of an analysis of hydride-related degradation of commercial spent-nuclear-fuel cladding under repository conditions. Based on applicable laboratory data on critical stress intensity obtained under isothermal conditions, occurrence of delayed hydride cracking from the inner-diameter side of cladding is concluded to be extremely unlikely. The key process for potential initiation of delayed hydride cracking at the outer-diameter side is long-term microstructural evolution near the localized regions of concentrated hydrides, i.e., nucleation, growth, and cracking of hydride blisters. Such locally concentrated hydrides are, however, limited to some high-burnup cladding only, and the potential for crack initiation and propagation at the outer-diameter side is expected to be insignificant for most spent fuels. Some degree of hydride reorientation could occur in high-burnup spent-fuel cladding. However, even if hydride reorientation occurs, accompanying stress-rupture failure in spent-fuel cladding is unlikely to occur.

  10. 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.

  11. 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

  12. ATR WG-MOX Fuel Pellet Burnup Measurement by Monte Carlo - Mass Spectrometric Method

    SciTech Connect

    Chang, Gray Sen I

    2002-10-01

    This paper presents a new method for calculating the burnup of nuclear reactor fuel, the MCWO-MS method, and describes its application to an experiment currently in progress to assess the suitability for use in light-water reactors of Mixed-OXide (MOX) fuel that contains plutonium derived from excess nuclear weapons material. To demonstrate that the available experience base with Reactor-Grade Mixed uranium-plutonium OXide (RGMOX) can be applied to Weapons-Grade (WG)-MOX in light water reactors, and to support potential licensing of MOX fuel made from weapons-grade plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory. Fuel burnup is an important parameter needed for fuel performance evaluation. For the irradiated MOX fuel’s Post-Irradiation Examination, the 148Nd method is used to measure the burnup. The fission product 148Nd is an ideal burnup indicator, when appropriate correction factors are applied. In the ATR test environment, the spectrum-dependent and burnup-dependent correction factors (see Section 5 for detailed discussion) can be substantial in high fuel burnup. The validated Monte Carlo depletion tool (MCWO) used in this study can provide a burnup-dependent correction factor for the reactor parameters, such as capture-to-fission ratios, isotopic concentrations and compositions, fission power, and spectrum in a straightforward fashion. Furthermore, the correlation curve generated by MCWO can be coupled with the 239Pu/Pu ratio measured by a Mass Spectrometer (in the new MCWO-MS method) to obtain a best-estimate MOX fuel burnup. A Monte Carlo - MCWO method can eliminate the generation of few-group cross sections. The MCWO depletion tool can analyze the detailed spatial and spectral self-shielding effects in UO2, WG-MOX, and reactor-grade mixed oxide (RG-MOX) fuel pins. The MCWO-MS tool only

  13. Value of burnup credit beyond actinides

    SciTech Connect

    Lancaster, D.; Fuentes, E.; Kang, Chi

    1997-12-01

    DOE has submitted a topical report to the NRC justifying burnup credit based only on actinide isotopes (U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241). When this topical report is approved, it will allow a great deal of the commercial spent nuclear fuel to be transported in significantly higher capacity casks. A cost savings estimate for shipping fuel in 32 assembly (burnup credit) casks as opposed to 24 assembly (non-burnup credit) casks was previously presented. Since that time, more detailed calculations have been performed using the methodology presented in the Actinide-Only Burnup Credit Topical Report. Loading curves for derated casks have been generated using actinide-only burnup credit and are presented in this paper. The estimates of cost savings due to burnup credit for shipping fuel utilizing 32, 30, 28, and 24 assembly casks where only the 24 assembly cask does not burnup credit have been created and are discussed. 4 refs., 2 figs.

  14. 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.

  15. Evaluation of Missing Pellet Surface Geometry on Cladding Stress Distribution and Magnitude

    SciTech Connect

    Capps, Nathan A.; Montgomery, Robert O.; Sunderland, Dion J.; Spencer, Ben; Pytel, Martin; Wirth, Brian D.

    2014-10-01

    Missing pellet surface (MPS) defects are local geometric defects that periodically occur in nuclear fuel pellets, usually as a result of the mishandling during the manufacturing process. The presences of these defects can lead to clad stress concentrations that are substantial enough to cause a through wall failure for certain conditions of power level, burnup, and power increase. Consequently, the impact of potential MPS defects has limited the rate of power increase or ramp rates in both PWR and BWR systems. Improved 3D MPS models that consider the effect of the MPS geometry can provide better understanding of the margins against PCMI clad failure. The Peregrine fuel performance code has been developed as a part the Consortium of Advanced Simulations of Light Water Reactors (CASL) to consider the inherently multi-physics and multi-dimensional mechanisms that control fuel behavior, including cladding failure by the presence of MPS defects. This paper presents an evaluation of the cladding stress concentrations as a function of MPS defect geometry. The results are the first step in a probabilistic approach to assess cladding failure during power maneuvers. This analysis provides insight into how varying pellet defect geometries affect the distribution of the cladding stress and fuel and cladding temperature and will be used to develop stress concentration factors for 2D and 3D models.

  16. Review of fuel/cladding eutectic formation in metallic SFR fuel pins

    SciTech Connect

    Denman, M.; Todreas, N.; Driscoll, M.

    2012-07-01

    Sodium-cooled Fast Reactors (SFRs) remain a strong contender amongst the Generation IV reactor concepts. Metallic fuel has been a primary fuel option for SFR designers in the US and was used extensively in the first generation of SFRs. One of the benefits of metallic fuel is its chemical compatibility with the coolant; unfortunately this compatibility does not extend to steel cladding at elevated temperatures. It has been known that uranium, plutonium, and rare earths diffuse with cladding constituents to form a low melting point fuel/cladding eutectic which acts to thin the cladding once the interfacial temperature rises above the system liquidus temperature. Since the 1960's, many experiments have been performed and published to evaluate the rate of fuel/cladding eutectic formation and the temperature above which melting will begin as a function of fuel/cladding interfacial temperature, time at temperature, fuel constituents (uranium, fissium or uranium (plutonium) zirconium), cladding type (stainless steel 316, stainless steel 306, D9 or HT9), beginning of life linear power, plutonium enrichment and burnup. The results of these tests, however, remain scattered across conference and journal papers spanning 50 years. The tests used to collect this data also varied in experimental procedure throughout the years. This paper will consolidate the experimental data into four groups of similar test conditions and expand upon the testing performed for each group in detail. A companion paper in PSA 2011 will discuss predictive correlations formulated from this database. (authors)

  17. 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

  18. Characterization of spent fuel approved testing material--ATM-104

    SciTech Connect

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

    1991-12-01

    The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. On microstructure and flexural strength of metal-ceramic composite cladding developed through microwave heating

    NASA Astrophysics Data System (ADS)

    Sharma, Apurbba Kumar; Gupta, Dheeraj

    2012-05-01

    A domestic multimode microwave applicator was used to develop carbide reinforced (tungsten-based) metal-matrix composite cladding on austenitic stainless steel substrate. Cladding was developed through microwave irradiation of the preplaced clad materials at 2.45 GHz for 420 s. Clads show metallurgical bonding with substrate by partial dilution of materials. Back scattered images of clad section confirm uniformly distributed reinforced particles in the metallic matrix. Presence of WC, W2C, NiSi, NiW and Co3W3C phases was detected in the clad. Flexural characteristics show two distinct load transitions attributable to deformations of the matrix and the reinforced particles. Clads fail at the upper transition load; further load is taken by the SS-316 substrate. Clads exhibit good stiffness and good adhesion with the substrate. Multi directional cracks were observed at the clad surface; on further loading, cracks get propagated into the clad thickness without getting peeled-off. Mechanism of clad development has been introduced.

  4. 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.

  5. 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).

  6. Burnup verification tests with the FORK measurement system-implementation for burnup credit

    SciTech Connect

    Ewing, R.I.

    1994-08-01

    Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. It was designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program and is well suited to verify burnup and cooling time records at commercial Pressurized Water Reactor (PWR) sites. This report deals with the application of the FORK system to burnup credit operations.

  7. 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.

  8. 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.

  9. Femtosecond writing of depressed cladding waveguides in strongly cumulative regime

    NASA Astrophysics Data System (ADS)

    Bukharin, Mikhail A.; Khudyakov, Dmitriy V.; Vartapetov, Sergey K.

    2015-05-01

    We proposed a novel approach for direct femtosecond inscription of waveguides. It consisted in formation of cladding with reduced refractive index in fused silica. Depressed cladding was based on peripheral regions of individually written neighbored tracks, which should be inscribed in strongly cumulative regime. It was shown, that due to shot time interval between femtosecond laser pulses and relatively slow thermal diffusion, the exposed focal region surrounds by significantly wide cladding with reduced refracted index. Based on proposed approach we demonstrated depressed cladding waveguide inscription in fused silica using emission directly from commercially available femtosecond oscillator without correcting optical systems and second harmonic generation. It was shown, that the new approach provides formation of easily adjustable single mode waveguides with desired mode field diameter. Such depressed cladding waveguides exploit both advantages of fused silica material and depressed cladding geometry. We also verified our suggestion by experiment and inscribed depressed cladding waveguides with two different mode field diameters at similar femtosecond pulse characteristics. The obtained structures provided low propagation losses and good coupling with Gaussian mode. The waveguides supported propagation of both polarizations with nearly identical characteristics. Obtained experimental results were in good agreement with numerical simulation.

  10. MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis

    SciTech Connect

    Gray S Chang

    2005-04-01

    The UNIX BASH (Bourne Again Shell) script MCWO has been developed at the Idaho National Engineering and Environment Laboratory (INEEL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN2. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. MCWO can handle a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) powers, and irradiation time intervals. The program processes input from the user that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN2, and data process module calculations are then output successively as the code runs. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN2 back to MCNP in a repeated, cyclic fashion. The basic requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with ORIGEN2 and other calculations are performed by UNIX BASH script MCWO. This paper presents the MCWO-calculated results of the RERTR-1 and -2 , and the Weapons-Grade Mixed Oxiide fuel (Wg-MOX) fuel experiments in ATR and compares the MCWO-calculated results with the measured data.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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

  18. 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.

  19. 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)

  20. Three-Layer Zn/Al/Zn Clad Solder for Die Attachment

    NASA Astrophysics Data System (ADS)

    Yamaguchi, T.; Ikeda, O.; Oda, Y.; Hata, S.; Kuroki, K.; Kuroda, H.; Hirose, A.

    2015-02-01

    Three-layer Zn/Al/Zn clad solders have been developed for high-temperature die attachment. The clad structure is used to improve the wettability and bondability of Zn-Al eutectic solder by preventing oxidation of the Al. The materials were produced by clad-rolling Zn and Al strips. TEM observations revealed that the Zn/Al clad interface was metallurgically bonded and that the Al oxide was almost entirely removed. The melting behavior of Zn/Al/Zn clad solder was examined. Eutectic melting began at the Zn/Al clad interface at 382°C, and all of the material melted within approximately 10 s. Unlike conventional Zn-Al solders, Zn/Al/Zn clad solders were successfully bonded without flux. The shear strength of a Zn/Al/Zn clad solder joint was three times that of a Pb-based solder joint. The bondability of Zn/Al/Zn clad solder was superior because the Al oxide films, which prevent bonding between chip and substrate, were fragmented by clad-rolling, and the outer Zn layers prevented Al oxidation during the bonding process.

  1. 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.

  2. Explosion Clad for Upstream Oil and Gas Equipment

    NASA Astrophysics Data System (ADS)

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

    2011-01-01

    Today's upstream oil and gas facilities frequently involve the combination of high pressures, high temperatures, and highly corrosive environments, requiring equipment that is thick wall, corrosion resistant, and cost effective. When significant concentrations of CO2 and/or H2S and/or chlorides are present, corrosion resistant alloys (CRA) can become the material of choice for separator equipment, piping, related components, and line pipe. They can provide reliable resistance to both corrosion and hydrogen embrittlement. For these applications, the more commonly used CRA's are 316L, 317L and duplex stainless steels, alloy 825 and alloy 625, dependent upon the application and the severity of the environment. Titanium is also an exceptional choice from the technical perspective, but is less commonly used except for heat exchangers. Explosion clad offers significant savings by providing a relatively thin corrosion resistant alloy on the surface metallurgically bonded to a thick, lower cost, steel substrate for the pressure containment. Developed and industrialized in the 1960's the explosion cladding technology can be used for cladding the more commonly used nickel based and stainless steel CRA's as well as titanium. It has many years of proven experience as a reliable and highly robust clad manufacturing process. The unique cold welding characteristics of explosion cladding reduce problems of alloy sensitization and dissimilar metal incompatibility. Explosion clad materials have been used extensively in both upstream and downstream oil, gas and petrochemical facilities for well over 40 years. The explosion clad equipment has demonstrated excellent resistance to corrosion, embrittlement and disbonding. Factors critical to insure reliable clad manufacture and equipment design and fabrication are addressed.

  3. 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.

  4. 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

  5. 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.

  6. An Automated, Multi-Step Monte Carlo Burnup Code System.

    2003-07-14

    Version 02 MONTEBURNS Version 2 calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. MONTEBURNS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with a radioactive decay and burnup code. Highlights on changes to Version 2 are listed in the transmittal letter. Along with other minor improvements in MONTEBURNS Version 2,more » the option was added to use CINDER90 instead of ORIGEN2 as the depletion/decay part of the system. CINDER90 is a multi-group depletion code developed at LANL and is not currently available from RSICC. This MONTEBURNS release was tested with various combinations of CCC-715/MCNPX 2.4.0, CCC-710/MCNP5, CCC-700/MCNP4C, CCC-371/ORIGEN2.2, ORIGEN2.1 and CINDER90. Perl is required software and is not included in this distribution. MCNP, ORIGEN2, and CINDER90 are not included.« less

  7. An Automated, Multi-Step Monte Carlo Burnup Code System.

    SciTech Connect

    TRELLUE, HOLLY R.

    2003-07-14

    Version 02 MONTEBURNS Version 2 calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. MONTEBURNS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with a radioactive decay and burnup code. Highlights on changes to Version 2 are listed in the transmittal letter. Along with other minor improvements in MONTEBURNS Version 2, the option was added to use CINDER90 instead of ORIGEN2 as the depletion/decay part of the system. CINDER90 is a multi-group depletion code developed at LANL and is not currently available from RSICC. This MONTEBURNS release was tested with various combinations of CCC-715/MCNPX 2.4.0, CCC-710/MCNP5, CCC-700/MCNP4C, CCC-371/ORIGEN2.2, ORIGEN2.1 and CINDER90. Perl is required software and is not included in this distribution. MCNP, ORIGEN2, and CINDER90 are not included.

  8. 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.

  9. 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.

  10. 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.

  11. Real-time laser cladding control with variable spot size

    NASA Astrophysics Data System (ADS)

    Arias, J. L.; Montealegre, M. A.; Vidal, F.; Rodríguez, J.; Mann, S.; Abels, P.; Motmans, F.

    2014-03-01

    Laser cladding processing has been used in different industries to improve the surface properties or to reconstruct damaged pieces. In order to cover areas considerably larger than the diameter of the laser beam, successive partially overlapping tracks are deposited. With no control over the process variables this conduces to an increase of the temperature, which could decrease mechanical properties of the laser cladded material. Commonly, the process is monitored and controlled by a PC using cameras, but this control suffers from a lack of speed caused by the image processing step. The aim of this work is to design and develop a FPGA-based laser cladding control system. This system is intended to modify the laser beam power according to the melt pool width, which is measured using a CMOS camera. All the control and monitoring tasks are carried out by a FPGA, taking advantage of its abundance of resources and speed of operation. The robustness of the image processing algorithm is assessed, as well as the control system performance. Laser power is decreased as substrate temperature increases, thus maintaining a constant clad width. This FPGA-based control system is integrated in an adaptive laser cladding system, which also includes an adaptive optical system that will control the laser focus distance on the fly. The whole system will constitute an efficient instrument for part repair with complex geometries and coating selective surfaces. This will be a significant step forward into the total industrial implementation of an automated industrial laser cladding process.

  12. Radiographic Inspection of Fueled Clads

    SciTech Connect

    Timothy J. Roney; Karen M. Wendt

    2005-04-01

    Five general purpose heat source (GPHS) fueled clads were radiographically inspected at the Idaho National Laboratory (INL). The girth weld region of each clad had previously passed visual examination, ring gauge test, and leak test but showed “positive” indications on the ultrasonic (UT) test. Positive ultrasonic indications are allowable under certain weld conditions; radiographic inspection provides a secondary nonintrusive means of clad inspection and may confirm allowable anomalies from the UT inspection. All the positive UT indications were found to exhibit allowable weld shield fusion or mismatch conditions. No indication of void defects was found. One additional clad (FCO371) was deemed unacceptable for radiographic inspection due to an unknown black substance that obscured the angular origin on the weld so that the angular offset to the UT indication could not be found.

  13. 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.

  14. Characterization of spent fuel approved testing material---ATM-105

    SciTech Connect

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

    1991-12-01

    The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

  15. 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

  16. Spent fuel dissolution rates as a function of burnup and water chemistry

    SciTech Connect

    Gray, W.J.

    1998-06-01

    To help provide a source term for performance-assessment calculations, dissolution studies on light-water-reactor (LWR) spent fuel have been conducted over the past few years at Pacific Northwest National Laboratory in support of the Yucca Mountain Site Characterization Project. This report describes that work for fiscal years 1996 through mid-1998 and includes summaries of some results from previous years for completeness. The following conclusions were based on the results of various flowthrough dissolution rate tests and on tests designed to measure the inventories of {sup 129}I located within the fuel/cladding gap region of different spent fuels: (1) Spent fuels with burnups in the range 30 to 50 MWd/kgM all dissolved at about the same rate over the conditions tested. To help determine whether the lack of burnup dependence extends to higher and lower values, tests are in progress or planned for spent fuels with burnups of 13 and {approximately} 65 MWd/kgM. (2) Oxidation of spent fuel up to the U{sub 4}O{sub 9+x} stage does not have a large effect on intrinsic dissolution rates. However, this degree of oxidation could increase the dissolution rates of relatively intact fuel by opening the grain boundaries, thereby increasing the effective surface area that is available for contact by water. From a disposal viewpoint, this is a potentially more important consideration than the effect on intrinsic rates. (3) The gap inventories of {sup 129}I were found to be smaller than the fission gas release (FGR) for the same fuel rod with the exception of the rod with the highest FGR. Several additional fuels would have to be tested to determine whether a generalized relationship exists between FGR and {sup 129}I gap inventory for US LWR fuels.

  17. 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.

  18. Thermo-mechanical analysis of LWR SiC/SiC composite cladding

    NASA Astrophysics Data System (ADS)

    Ben-Belgacem, M.; Richet, V.; Terrani, K. A.; Katoh, Y.; Snead, L. L.

    2014-04-01

    A dedicated framework for thermo-mechanical analysis of the in-pile performance of SiC/SiC composite fuel cladding concepts in LWRs has been developed. This analysis framework focuses on cladding and omits any fuel-cladding interaction and fuel behavior. Since radial expansion of the cladding occurs early in life for these ceramic structures, fuel-cladding contact is expected to be delayed or eliminated and therefore it is not considered in this analysis. The analysis inputs recent out-of-pile and in-pile materials property data and phenomenological understanding of material evolution under neutron irradiation for nuclear-grade SiC/SiC composites to provide a best-estimate analysis. The analysis provides insight into the concept design and feasibility of SiC/SiC composite cladding concepts that exhibit significantly different behavior than metallic cladding structures. In particular, absence of any tangible creep (thermal or irradiation) coupled with a large and temperature-gradient-driven irradiation swelling strain gradient across the cladding, drive development of large stresses across the cladding thickness. The resulting analysis indicates that significant stresses develop after a modest neutron dose (∼1 dpa) and a pronounced variation across the cladding thickness exists and is opposite to that observed for metallic cladding structures where swelling or growth strains are either negligible (with small temperature dependence) or absent. Following this thermo-mechanical analysis, a best-estimate and parametric examination of SiC/SiC fuel rod cladding structures has been performed using appropriate Weibull statistics to prescribe basic design guidelines and to begin to define a probable design space.

  19. Spatial mode-selective waveguide with hyperbolic cladding.

    PubMed

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

    2016-09-15

    Hyperbolic metamaterials (HMMs) are anisotropic materials with a permittivity tensor that has both positive and negative eigenvalues. Here we report that by using a type II HMM as a cladding material, a waveguide that only supports higher-order modes can be achieved, while the lower-order modes become leaky and are absorbed in the HMM cladding. This counter-intuitive property can lead to novel application in optical communications and photonic integrated circuits. The loss in our HMM insulator-HMM (HIH) waveguide is smaller than that of similar guided modes in a metal-insulator-metal (MIM) waveguide. PMID:27628378

  20. CHARACTERIZATION OF HYDROGEN CONTENT IN ZIRCALOY-4 NUCLEAR FUEL CLADDING

    SciTech Connect

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

    2010-02-22

    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.

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

    SciTech Connect

    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.

  2. 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.

  3. 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.

  4. Design of intrinsically single-mode double clad crystalline fiber waveguides for high power lasers

    NASA Astrophysics Data System (ADS)

    Li, Da; Hong, Pengda; Meissner, Stephanie K.; Meissner, Helmuth E.

    2016-03-01

    Recently, double-clad crystalline fiber waveguides (CFWs), consisting of single crystalline or ceramic RE3+:YAG cores of square cross section and inner claddings of either undoped or laser-inactive-ion-doped YAG and outer claddings of sapphire, have been successfully demonstrated. These waveguides, manufactured by an Adhesive-Free Bonding (AFB®) technique, can be precisely engineered and fabricated with predictable beam propagation behavior. In this work, with high power laser designs in mind, minimum thicknesses for inner cladding are derived for different core cross sections and refractive index differences between the core and inner cladding and sapphire as outer cladding material for common laser core dopants such as Nd3+, Yb3+, Er3+, Tm3+ and Ho3+. All designs are intended to use high NA high power laser diode pumping to obtain high power intrinsically single transverse mode laser output. The obtained data are applicable to any crystalline fiber waveguide design, regardless of fabrication technique. As an example, a CFW with 40 μm × 40 μm 4% Tm:YAG core, 5% Yb:YAG inner cladding, and sapphire outer cladding was calculated to be intrinsically single transverse mode, with the minimum inner cladding width of 21.7 μm determined by the effective index technique [1].

  5. Structure-property correlations in nanostructured WC-12Co microwave clad

    NASA Astrophysics Data System (ADS)

    Zafar, Sunny; Sharma, Apurbba Kumar

    2016-05-01

    Nanostructured materials are known for enhanced properties as compared to their conventional counterparts. In the present work, microwave cladding technique was explored for depositing nanostructured WC-12Co clads on stainless steel substrates. Phase analysis of the WC-12Co microwave clads revealed the presence of Co6W6C, Co7W6, Co3W9C4, W2C and WC phases. The microstructure of the WC-12Co clads confirmed uniform distribution of nano-carbides in the form of clusters enclosed in the carbide network. Mechanical characterisation of the nanostructured clads was carried in terms of microhardness assessment and flexural strength measurement. The microwave induced clads exhibited excellent metallurgical bonding with the substrate and were free from interfacial cracks. The average microhardness of the developed clads was found in the order of 1760 ± 128HV. The flexural strength of the developed clads was observed to be 671 ± 28 MPa. The nanostructured clads exhibited good adhesion with the substrate without getting peeled-off under a load of 3.75 kN and a displacement of 3.72 mm.

  6. Study on the characteristics of an Er/Yb co-doped double cladding fiber laser

    NASA Astrophysics Data System (ADS)

    Qu, Zhou; Li, Qiushi; Yan, Mingliang

    2009-07-01

    An Er/Yb co-doped double cladding fiber laser pumped at 980 nm was optimized. The double-cladding fiber laser with whole fiber was obtained by end-pumping and utilizing fiber bragg grating as a resonator. The output power of laser was analyzed along the changes of output grating reflectance (L=10m) as well as the fiber length (R2=4%). Consequently, a fiber with 4 m Er / Yb co-doped double cladding was employed as gain medium while a fiber of which the reflectance was approximately 15% was used as output resonator mirror. Thereafter the technical indexes of EYDF(Er / Yb Double cladding Fiber) were measured. The absorption maximum of fiber core Er3+ was higher than 30dB/m and material gain maximum was observed at 1535nm. Moreover, the diameters of fiber core and inner cladding of double-cladding fiber grating were 6μm and 125μm respectively however the diameters of fiber core and inner cladding of Er/Yb co-doped double cladding fiber were 7μm and 130μm separately.According to the experimental data, a fiber laser with 4 m Er / Yb co-doped double cladding and launched maximum pump power of 3.4 W was set up. Proposed laser shows the maximum output power of 1.25 W and slope efficiency of 40%.

  7. 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.

  8. Thermal Behavior of Advanced UO{sub 2} Fuel at High Burnup

    SciTech Connect

    Muller, E.; Lambert, T.; Silberstein, K.; Therache, B.

    2007-07-01

    To improve the fuel performance, advanced UO{sub 2} products are developed to reduce significantly Pellet-Cladding Interaction and Fission Gas Release to increase high burnup safety margins on Light Water Reactors. To achieve the expected improvements, doping elements are currently used, to produce large grain viscoplastic UO{sub 2} fuel microstructures. In that scope, AREVA NP is conducting the qualification of a new UO{sub 2} fuel pellet obtained by optimum chromium oxide doping. To assess the fuel thermal performance, especially the fuel conductivity degradation with increasing burnup and also the kinetics of fission gas release under transient operating conditions, an instrumented in-pile experiment, called REMORA, has been developed by the CEA. One segment base irradiated for five cycles in a French EDF commercial PWR ({approx} 62 GWd/tM) was consequently re-instrumented with a fuel centerline thermocouple and an advanced pressure sensor. The design of this specific sensor is based on the counter-pressure principle and avoids any drift phenomenon due to nuclear irradiation. This rodlet was then irradiated in the GRIFFONOS rig of the Osiris experimental reactor at CEA Saclay. This device, located in the periphery of the core, is designed to perform test under conditions close to those prevailing in French PWR reactor. Power variations are carried out by translating the device relatively to the core. Self - powered neutron detectors are positioned in the loop in order to monitor the power the whole time of the irradiation. The re-irradiation of the REMORA experiment consisted of a stepped ramp to power in order to point out a potential degradation of the fuel thermal conductivity with increasing burnup. During the first part of the irradiation, most of the measurements were performed at low power in order to take into account the irradiation effects on UO{sub 2} thermal conductivity at high burnup in low range of temperature. The second part of the irradiation

  9. Weapons-Grade MOX Fuel Burnup Characteristics in Advanced Test Reactor Irradiation

    SciTech Connect

    G. S. Chang

    2006-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory (LANL) by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, 40, and 50 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). 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(MCWO). MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. The fuel burnup analyses presented in this study were performed using MCWO. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements and the irradiated WG-MOX post irradiation examination (PIE) data.

  10. Local Burn-Up Effects in the NBSR Fuel Element

    SciTech Connect

    Brown N. R.; Hanson A.; Diamond, D.

    2013-01-31

    This study addresses the over-prediction of local power when the burn-up distribution in each half-element of the NBSR is assumed to be uniform. A single-element model was utilized to quantify the impact of axial and plate-wise burn-up on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, including neutron energy spectrum, power distribution, and integral U-235 vector. The power distribution changes significantly when incorporating local burn-up effects and has lower power peaking relative to the uniform burn-up case. In the uniform burn-up case, the axial relative power peaking is over-predicted by as much as 59% in the HEU single-element and 46% in the LEU single-element with uniform burn-up. In the uniform burn-up case, the plate-wise power peaking is over-predicted by as much as 23% in the HEU single-element and 18% in the LEU single-element. The degree of over-prediction increases as a function of burn-up cycle, with the greatest over-prediction at the end of Cycle 8. The thermal flux peak is always in the mid-plane gap; this causes the local cumulative burn-up near the mid-plane gap to be significantly higher than the fuel element average. Uniform burn-up distribution throughout a half-element also causes a bias in fuel element reactivity worth, due primarily to the neutronic importance of the fissile inventory in the mid-plane gap region.

  11. Recycle of Zirconium from Used Nuclear Fuel Cladding: A Major Element of Waste Reduction

    SciTech Connect

    Collins, Emory D; DelCul, Guillermo D; Terekhov, Dmitri; Emmanuel, N. V.

    2011-01-01

    Feasibility tests were initiated to determine if the zirconium in commercial used nuclear fuel (UNF) cladding can be recovered in sufficient purity to permit re-use, and if the recovery process can be operated economically. Initial tests are being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Early results indicate that quantitative recovery can be accomplished and product contamination with alloy constituents can be controlled sufficiently to meet purification requirements. Future tests with actual radioactive UNF cladding are planned. The objective of current research is to determine the feasibility of recovery and recycle of zirconium from used fuel cladding wastes. Zircaloy cladding, which contains 98+% of hafnium-free zirconium, is the second largest mass, on average {approx}25 wt %, of the components in used U.S. light-water-reactor fuel assemblies. Therefore, recovery and recycle of the zirconium would enable a large reduction in geologic waste disposal for advanced fuel cycles. Current practice is to compact or grout the cladding waste and store it for subsequent disposal in a geologic repository. This paper describes results of initial tests being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Future tests with actual radioactive UNF cladding are planned.

  12. 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

  13. 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.

  14. 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.

  15. 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.

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

    NASA Astrophysics Data System (ADS)

    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 self-mated 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.

  17. SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

    SciTech Connect

    BSC

    2004-12-01

    Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier

  18. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    SciTech Connect

    Rudisill, T; John Mickalonis, J

    2006-09-27

    process by obstructing the removal of contamination. The thermal decomposition of this material is also undesirable if the cladding hulls are melted for volume reduction or to produce waste forms. Handling and disposal of the corrosive off-gas stream and ZrO{sub 2}-containing dross must be addressed. The stability of Zr{sup 4+} in the NHF{sub 4}/NH{sub 4}NO{sub 3} solution is also a concern. Precipitation of ammonium zirconium fluorides upon cooling of the dissolving solution was observed in the feasibility experiments. Precipitation of the solids was attributed to the high fluoride to Zr ratios used in the experiments. The solubility of Zr{sup 4+} in NH{sub 4}F solutions decreases as the free fluoride concentration increases. The removal of the ZrO{sub 2} layer from Zircaloy-4 coupons with HF showed a strong dependence on both the concentration and temperature. Very rapid dissolution of the oxide layer and significant amounts of metal was observed in experiments using HF concentrations {ge} 2.5 M. Treatment of the coupons using HF concentrations {le} 1.0 M was very effective in removing the oxide layer. The most effective conditions resulted in dissolution rates which were less than approximately 2 mg/cm{sup 2}-min. With dissolution rates in this range, uniform removal of the oxide layer was obtained and a minimal amount of Zircaloy metal was dissolved. Future HF dissolution studies should focus on the decontamination of actual spent fuel cladding hulls to determine if the treated hulls meet criteria for disposal as a LLW.

  19. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

    NASA Astrophysics Data System (ADS)

    Terrani, K. A.; Zinkle, S. J.; Snead, L. L.

    2014-05-01

    Application of advanced oxidation-resistant iron alloys as light water reactor fuel cladding is proposed. The motivations are based on specific limitations associated with zirconium alloys, currently used as fuel cladding, under design-basis and beyond-design-basis accident scenarios. Using a simplified methodology, gains in safety margins under severe accidents upon transition to advanced oxidation-resistant iron alloys as fuel cladding are showcased. Oxidation behavior, mechanical properties, and irradiation effects of advanced iron alloys are briefly reviewed and compared to zirconium alloys as well as historic austenitic stainless steel cladding materials. Neutronic characteristics of iron-alloy-clad fuel bundles are determined and fed into a simple economic model to estimate the impact on nuclear electricity production cost. Prior experience with steel cladding is combined with the current understanding of the mechanical properties and irradiation behavior of advanced iron alloys to identify a combination of cladding thickness reduction and fuel enrichment increase (∼0.5%) as an efficient route to offset any penalties in cycle length, due to higher neutron absorption in the iron alloy cladding, with modest impact on the economics.

  20. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    NASA Astrophysics Data System (ADS)

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

    2015-06-01

    FeCrAl, 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 this 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. 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.

  1. 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.

  2. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    DOE PAGESBeta

    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 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. Simulated transient behavior of HT9 cladding

    SciTech Connect

    Cannon, N.S.; Huang, F.H.; Hamilton, M.L.

    1988-09-01

    Simulated transient tests were performed on sections of HT9 fast- reactor fuel pin cladding irradiated to a fast fluence of nearly 16 /times/ 10/sup 22/ n/cm/sup 2/ at temperatures ranging from 370 to 620/degree/C. After removing fuel, these specimens were internally pressurized and heated at one of several constant rates (0.56, 5.6, or 110/degree/C/s) until specimen failure occurred. A slight reduction of strength was observed in irradiated cladding, particularly at 110/degree/C/s, when compared with transient results from unirradiated HT9 control specimens; however, this strength reduction did not correlate with either fluence or irradiation temperature. A small reduction of ductility was also observed for irradiated cladding failing at temperatures above 800/degree/C at the lower heating rates (0.56 or 5.6/degree/C/s); irradiated cladding was generally more ductile at 110/degree/C/s than unirradiated HT9 cladding. The HT9 cladding results were compared with similar transient data obtained previously from 20% Cold-Worked Type 316 Stainless Steel (316 SS) cladding. In the unirradiated state, this austenitic cladding is stronger and less ductile than HT9 cladding. However, the 316 SS cladding undergoes a significant loss of strength and ductility during irradiation when in contact with oxide fuel, by a mechanism labeled the fuel adjacency effect (FAE). The FAE is believed to be liquid metal embrittlement from fission products. The HT9 fuel pin cladding remained as strong or stronger than the 316 SS cladding when irradiated in contact with fuel, showing no evidence of the FAE up to the high fluences reported here. The ductility of the irradiated HT9 fuel pin cladding remained significantly greater than that of irradiated 316 SS cladding. 14 refs., 11 figs., 1 tab.

  4. Image processing applied to laser cladding process

    SciTech Connect

    Meriaudeau, F.; Truchetet, F.

    1996-12-31

    The laser cladding process, which consists of adding a melt powder to a substrate in order to improve or change the behavior of the material against corrosion, fatigue and so on, involves a lot of parameters. In order to perform good tracks some parameters need to be controlled during the process. The authors present here a low cost performance system using two CCD matrix cameras. One camera provides surface temperature measurements while the other gives information relative to the powder distribution or geometric characteristics of the tracks. The surface temperature (thanks to Beer Lambert`s law) enables one to detect variations in the mass feed rate. Using such a system the authors are able to detect fluctuation of 2 to 3g/min in the mass flow rate. The other camera gives them information related to the powder distribution, a simple algorithm applied to the data acquired from the CCD matrix camera allows them to see very weak fluctuations within both gaz flux (carriage or protection gaz). During the process, this camera is also used to perform geometric measurements. The height and the width of the track are obtained in real time and enable the operator to find information related to the process parameters such as the speed processing, the mass flow rate. The authors display the result provided by their system in order to enhance the efficiency of the laser cladding process. The conclusion is dedicated to a summary of the presented works and the expectations for the future.

  5. 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

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

    DOE PAGESBeta

    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 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

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

    DOE PAGESBeta

    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. 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

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. DECONTAMINATION OF ZIRCALOY CLADDING HULLS FROM SPENT NUCLEAR FUEL

    SciTech Connect

    Rudisill, T.

    2010-09-29

    The feasibility of decontaminating spent fuel cladding hulls using hydrofluoric acid (HF) was investigated as part of the Global Energy Nuclear Partnership (GNEP) Separations Campaign. The concentrations of the fission product and transuranic (TRU) isotopes in the decontaminated hulls were compared to the limits for determining the low level waste (LLW) classification in the United States (US). The {sup 90}Sr and {sup 137}Cs concentrations met the disposal criteria for a Class C LLW; although, in a number of experiments the criteria for disposal as a Class B LLW were met. The TRU concentration in the hulls generally exceeded the Class C LLW limit by at least an order of magnitude. The concentration decreased sharply as the initial 30-40 {micro}m of the cladding hull surface were removed. At depths beyond this point, the TRU activity remained relatively constant, well above the Class C limit. Reprocessing of spent nuclear fuel generates a cladding waste which would likely require disposal as a Greater than Class C LLW in the US. If the cladding hulls could be treated to remove a majority of the actinide and fission product contamination, the hulls could potentially meet acceptance criteria for disposal as a LLW or allow recycle of the Zr metal. Discard of the hulls as a LLW would result in significant cost savings compared to disposal as a Greater than Class C waste which currently has no disposition path. During fuel irradiation and reprocessing, radioactive materials are produced and deposited in the Zircaloy cladding. Due to short depths of penetration, the majority of the fission products and actinide elements are located in the ZrO{sub 2} layer which forms on the surface of the cladding during fuel irradiation. Therefore, if the oxide layer is removed, the majority of the contamination should also be removed. It is very difficult, if not impossible to remove all of the activity from spent fuel cladding since traces of U and Th in the unirradiated Zircaloy

  13. 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.

  14. Analysis of Cladding Residues from the Dissolution of Irradiated Dresden-1 Reactor Fuel

    SciTech Connect

    KESSINGER, GF.

    2004-05-27

    The primary goal of this work was to evaluate the efficacy of the chop-leach spent fuel dissolution process, with nitric acid dissolvent, for removing actinides and fission products from Zircaloy cladding to produce a cladding capable of meeting low-level waste (LLW) disposal criteria. Analysis of the cladding shows that actinides are present in the cladding at concentrations 50 to 400 times greater than the acceptable TRU limit in LLW. It appears that the nitric acid used for dissolution (initial concentration 4 M, with 10 M added as the dissolution proceeded) was inadequate for solubilizing the fuel meat. Scanning electron micrographs of the as-sampled cladding surface showed particles of material high in U on the surface of the cut samples, suggesting the fuel meat was not completely dissolved. If the cladding is to meet LLW disposal limits, it is likely that a more robust chemical treatment will be required to more completely digest the fuel meat. Based on the available analytical results, and the interpretation of those results, the following conclusions are drawn. The chop-leach method, as performed initially was inadequate for complete digestion of the fuel meat present in the Dresden-1 fuel samples studied. This failure of the dissolution process resulted in cladding samples that contained TRU actinides about 400 times more TRU content than is allowable in LLW. The inductively-coupled plasma emission spectroscopy (ICP-ES) method appears to overestimate the quantities of a variety of elements, such as Ag, Ce, Gd, La, and Sb. It is believed this phenomenon is related to spectral interferences from d-block and f-block elements. Future studies should be performed to determine the efficacy of leaching to chemically polish the cladding surface and remove undissolved fuel meat. Further analyses of cladding samples, to more thoroughly characterize the nature of the U- and TRU-bearing phases present in the cladding, are recommended.

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

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-01

    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.

  16. 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.

  17. 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.

  18. Triton burnup study in JT-60U

    SciTech Connect

    Nishitani, T.; Hoek, M.; Isobe, M.; Tobita, K.; Kusama, Y.; Harano, H.; Wurden, G.A.; Chrien, R.E.

    1995-10-01

    The behavior of 1 MeV tritons produced in the d(d,p)t reaction is important to predict the properties of D-T produced 3.5 MeV alphas because 1 MeV tritons and 3.5 MeV alphas have similar kinematic properties, such as Larmor radius and precession frequency. The confinement and slowing down of the fast tritons were investigated by measuring the 14 MeV and the 2.5 MeV neutron production rates. Here the time resolved triton burnup measurements have been performed using a new type 14 MeV neutron detector based on scintillating fibers, as part of a US-Japan tokamak collaboration. Loss of alpha particles due to toroidal ripple is one of the most important issues to be solved for a fusion reactor such as ITER. The authors investigated the toroidal ripple effect on the fast triton by analyzing the time history of the 14 MeV emission after NB turn-off.

  19. 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.

  20. 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.

  1. 77 FR 31834 - Clad Steel Plate from Japan: Final Results of the Expedited Third Sunset Review of the...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-30

    ... Five-Year (Sunset) Review, 77 FR 4995 (Feb. 1, 2012) (Notice of Initiation). On February 15, 2012, the... Value: Clad Steel Plate From Japan, 61 FR 21158, 21159 (May 9, 1996). ] Weighted- average Manufacturers... layer of cladding material (usually stainless steel or nickel) which is metallurgically bonded to a...

  2. Termination of plastic-clad fiber. [Plastic-clad silica

    SciTech Connect

    Nance, W.R.

    1982-03-01

    Optical waveguides are ideal in a nuclear weapon environment because of their resistance to electromagnetic interference. Of the fibers on today's market, plastic-clad silica (PCS) is the most radiation resistant and therfore the best choice. Because terminating PCS is complex, this paper attemps to address the major problems associated with these terminations including selecting the proper connector and optimizing the terminating procedures. The sources of losses in the connectors are summarized and typical loss values are given for four connectors which were tested.

  3. 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.

  4. 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).

  5. Coatings and claddings for the reduction of plasma contamination and surface erosion in fusion reactors

    SciTech Connect

    Kaminsky, M.

    1980-01-01

    For the successful operation of plasma devices and future fusion reactors it is necessary to control plasma impurity release and surface erosion. Effective methods to obtain such controls include the application of protective coatings to, and the use of clad materials for, certain first wall components. Major features of the development programs for coatings and claddings for fusion applications will be described together with an outline of the testing program. A discussion of some pertinent test results will be included.

  6. 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.

  7. HIP-clad products for the plastics industry

    NASA Astrophysics Data System (ADS)

    Bishop, Morley F.; Nickel, Clinton F.

    1999-07-01

    The production of plastics and plastics components requires equipment that can withstand severe wear and, in a high percentage of cases, wear and corrosion environments. There are two basic elements of plastic extrusion equipment: the barrels and the screws. Both must manifest similar properties, but since screw elements are less costly and easier to replace, they are usually designed to wear out first. Due to the high cost of wear/corrosion-resistance materials, the industry used clad (i.e., bimetallic) components. Barrel sections and screw segments are both produced as hot-isostatic press clad components using similar processes. There are any number of material combinations that are used and that are possible for the right application.

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

    SciTech Connect

    Maloy, S. A.; Toloczko, Mychailo B.; Cole, J. I.; Byun, Thak Sang

    2011-12-31

    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

  9. 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

  10. 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.

  11. 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.

  12. Inherent safety of minimum-burnup breed and burn reactors

    SciTech Connect

    Qvist, S.; Reenspan, E.

    2012-07-01

    Reactors that aim to sustain the breed and burn (B and B) mode of operation at minimum discharge burnup require excellent neutron economy, Minimum-burnup B and B cores are generally large and feature low neutron leakage probability and a hard neutron spectrum. While highly promising fuel cycles can be achieved with such designs, the very same features are pushing the limits of the core's ability to passively respond safely to unprotected accidents. Low leakage minimum-burnup sodium-cooled B and B cores have a large positive coolant void-worth and coolant temperature reactivity coefficient. In this study, the applicability of major approaches for fast reactor void-worth reduction is evaluated specifically for B and B cores. The design, shuffling scheme and performance of a new metallic-fueled, sodium-cooled minimum burnup B and B core, used as basis for the void-worth reduction analysis, is presented. The analysis shows that reactivity control systems based on passive {sup 6}Li injection during temperature excursions are the only option able to provide negative void-worth without significantly increasing the minimum burnup required for sustaining the B and B mode of operation. A new type of lithium expansion module (LEM) system was developed specifically for B and B cores and its effect on core performance is presented. (authors)

  13. 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.

  14. 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.

  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. Monte-Carlo Continuous Energy Burnup Code System.

    2007-08-31

    Version 00 MCB is a Monte Carlo Continuous Energy Burnup Code for a general-purpose use to calculate a nuclide density time evolution with burnup or decay. It includes eigenvalue calculations of critical and subcritical systems as well as neutron transport calculations in fixed source mode or k-code mode to obtain reaction rates and energy deposition that are necessary for burnup calculations. The MCB-1C patch file and data packages as distributed by the NEADB are verymore » well organized and are being made available through RSICC as received. The RSICC package includes the MCB-1C patch and MCB data libraries. Installation of MCB requires MCNP4C source code and utility programs, which are not included in this MCB distribution. They were provided with the now obsolete CCC-700/MCNP-4C package.« less

  17. Spent fuel burnup estimation by Cerenkov glow intensity measurement

    SciTech Connect

    Kuribara, Masayuki . Communication and Information Research Lab.)

    1994-10-01

    The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to the declared burnups for the tested spent fuel assemblies for cooling times ranging from 900--2.000 d.

  18. Issues related to criticality safety analysis for burnup credit applications

    SciTech Connect

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

    1995-12-01

    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. Parametric analyses are required to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models are evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. This paper discusses the results of studies to determine the effect of two important modeling assumptions on the criticality analysis of pressurized-water reactor (PWR) spent fuel: (1) the effect of assumed burnup history (i.e., specific power during and time-dependent variations in operational power) during depletion calculations, and (2) the effect of axial burnup distributions on the neutron multiplication factor calculated for a three-dimensional (3-D) conceptual cask design.

  19. 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.

  20. 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.

  1. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    SciTech Connect

    Gauld, I.C.

    2002-12-23

    Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103

  2. 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.

  3. High-burnup core design using minor actinide-containing metal fuel

    SciTech Connect

    Ohta, Hirokazu; Ogata, Takanari; Obara, T.

    2013-07-01

    A neutronic design study of metal fuel fast reactor (FR) cores is conducted on the basis of an innovative fuel design concept to achieve an extremely high burnup and realize an efficient fuel cycle system. Since it is expected that the burnup reactivity swing will become extremely large in an unprecedented high burnup core, minor actinides (MAs) from light water reactors (LWRs) are added to fresh fuel to improve the core internal conversion. Core neutronic analysis revealed that high burnups of about 200 MWd/kg for a small-scale core and about 300 MWd/kg for a large-scale core can be attained while suppressing the burnup reactivity swing to almost the same level as that of conventional cores with normal burnup. An actinide burnup analysis has shown that the MA consumption ratio is improved to about 60% and that the accumulated MAs originating from LWRs can be efficiently consumed by the high-burnup metal fuel FR. (authors)

  4. 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.

  5. Criticality validation for burnup credit using recycle Pu criticals

    SciTech Connect

    Fuentes, E.; Lancaster, D.

    1997-04-01

    A set of 23 additional critical experiments were analyzed to provide additional input to the criticality validation portion of spent fuel cask analysis. The results of this analyses were combined with the previously analyzed criticals to determine the upper safety limit on k{sub eff}. The combined set of criticals can be used used for criticality validation for burnup credit, and are better suited for the range of isotopics in spent nuclear fuels. A trend observed in the analysis was that the calculated k{sub eff} deviates from the criticals in the positive direction, implying that increased burnup results in increased safety margin. 6 refs., 2 figs., 1 tab.

  6. 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.

  7. Isotopic and criticality validation for actinide-only burnup credit

    SciTech Connect

    Fuentes, E.; Lancaster, D.; Rahimi, M.

    1997-07-01

    The techniques used for actinide-only burnup credit isotopic validation and criticality validation are presented and discussed. Trending analyses have been incorporated into both methodologies, requiring biases and uncertainties to be treated as a function of the trending parameters. The isotopic validation is demonstrated using the SAS2H module of SCALE 4.2, with the 27BURNUPLIB cross section library; correction factors are presented for each of the actinides in the burnup credit methodology. For the criticality validation, the demonstration is performed with the CSAS module of SCALE 4.2 and the 27BURNUPLIB, resulting in a validated upper safety limit.

  8. Fabrication of a tantalum-clad tungsten target for KENS

    NASA Astrophysics Data System (ADS)

    Kawai, Masayoshi; Kikuchi, Kenji; Kurishita, Hiroaki; Li, Jing-Feng; Furusaka, Michihiro

    2001-07-01

    Since the cold neutron source intensity of KENS (the spallation neutron source at High Energy Accelerator Research Organization) was decreased into about a third of the designed value because a cadmium liner at the cold neutron source deformed and obstructed the neutron beam line, the target-moderator-and-reflector assembly (TMRA) has been replaced by a new one aimed at improving the neutron performance and recovering the cold neutron source. The tantalum target has also been replaced by a tantalum-clad tungsten one. In order to bond the tantalum-clad with the tungsten block, a hot isostatic press (HIP) process was applied and optimized. It was found that gaseous interstitial impurity elements severely attacked tantalum and embrittled, and that the getter materials such as zirconium and tantalum were effective to reduce the embrittlement.

  9. 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.

  10. 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

  11. 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.

  12. 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

    2016-09-01

    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.

  13. GSGG edge cladding development: Final technical report

    SciTech Connect

    Izumitani, T.; Meissner, H.E.; Toratani, H.

    1986-11-15

    The objectives of this project have been: (1) Investigate the possibility of chemical etching of GSGG crystal slabs to obtain increased strength. (2) Design and construct a simplified mold assembly for casting cladding glass to the edges of crystal slabs of different dimensions. (3) Conduct casting experiments to evaluate the redesigned mold assembly and to determine stresses as function of thermal expansion coefficient of cladding glass. (4) Clad larger sizes of GGG slabs as they become available. These tasks have been achieved. Chemical etching of GSGG slabs does not appear possible with any other acid than H/sub 3/PO/sub 4/ at temperatures above 300/sup 0/C. A mold assembly has been constructed which allowed casting cladding glass around the edges of the largest GGG slabs available (10 x 20 x 160 mm) without causing breakage through the annealing step.

  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. 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.

  16. 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.

  17. 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)

  18. Non-Invasive Characterization of Burnup for PWR Spent Fuel Rods with Burnups > 80 GWd/t

    SciTech Connect

    Caruso, S.; Murphy, M.; Jatuff, F.; Chawla, R.

    2006-07-01

    High-resolution gamma spectroscopy has been employed for the measurement of {sup 134}Cs/{sup 137}Cs, {sup 154}Eu/{sup 137}Cs and {sup 134}Cs/{sup 154}Eu gamma intensity ratios from spent fuel with the purpose of deriving pin-averaged single-ratio burnup indicators for high and ultra-high burnups. Two UO{sub 2} pressurised water reactor (PWR) fuel rod segments with record burnup levels >80 GWd/t have been experimentally characterised. Additionally, pin cell depletion calculations have been performed for each sample with the deterministic code CASMO-4, using both its JEF2.2- and its ENDF/B-IV-based libraries, for three different descriptions of the fuel rod irradiation histories, in order to test the sensitivity of the results to neutron cross sections and to the depletion model employed. Measured and calculated ratios have then been compared. It is shown that the {sup 134}Cs/{sup 137}Cs ratio, frequently used as burnup monitor, is considerably less accurate for values exceeding 50 GWd/t; discrepancies of up to {approx}25% are found between measured and calculated values. The ratios built with the {sup 154}Eu concentration show much larger discrepancies, essentially because this isotope is rather poorly predicted as revealed by the use of different basic cross section data. (authors)

  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. Aluminum alloy clad fiber optic corrosion sensor

    NASA Astrophysics Data System (ADS)

    Rutherford, Paul S.; Ikegami, Roy; Shrader, John E.; Sherrer, David; Zabaronick, Noel; Zeakes, Jason S.; Murphy, Kent A.; Claus, Richard O.

    1997-06-01

    Life extension programs for military metallic aircraft are becoming increasingly important as defense budgets shrink and world economies realign themselves to an uncertain future. For existing military weapon systems, metallic corrosion damage costs as estimated $DOL8 billion per year. One approach to reducing this cost is to develop a reliable method to detect and monitor corrosion in hidden metallic structure with the use of corrosion sensors which would give an early indication of corrosion without significant disassembly, thereby reducing maintenance costs. This presentation describes the development, analysis, and testing of a fiber optic corrosion sensor developed jointly with the Virginia Polytechnic Fiber and Electro-Optics Research Center and sponsored by Wright Laboratory Materials Directorate. In the sensor which was researched, the normal cladding is removed in the sensor region, and replaced with aluminum alloy and allowed to corrode on coupons representative of C/KC-135 body structure in an ASTM B117 salt spray chamber and a Boeing developed Crevice Corrosion Cell. In this approach, the optical signal output of the sensor was originally designed to increase as corrosion takes place, however interaction with the corrosion byproducts yielded different results than anticipated. These test results to determine a correlation between the sensor output and the structural degradation due to corrosion are discussed.

  1. 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.

  2. Technical development of double-clad process for thin strip casting of carbon steel

    SciTech Connect

    Brown, H.L.; Forkel, C.E.; Knudson, D.L.

    1984-08-01

    This report documents the technical development for a patent disclosure of a double-clad process for the continuous casting of thin-strip carbon steel. The fundamental idea of the disclosure is to form a product strip by depositing molten steel between two, cooled, clad strips of the same material. The claimed benefits include: (a) the conservation of energy in steel making through the elimination of soaking pits and reheat cycles, and (b) an improved surface on both sides of the as-cast product such that it will be suitable for direct feed to a cold-reduction mill. However, the process as conceived is not necessarily limited to the casting of carbon steel, but may be also applied to other metals and alloys. The work is described under three headings as follows. Preliminary Considerations and Scoping Analysis presents the basic idea of the double-clad, thin-strip casting process; the energy conservation potential; scoping heat transfer calculations for the casting process; and independent review of this work. Thermal Analysis for Roller Configuration of Double-Clad Process, presents the development, results, and independent review of a finite-element thermal analysis for the casting process as originally conceived (using only chilled rollers in direct contact with the clad material of the product strip). Further Considerations for Belt Configuration of Double-Clad Process deals with a modified equipment design which interposes two product support belts, one on each side of the product, between the clad strip and the rollers. In addition to the process description, this section presents the preliminary mechanical calculations for the endless metal belts and the work scope and results for the computer model revision and thermal analysis for the modified concept.

  3. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    SciTech Connect

    Broadhead, B.L.

    1991-08-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications.

  4. High-strength clad current collector for silicon-based negative electrode in lithium ion battery

    NASA Astrophysics Data System (ADS)

    Kataoka, Riki; Oda, Yoshimitsu; Inoue, Ryouji; Kitta, Mitsunori; Kiyobayashi, Tetsu

    2016-01-01

    We develop a clad foil current collector with a high tensile strength that endures a large volume change in the active material during the charge and discharge, such as the Si-based materials. The nano-Si negative electrode with the clad current collector retains 76% of the initial capacity after 40 cycles, while the capacity of the nano-Si electrode with a conventional Cu foil drops to less than 70% only after 10 cycles. A full cell with the SiO negative electrode and the LiFePO4 positive electrode retains more than 90% of its capacity at the 10th cycle after 800 cycles. The conventional rolled Cu foil wrinkles during the cycling test. The high-strength clad current foil hardly deforms during the test regardless of the electrode size.

  5. Burn-up and neutron economy of accelerator-driven reactor

    SciTech Connect

    Takahashi, H.; Yang, W.; An, Y.; Yamazaki, Y.

    1997-07-01

    It is desirable to have only a small reactivity change in the large burn-up of a solid fuel fast reactor, so that the number of replacements or shuffling of the fuel can be reduced, and plant factor accordingly increased. Also, this reduces the number of control rods needed for the change in burn-up reactivity. In subcritical operation, power controlled by beam power is suggested, but this practice is not as economical as the use of control rods and makes more careful operation of the accelerator is required due to changes in the wake field. In subcritical operation, even a slightly subcritical one, the safety problems associated with a hard neutron spectrum can be alleviated. Neutron leakage from a flattened core, which is needed for operation of the critical fast reactor can be lessen by using the non flat core which has good neutron economy. For generating nuclear energy, it is essential to have a high neutron economy, although breeding the fuel is not welcomed in the present political climate, as is needed for transmuting long lived fission products. In contrast to the breeder, the accelerator driven reactor can separate the energy production from fuel production and processing. Thus, it is suited for non-proliferation of nuclear material by prohibiting the processing and production of fuel in the unrestricted area so this can be only done in international controlled areas which are restricted and remote.

  6. Weld overlay cladding with iron aluminides

    SciTech Connect

    Goodwin, G.M.

    1996-11-01

    The hot and cold cracking tendencies of some early iron aluminide alloy compositions limited their use to applications where good weldability was not required. Considerable progress has been made toward improving this situation. Using hot crack testing techniques developed at ORNL and a systematic study of alloy compositional effects, we have established a range of compositions within which hot cracking resistance is very good, essentially equivalent to stainless steel. Cold cracking, however, remains an issue, and extensive efforts are continuing to optimize composition and welding parameters, especially preheat and postweld heat treatment, to minimize its occurrence. In terms of filler metal and process development, we have progressed from sheared strip through aspiration cast rod and shielded metal arc electrodes to the point where we can now produce composite wire with a steel sheath and aluminum core in coil form, which permits the use of both the gas tungsten arc and gas metal arc processes. This is a significant advancement in that the gas metal arc process lends itself well to automated welding, and is the process of choice for commercial weld overlay applications. Using the newly developed filler metals, we have prepared clad specimens for testing in a variety of environments both in-house and outside ORNL, including laboratory and commercial organizations. As a means of assessing the field performance of this new type of material, we have modified several non-pressure boundary boiler components, including fuel nozzles and port shrouds, by introducing areas of weld overlay in strategic locations, and have placed these components in service in operating boilers for a side-by-side comparison with conventional corrosion-resistant materials.

  7. 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.

  8. ATR PDQ and MCWO Fuel Burnup Analysis Codes Evaluation

    SciTech Connect

    G.S. Chang; P. A. Roth; M. A. Lillo

    2009-11-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is being studied to determine the feasibility of converting it from the highly enriched Uranium (HEU) fuel that is currently uses to low enriched Uranium (LEU) fuel. In order to achieve this goal, it would be best to qualify some different computational methods than those that have been used at ATR for the past 40 years. This paper discusses two methods of calculating the burnup of ATR fuel elements. The existing method, that uses the PDQ code, is compared to a modern method that uses A General Monte Carlo N-Particle Transport Code (MCNP) combined with the Origen2.2 code. This modern method, MCNP with ORIGEN2.2 (MCWO), is found to give excellent agreement with the existing method (PDQ). Both of MCWO and PDQ are also in a very good agreement to the 235U burnup data generated by an analytical method.

  9. 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.

  10. An overview of the HSST Full-Thickness Shallow-Crack Clad Beam Testing Program

    SciTech Connect

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

    1994-09-01

    A testing program is described that will utilize full-thickness clad beam specimens to quantify fracture toughness for shallow flaws 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 that includes weld, plate and clad material. Metallurgical factors potentially influencing fracture toughness for shallow flaws in the beam specimen include material gradients due to welding and cladding applications, as well as material inhomogeneities in welded regions due to reheating in multiple weld passes. Fracture toughness tests focusing on shallow flaws in plate and weld material will also provide data for evaluating the relative influence of absolute and normalized crack depth on constraint conditions. Pretest finite-element analyses are described that provide near-tip stress and strain fields for characterization of constraint in the shallow-crack specimens in terms of the Q-stress. Analysis results predict a constraint loss in the shallow-crack clad beam specimen similar to that determined for a previously tested shallow-crack single-edge notch homogeneous bend specimen with the same normalized crack depth.

  11. 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.

  12. 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

  13. Parametric Study and Multi-Criteria Optimization in Laser Cladding by a High Power Direct Diode Laser

    NASA Astrophysics Data System (ADS)

    Farahmand, Parisa; Kovacevic, Radovan

    2014-12-01

    In laser cladding, the performance of the deposited layers subjected to severe working conditions (e.g., wear and high temperature conditions) depends on the mechanical properties, the metallurgical bond to the substrate, and the percentage of dilution. The clad geometry and mechanical characteristics of the deposited layer are influenced greatly by the type of laser used as a heat source and process parameters used. Nowadays, the quality of fabricated coating by laser cladding and the efficiency of this process has improved thanks to the development of high-power diode lasers, with power up to 10 kW. In this study, the laser cladding by a high power direct diode laser (HPDDL) as a new heat source in laser cladding was investigated in detail. The high alloy tool steel material (AISI H13) as feedstock was deposited on mild steel (ASTM A36) by a HPDDL up to 8kW laser and with new design lateral feeding nozzle. The influences of the main process parameters (laser power, powder flow rate, and scanning speed) on the clad-bead geometry (specifically layer height and depth of the heat affected zone), and clad microhardness were studied. Multiple regression analysis was used to develop the analytical models for desired output properties according to input process parameters. The Analysis of Variance was applied to check the accuracy of the developed models. The response surface methodology (RSM) and desirability function were used for multi-criteria optimization of the cladding process. In order to investigate the effect of process parameters on the molten pool evolution, in-situ monitoring was utilized. Finally, the validation results for optimized process conditions show the predicted results were in a good agreement with measured values. The multi-criteria optimization makes it possible to acquire an efficient process for a combination of clad geometrical and mechanical characteristics control.

  14. Using SERPENT Monte Carlo and Burnup code to model Traveling Wave Reactors (TWR)

    NASA Astrophysics Data System (ADS)

    Gulik, Volodymyr; Pavlovych, Volodymyr; Tkaczyk, Alan Henry

    2014-06-01

    This paper is mainly devoted to the proof-of-principle implementation of the SERPENT code for the simulation of traveling wave reactors. The investigation of SERPENT 1.1.19 code for multiprocessor tasks with long burnup steps was performed. The investigation of SERPENT 2 code for multiprocessor tasks with long burnup steps was performed. Methods to remove the influence of the ignition zone were considered, and neutronics simulations with various fragmentations of the burnup zone were performed.

  15. Numerical simulation of powder transport behavior in laser cladding with coaxial powder feeding

    NASA Astrophysics Data System (ADS)

    Liu, Hao; He, XiuLi; Yu, Gang; Wang, ZhongBin; Li, ShaoXia; Zheng, CaiYun; Ning, WeiJian

    2015-10-01

    Laser cladding with coaxial powder feeding is one of the new processes applied to produce well bonding coating on the component to improve performance of its surface. In the process, the clad material is transported by the carrying gas through the coaxial nozzle, generating gas-powder flow. The powder feeding process in the coaxial laser cladding has important influence on the clad qualities. A 3D numerical model was developed to study the powder stream structure of a coaxial feeding nozzle. The predicted powder stream structure was well agreed with the experimental one. The validated model was used to explore the collision behavior of particles in the coaxial nozzle, as well as powder concentration distribution. It was found that the particle diameter and restitution coefficient greatly affect the velocity vector at outlet of nozzle due to the collisions, as well as the powder stream convergence characteristics below the nozzle. The results indicated a practical approach to optimize the powder stream for the coaxial laser cladding.

  16. 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.

  17. Effect of hydrogen injection on hydrogen uptake by BWR fuel cladding. Final report

    SciTech Connect

    Cox, B.

    1983-06-01

    The hydrogen uptake rates reported for zirconium alloys in BWRs, PWRs, HWRs and various experimental in-reactor loops have been surveyed. The scatter in the data is large and arises from a variety of sources: variability in material properties; variations in temperature, irradiation flux, and water chemistry at different points in the reactor; and a lack of accurate knowledge of the sources of hydrogen which ends up in the zirconium-alloy cladding. An attempt has been made to assess the significance of these sources of variability in order to estimate the baseline for hydrogen uptake from the outside of good-quality cladding under normal BWR operating conditions. The probable effect of continuous hydrogen injection on this baseline has been estimated. It is concluded that severe localized hydrogen uptake, which might lead to hydride blisters is very improbable, but that an increase in the uniform rate of hydrogen absorption at the outside of the cladding may be expected.

  18. Localized in situ cladding annealing for post-fabrication trimming of silicon photonic integrated circuits.

    PubMed

    Spector, Steven; Knecht, Jeffrey M; Juodawlkis, Paul W

    2016-03-21

    We report the use of localized annealing via in situ heaters to induce a semi-permanent change in the refractive index of the cladding in ring resonator filters. When compared to other methods for post-fabrication trimming, this method has the advantage that no additional equipment, other than a supply of electrical power, is necessary to cause the index change. Two cladding materials were used: hydrogen silsesquioxane (HSQ) for samples that were externally annealed, and PECVD oxide for samples that were annealed with in situ heaters. The resonant wavelengths could be adjusted by as much as 3.0 nm and 1.7 nm for the HSQ and PECVD cladded filters, respectively. The trimming of a 5 channel, single ring filter bank, and a single, double ring filter is demonstrated. PMID:27136793

  19. 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.

  20. A strain sensor based on cladding mode resonance of optical double-cladding fiber

    NASA Astrophysics Data System (ADS)

    Zhang, Jian; Pang, Fufei; Guo, Hairun; Chen, Zhenyi; Wang, Tingyun

    2010-11-01

    A strain sensor based on cladding mode resonance of optical double-cladding fiber (DCF) was proposed and experimentally demonstrated. The sensor head was fabricated by splicing a section of DCF into a standard single mode fiber (SMF). Attributed to the thin thickness of the inner cladding, the core mode can be coupled with cladding modes which generatess a strong resonant spectrum at the phase-matching wavelength. When the DCF sensor is applied an axial strain, the refractive index of the DCF decreases due to the photoelastic effect. According to the coupled mode theory, the phase-matching wavelength will shift to a shorter wavelength. By detecting the resonant spectrum variation, the stain sensor can be realized. The strain sensitivity was achieved as -2.87 pm μɛ over 800μɛ measurement range with good repeatability. With the simple configuration and attractive performance, the specialty DCF strain sensor can be explored for wide sensing applications.

  1. 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

  2. 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.

  3. 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.

  4. Ceramic breeder materials

    SciTech Connect

    Johnson, C.E.

    1990-01-01

    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 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. 32 refs., 1 fig., 1 tab.

  5. 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.

  6. Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

    SciTech Connect

    Ott, Larry J; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J; Bevard, Bruce Balkcom

    2009-01-01

    The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

  7. Thermal conductance at millikelvin temperatures of woven ribbon cable with phosphor-bronze clad superconducting wires

    NASA Astrophysics Data System (ADS)

    Woodcraft, Adam L.; Ventura, Guglielmo; Martelli, Valentina; Holland, Wayne S.

    2010-08-01

    Woven Nomex® ribbon cables made up with superconducting niobium-titanium wire are used at millikelvin temperatures in many large cryogenic instruments. It is important to know how much heat in transmitted down such cables. However, the conductivity of the materials used is not well known. Another problem is that the wires are normally clad with alloys which exhibit some magnetism. This is a potential problem for instruments employing superconducting detectors. A safe non-magnetic alternative to the usual materials is phosphor-bronze clad niobium-titanium wiring. However, there is little experience with such wires. We have therefore measured the conductance of a ribbon cable made up with these wires. The measured values are in good agreement with our predictions, suggesting that the values we have used to model the cable are sufficiently accurate, and could therefore be used to predict the performance of ribbon cables using other cladding materials, so long as the conductivity of the cladding is reasonably well known. As part of our analysis, we consider the likely variation in thermal conductivity values for C51000 phosphor bronze caused by legitimate variations in composition.

  8. 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.

  9. Hydrogen motion in Zircaloy-4 cladding during a LOCA transient

    NASA Astrophysics Data System (ADS)

    Elodie, T.; Jean, D.; Séverine, G.; M-Christine, B.; Michel, C.; Berger, P.; Martine, B.; Antoine, A.

    2016-04-01

    Hydrogen and oxygen are key elements influencing the embrittlement of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA). The understanding of the mechanisms influencing the motion of these two chemical elements in the metal is required to fully describe the material embrittlement. High temperature steam oxidation tests were performed on pre-hydrided Zircaloy-4 samples with hydrogen contents ranging between 11 and 400 wppm prior to LOCA transient. Thanks to the use of both Electron Probe Micro-Analysis (EPMA) and Elastic Recoil Detection Analysis (μ-ERDA), the chemical elements partitioning has been systematically quantified inside the prior-β phase. Image analysis and metallographic examinations were combined to provide an average oxygen profile as well as hydrogen profile within the cladding thickness after LOCA transient. The measured hydrogen profile is far from homogeneous. Experimental distributions are compared to those predicted numerically using calculations derived from a finite difference thermo-diffusion code (DIFFOX) developed at IRSN.

  10. 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.

  11. Influence of porosity formation on irradiated UO2 fuel thermal conductivity at high burnup

    NASA Astrophysics Data System (ADS)

    Roostaii, B.; Kazeminejad, H.; Khakshournia, S.

    2016-10-01

    Based on the existing low temperature high burnup gaseous swelling model for UO2 fuel, the matrix swelling terms are calculated and the formation of total volume porosity up to burnup of 120 MWd/KgU is computed. The irradiated UO2 thermal conductivity model based on the Maxwell-Eucken correlation for porosity factor is selected as a case study and the calculation of porosity evolution with burnup is carried out. It is shown that taking into account the formation of porosity with burnup compared to the case with constant porosity equal to as-fabricated value leads to a decrease in the UO2 fuel thermal conductivity up to 15% at high burnup values of 120 MWd/kgU. Results of the calculations are also compared with the available experimental data and good agreement was found. The conducted parametric study clearly demonstrated the impact of the key parameters on the results of the present investigation.

  12. End effect K{sub eff} bias curve for actinide-only burnup credit casks

    SciTech Connect

    Kang, C.H.; Lancaster, D.B.

    1997-03-01

    A conservative end effect k{sub eff} bias curve for actinide-only burnup credit for spent fuel casks is presented in this paper. The k{sub eff} bias values can be added to the uniform axial burnup analysis to conservatively bound the actinide-only end effect. A normalized axial burnup distribution for the standard Westinghouse 17 x 17 assembly design is used for calculating k{sub eff}. The end effect calculated is a strong function of burnup, and increases as cask size size decreases. The presence of poison plates increases the end effect. The bias curve presented is based on the most limiting cask configuration of a single PWR assembly with completely black poison plates. Therefore, axially uniform criticality calculations with application of the proposed k{sub eff} could eliminate the need for axially burnup dependent analyses. 7 refs., 1 fig.

  13. 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

  14. 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.

  15. 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.

  16. 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.

  17. Modelling of laser cladding of magnesium alloys with pre-placed powders

    NASA Astrophysics Data System (ADS)

    Guo, Li-Feng

    As a surface engineering technique, high-power laser cladding, has shown great potential for improving the corrosion resistance of magnesium alloys. Its main advantage over other processes, is its ability to form relatively thick protective coatings on selected areas where improved properties are desired. It is also a 'clean' process. However, previous research studies have found that in laser cladding of magnesium alloys, the problem of a high degree of dilution cannot be easily overcome. Moreover, in-depth studies using analytical or numerical modelling can rarely be found in the literature for addressing laser cladding with pre-placed powders with the aim of predicting the level of dilution. In the first phase of this study, a simplified thermal model based on the finite element method (FEM) was developed to study the phenomenon of dilution in laser cladding of a magnesium alloy. In the model, the powder bed was treated as a continuum, and a high power continuous wave (CW) laser was employed. The results of the simulations of the FEM model together with those of the statistical analyses showed that although, under normal cladding conditions, a process window can be established for achieving good interfacial bonding between the substrate and the clad coating, a low dilution level was extremely difficult to achieve. This was primarily attributed to the low melting point and the high thermal diffusivity of magnesium as well as the relatively long laser-material interaction time. To overcome the dilution problem, the double-layer cladding technique was explored, and was found to be able to produce low dilution clads with improved corrosion resistance. In considering the improvement of corrosion resistance that can be caused by laser surface modification to magnesium alloys, a comparison was made between the techniques of laser surface melting and laser cladding. The results of the potentiodynamic polarisation tests showed that the improvement obtained from laser

  18. 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

  19. Extension and validation of the TRANSURANUS burn-up model for helium production in high burn-up LWR fuels

    NASA Astrophysics Data System (ADS)

    Botazzoli, Pietro; Luzzi, Lelio; Brémier, Stephane; Schubert, Arndt; Van Uffelen, Paul; Walker, Clive T.; Haeck, Wim; Goll, Wolfgang

    2011-12-01

    The TRANSURANUS burn-up model (TUBRNP) calculates the local concentration of the actinides, the main fission products, and 4He as a function of the radial position across a fuel rod. In this paper, the improvements in the helium production model as well as the extensions in the simulation of 238-242Pu, 241Am, 243Am and 242-245Cm isotopes are described. Experimental data used for the extended validation include new EPMA measurements of the local concentrations of Nd and Pu and recent SIMS measurements of the radial distributions of Pu, Am and Cm isotopes, both in a 3.5% enriched commercial PWR UO 2 fuel with a burn-up of 80 and 65 MWd/kgHM, respectively. Good agreement has been found between TUBRNP and the experimental data. The analysis has been complemented by detailed neutron transport calculations (VESTA code), and also revealed the need to update the branching ratio for the 241Am(n,γ) 242mAm reaction in typical PWR conditions.

  20. 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

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

    SciTech Connect

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

    1985-02-01

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

  2. 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.

  3. 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

  4. Anisotropic Properties of Stainless Steel—Clad Aluminum Sheet

    NASA Astrophysics Data System (ADS)

    Kim, Daeyong; Hwang, Bum Kyu; Lee, Young Seon; Kim, Ji Hoon; Kim, Min-Joong

    2010-06-01

    The production of a stainless steel—clad aluminum sheet by the cold rolling process is a more efficient and economical approach compared with the other types of processes utilized for the production of such sheets. Because both the stainless steel and aluminum sheets show the highly anisotropic behavior, it is necessary to investigate anisotropic properties of clad sheets for the design of process. In this paper, to investigate the anisotropic properties of stainless steel—clad aluminum sheet, two kinds of clad sheets were considered: STS439/AA3003 and STS439/AA1050/STS304 clad sheets. The uni-axial tension tests at 0, 45 and 90 degrees for the rolling direction were performed to obtained yield stresses and R values. The strain ratio at balanced biaxial tension state was measured from compression disk test. In order to describe the anisotropic behavior of the clad sheet, nonquadratic anisotropic yield function Yld2000-2d was utilized.

  5. 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

  6. 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).

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

    SciTech Connect

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

    1988-02-01

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

  8. 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.

  9. 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.

  10. Clad fiber capacitor and method of making same

    DOEpatents

    Tuncer, Enis

    2013-11-26

    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; a ductile, electrically conductive sleeve positioned over the cladding. One or more of the 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.

  11. 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.

  12. 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)

  13. 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.

  14. 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

  15. 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%.

  16. An acoustic vibration sensor based on tapered triple cladding fiber

    NASA Astrophysics Data System (ADS)

    Xu, Hui; Pang, Fufei; Zhao, Shiqi; Chen, Zhenyi; Wang, Tingyun

    2014-05-01

    An acoustic vibration sensor is investigated and demonstrated by using a tapered triple cladding fiber (TCF). It is fabricated by tapering a length of 2 cm TCF which is spliced between two single mode fibers (SMF). The TCF consists of core, inner cladding, middle cladding and outer cladding. After the tapering process, this structure becomes a tapered coaxial fiber coupler which presents a periodic filtering transmission spectrum. The surrounding vibration perturbation can be directly demodulated by intensity detection of the transmission power at a particular wavelength. The experimental result shows that the maximum frequency response of 700 kHz is achieved.

  17. 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.

  18. Electroslag Strip Cladding of Steam Generators With Alloy 690

    SciTech Connect

    Consonni, M.; Maggioni, F.; Brioschi, F.

    2006-07-01

    The present paper details the results of electroslag cladding and tube-to-tubesheet welding qualification tests conducted by Ansaldo-Camozzi ESC with Alloy 690 (Alloy 52 filler metal) on steel for nuclear power stations' steam generators shell, tubesheet and head; the possibility of submerged arc cladding on first layer was also considered. Test results, in terms of chemical analysis, mechanical properties and microstructure are reproducible and confidently applicable to production cladding and show that electroslag process can be used for Alloy 52 cladding with exceptionally stable and regular operation and high productivity. The application of submerged arc cladding process to the first layer leads to a higher base metal dilution, which should be avoided. Moreover, though the heat affected zone is deeper with electroslag cladding, in both cases no coarsened grain zone is found due to recrystallization effect of second cladding layer. Finally, the application of electroslag process to cladding of Alloy 52 with modified chemical composition, was proved to be highly beneficial as it strongly reduces hot cracking sensitivity, which is typical of submerged arc cladded Alloy 52, both during tube-to-tubesheet welding and first re-welding. (authors)

  19. Neutronic and burnup characteristics of an actinide-reduced plutonium fuel with tungsten

    SciTech Connect

    Chang, G.S.

    1998-04-01

    The US and Russia expect to have a surplus of {approximately} 150 tonnes of weapons-grade plutonium (WGP) and 1,000 tonnes of weapons-grade uranium resulting from drastic reductions in nuclear weapons programs. One of the most favored candidate methods for disposing of the WGP is to blend it with natural or depleted uranium down to 5 to 7 wt% of WGP for light water reactor (LWR) fuel pellet fabrication. However, this approach, with a conversion ratio of 0.6, will produce more plutonium and other actinides in the spent fuel than the nonfertile fuel and the proposed actinide-reduced plutonium fuel (ARPF). This process only transforms the weapons-grade fissile materials to civilian-grade plutonium, which is still a non-proliferation concern, so it does not completely solve the plutonium disposition problem. Disposition of WGP in reactors without fertile material has been proposed by industry and national laboratories. A new ARPF is described that would use WGP mixed with medium-enrichment (20 at.% < {sup 235}U < 93 at.%) UO{sub 2} and the nonfertile material tungsten to achieve a conversion ratio < 0.1. The ARPF can meet the WGP disposal goal while minimizing the plutonium production. Its physics and burnup characteristics are analyzed, and the results are compared with LWR UO{sub 2} and mixed-oxide fuel.

  20. Stress analysis and probabilistic assessment of multi-layer SiC-based accident tolerant nuclear fuel cladding

    NASA Astrophysics Data System (ADS)

    Stone, J. G.; Schleicher, R.; Deck, C. P.; Jacobsen, G. M.; Khalifa, H. E.; Back, C. A.

    2015-11-01

    Silicon carbide (SiC) fiber, SiC matrix composites (SiC/SiC) are being considered as a cladding material for light water reactors in order to improve safety performance. Engineered, multi-layer cladding designs consisting of both monolithic SiC (mSiC) and SiC/SiC have been examined as promising concepts to meet both strength and impermeability requirements. A new model has been developed to calculate stresses and failure probabilities for multi-layer cladding consisting of SiC-based materials in reactor operating conditions. The results show that stresses in SiC-based cladding are dominated by temperature-dependent irradiation-induced swelling, with the largest stresses occurring during the cold shutdown conditions. Failure probabilities are driven by the resulting tensile stresses at the cladding inner wall, while the outer wall is subject to compressive stresses. This indicates that the inner SiC/SiC, outer mSiC concept has the lowest failure probability, as the pseudo-plastic deformation of the composite reduces tensile loading and the compressed monolith provides a reliable, impermeable barrier to fission product release.

  1. 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).

  2. 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.

  3. 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.

  4. Photonic lantern with cladding-removable fibers

    NASA Astrophysics Data System (ADS)

    Sun, Weimin; Yan, Qi; Bi, Yao; Yu, Haijiao; Liu, Xiaoqi; Xue, Jiuling; Tian, He; Liu, Yongjun

    2014-07-01

    Recently, spectral measurement becomes an important tool in astronomy to find exoplanets etc. The fibers are used to transfer light from the focal plate to spectrometers. To get high-resolution spectrum, the input slits of the spectrometers should be as narrow as possible. In opposite, the light spots from the fibers are circle, which diameters are clearly wider than the width of the spectrometer slits. To reduce the energy loss of the fiber-guide star light, many kinds of image slicers were designed and fabricated to transform light spot from circle to linear. Some different setup of fiber slicers are introduced by different research groups around the world. The photonic lanterns are candidates of fiber slicers. Photonic lantern includes three parts: inserted fibers, preform or tubing, taped part of the preform or tubing. Usually the optical fields concentrate in the former-core area, so the light spots are not uniform from the tapered end of the lantern. We designed, fabricated and tested a special kind of photonic lantern. The special fibers consist polymer cladding and doped high-index core. The polymer cladding could be easily removed using acetone bath, while the fiber core remains in good condition. We inserted the pure high-index cores into a pure silica tubing and tapered it. During the tapering process, the gaps between the inserted fibers disappeared. Finally we can get a uniform tapered multimode fiber end. The simulation results show that the longer the taper is, the lower the loss is. The shape of the taper should be controlled carefully. A large-zone moving-flame taper machine was fabricated to make the special photonic lantern. Three samples of photonic lanterns were fabricated and tested. The lanterns with cladding-removable fibers guide light uniform in the tapered ends that means these lanterns could collect more light from those ends.

  5. Zircaloy cladding degradation under repository conditions

    SciTech Connect

    Santanam, L.; Raghavan, S.; Chin, B.A.; Shaw, H.

    1990-12-01

    Creep, a potential degradation mechanism of Zircaloy cladding after repository disposal of spent nuclear fuel, has been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. Maximum allowable temperatures are 340{degree}C (613 K) for typically stressed rods (70--100 MPa) and 300{degree}C (573 K) for highly stressed rods (140--160 MPa). 10 refs., 2 figs.

  6. Effective absorption in cladding-pumped fibers

    NASA Astrophysics Data System (ADS)

    Zervas, Michalis N.; Marshall, Andy; Kim, Jaesun

    2011-02-01

    We investigate experimentally and theoretically the wavelength dependence of the pump absorption along Yb3+-doped fibers, for cladding-pumped single as well as coupled multimode (GTWaveTM) fibers. We show that significant spectral absorption distortions occur along the length with the 976nm absorption peak affected the most. We have developed a novel theoretical approach, based on coupled mode theory, to explain the observed effects. We have also investigated the mode mixing requirements in order to improve the absorption spectral distribution along the increase the overall absorption efficiency and discuss the implications on fiber laser performance.

  7. Development of Yttrium Stabilized Zirconia (YSZ) diffusion barrier coatings for mitigation of Fuel-Cladding Chemical Interactions

    NASA Astrophysics Data System (ADS)

    Firouzdor, Vahid; Brechtl, Jamieson; Wilson, Lucas; Semerau, Brandon; Sridharan, Kumar; Allen, Todd R.

    2013-07-01

    Fuel-Cladding Chemical Interactions (FCCIs) in a nuclear reactor occurs due to thermal and radiation enhanced inter-diffusion between the cladding and fuel materials. This can have the detrimental effects of reducing the effective cladding wall thickness and the formation of low melting point eutectic compounds. Deposition of thin diffusion barrier coatings in the inner surface of the cladding can potentially reduce or delay the onset of FCCI. This study examines the feasibility of using nanofluid-based electrophoretic deposition (EPD) process to deposit coatings of Yttrium Stabilized Zirconia (YSZ) as the diffusion barrier coating. The deposition parameters, including the nanofluid solvent, additive, particle size, current, and voltage were optimized using test flat substrates of T91 ferritic-martensitic steel. A post deposition sintering step was also conducted and optimized to improve the bonding and mechanical integrity of the coating. Diffusion characteristics of the coatings were investigated by diffusion couple experiments using cerium as a fuel fission product responsible for solid state FCCI. These diffusion couple studies performed at 575 °C for 100 h showed that the YSZ coatings significantly reduced the solid state inter-diffusion between cerium and steel. A heat transfer model was developed to simulate the changes in temperature profile inside the fuel cladding by addition of YSZ coating. It was found that even though the temperature can increase in the coated cladding, the temperature falls below the melting point of uranium and eutectic temperature in Fe-U phase diagram. Using a co-axial configuration in conjunction with the EPD process, YSZ was successfully deposited uniformly on the inner surfaces of 12″ length sections of cladding with 4 mm inner diameter. Such a coating is extremely hard to make by conventional coating technologies like thermal spray or vapor deposition.

  8. Laser Clad Nickel Based Superalloys: Microstructure Evolution And High Temperature Oxidation Studies

    NASA Astrophysics Data System (ADS)

    Sircar, S.; Ribaudo, C.; Mazumder, J.

    1988-10-01

    Application of alloy coatings with superior oxidation resistance at elevated temperatures (1200°C) on superalloy components is of interest at present. There is a general consensus that the addition of rare earths such as hafnium (Hf) to these alloys has a pronounced effect on their performance. An in situ laser cladding technique was used to produce Ni-Al-Cr-Hf alloys on a nickel alloy substrate. Scanning Electron Microscope (SEM), Transmission Electron Microscope (TEM), and Scanning Transmission Electron Microscope (STEM) attached with Energy Dispersive X-ray (EDX) analyzers were employed for microstructural evolution studies of alloys produced during the laser cladding process. The microstructure of these alloys mainly consists of dendrites of Y' of the Ni3Al type with about 11-14 wt% Hf and an interdendritic eutectic phase. Electron microscopy in the dendritic zones reveals ordered domains whose morphology depends on laser cladding process parameters. Variation in these parameters produced only subtle changes in the composition and cell spacing of the dendritic phase. The eutectic constituent consists of a Hf-rich phase and a Hf-lean phase in an alternating lamellar structure. Convergent beam diffraction and x-ray spectroscopy techniques were used to characterize the constituents. A possible phase transformation sequence has been suggested. Differential Thermal Analysis (DTA) work indicates that the Y' dissolution temperature for the claddings is at least as high as the substrate material (Rene 80). Single cycle oxidation tests of eight hours at 1200°C in slowly flowing air reveal that the claddings have a lower weight gain rate than the substrate itself. Microchemistry and microstructure of the oxidized samples are examined using SEM attached with EDX and Auger Electron Spectroscopic (AES) techniques. The improvement in the oxidation resistance is believed to be at least partially due to the mechanical pegging between alumina coated hafnia protrusions and the

  9. Consideration of Lower Allowable Impact Temperature for DOP-26 Iridium Alloy Fueled Clads

    NASA Astrophysics Data System (ADS)

    Skrabek, Emanuel A.

    2005-02-01

    The current DOE General Purpose Heat Source (GPHS) was developed specifically for a SiGe Radioisotope Thermoelectric Generator (RTG) which was to be launched on a Space Shuttle. In such a launch, if the final stage of the deployed spacecraft failed to ignite, the spacecraft and RTG would reenter the atmosphere after about three months in a Shuttle parking orbit. The indium alloy clad containing the radioisotope fuel would have then been at about 1300° C for the entire time on orbit. This would lead to some grain growth in the clad material, which would tend to decrease its ductility. Due to these constraints, it was concluded by the heat source community that safety considerations would require that the clads be maintained at a temperature of at least 930° C through impact. The safety testing for the GPHS RTG programs was therefore conducted with these conditions as guides. Within this set of guidelines, the heat source was shown to be able to safely handle all credible launch and reentry accident events in these missions. Currently proposed missions plan to use different power conversion systems and launch scenarios that are able to utilize much lower operating temperatures. Imposing the high operating and impact temperatures required by the SiGe RTG onto these new systems would lead to unnecessary added design complexity and probably a less robust system. If the heat source is operated so that the iridium alloy clads are generally maintained at less than 1000° C, essentially no grain growth would be experienced over normal mission lifetimes, and the clads would stay more ductile and able to safely withstand the resulting lower impact temperatures. Therefore DOE has embarked on a study to determine the lowest safe operating and impact temperatures for the GPHS. This paper presents the historical background and some preliminary results of this study.

  10. Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding

    DOE PAGESBeta

    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

  11. Radial Power Profile of MOX and LEU Fuel Pellet Versus Burnup

    SciTech Connect

    Chang, Gray S.; Pedersen, Robert C.

    2002-07-01

    One of challenge to burn the WG-Pu in Mixed Oxide (MOX) fuel in light water reactors (LWR) is to demonstrate that the differences between WG-MOX, RG-MOX, and LWR LEU fuel are minimal, and therefore, the commercial MOX and LEU fuel experience base is applicable. The MCWO-calculated Radial Power Profile of LEU, Weapons Grade-MOX and Reactor Grade-MOX fuel pellets at various burnups are similar toward the end of life (50 GWd/t). Therefore, the LEU fuel performance evaluation code - FRAPCON-3 with modifications, such as, the detailed fission power profiles versus burnup, can be used in the MOX fuel pellet performance analysis. MCWO also calculated the {sup 240}Pu/Pu ratio in WG-MOX versus burnup, which reaches an average of 31.25% at discharged burnup of 50 GWd/t. It meets the spent fuel standard for WG-Pu disposition in LWR. (authors)

  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. 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.

  14. High burn-up structure of U(Mo) dispersion fuel

    NASA Astrophysics Data System (ADS)

    Leenaers, A.; Van Renterghem, W.; Van den Berghe, S.

    2016-08-01

    The evolution of the high burn-up structure (HBS) in U(Mo) fuel irradiated up to a burn-up of ∼70% 235U or ∼5 × 1021 f/cm3 or ∼120 GWd/tHM is described and compared to the observation made on LWR fuel. Scanning and transmission electron microscopy was performed on several samples having different burn-ups in order to get a better understanding of the mechanisms leading to the high burn-up structure formation. Even though there are some substantial differences between the irradiation of ceramic and U(Mo) alloy fuels (crystal structure, enrichment, irradiation temperature …), it was found that in both fuels recrystallization initiates at the same threshold and progresses in a similar way with increasing fission density. In case of U(Mo), recrystallization leads to accelerated swelling of the fuel which could result in instability of the fuel plate.

  15. Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

    SciTech Connect

    Schneider, E.; Deinert, M.; Bingham Cady, K.

    2006-07-01

    Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)

  16. Burnup of fusion produced tritons and /sup 3/He ions in PLT and PDX

    SciTech Connect

    Heidbrink, W.W.; Chrien, R.E.; Strachan, J.D.

    1982-09-01

    The d(d,p)t and d(d,n)/sup 3/He fusion reactions produce 1 MeV tritons and 0.8 MeV /sup 3/He ions which can subsequently undergo d(t,n)..cap alpha.. and d(/sup 3/He,p)..cap alpha.. fusion reactions. The magnitude of this triton and /sup 3/He ion burnup was measured on the PLT and PDX tokamaks by detection of the 14 MeV neutron and 15 MeV proton emission. In discharges with B/sub phi/ greater than or equal to 2 T, the measured /sup 3/He burnup agrees well with predictions based on classical theories of ion confinement and slowing down, while the triton burnup was about four times lower than theoretically predicted. In discharges with weaker toroidal fields, the burnup of both ions fell by more than a factor of ten.

  17. Advanced Fabrication Technique and Thermal Performance Prediction of U-Mo/Zr-alloy Dispersion Fuel Pin for High Burnup PWR

    NASA Astrophysics Data System (ADS)

    Suwardi

    2010-06-01

    In recent years, a novel class of zirconium alloys having the melting temperature of 990-1160 K has been developed. Based on novel zirconium matrix alloys, high uranium content fuel pin with U-9Mo has been developed according to capillary impregnation technique. The pin shows it is thermal conductivity ranging from 18 to 22 w/m/K that is comparably higher than UO2 pellet pin. The paper presents the met-met fabrication and thermal performance analysis of the fuel in typical PWR. The fabrication consists of mixing UO2 powder or granules and a novel Zr-alloy powder having low melting point, filling the mixture in a cladding tube that one of its end has been plugged, heating the pin to above melting temperature of Zr-alloy for an hour, natural cooling and heat treating at 300 K for 1/2 hr. The thermal analysis takes into account the pore and temperature distribution and high burn up effect to pellet conductivity. The thermal diffusivity ratio of novel to conventional fuel has been used as correction factor for the novel fuel conductivity. The results show a significant lowering pellet temperature along the radius until 1000 K at the hottest position. The analysis underestimates since the gap conductivity has been treated as decreased by 2% fission gas released that is not real since the use of lower temperature, and also decreasing thermal conductivity by porosity formation will much lower. The analysis shows that the novel fuel has very good thermal properties which able to pass the barrier of 65 MWD/kg-U, the limit to day commercial fuel. The burn-up extension means fewer fresh fuel is needed to produce electricity, preserve natural uranium resource, easier fuel handling operational per energy produced

  18. Achieving Inclusion through CLAD: Collaborative Learning Assessment through Dialogue

    ERIC Educational Resources Information Center

    Fitch, E. Frank; Hulgin, Kathleen M.

    2008-01-01

    This study measures the effectiveness of Collaborative Learning Assessment through Dialogue (CLAD) on reading achievement in inclusive classrooms in the USA. The CLAD process involved students collaboratively completing multiple-choice quizzes, using dialogue and critical thinking to reach consensus and receiving immediate feedback on their…

  19. 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.

  20. 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).

  1. 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.

  2. Laser cladding of Ni-Cr-Al-Hf on inconel 718 for improved high-temperature oxidation resistance

    SciTech Connect

    Singh, J.; Nagarathnam, K.; Mazumder, J.

    1987-08-03

    In-situ Ni-Cr-Al-Hf alloy was developed by laser surface cladding with a mixed-powder feed for improved high-temperature oxidation resistance. Oxidation-resistant materials for operation at elevated temperatures must satisfy two requirements: diffusion through the oxide scale must occur at the lowest possible rate, and the oxide scale must resist spallation. Formation of an Al/sub 2/O/sub 3/ protective scale fulfills the former requirement but its adherence is poor. A reactive metal such as Hf is added to improve adhesion. A 10-kW CO/sub 2/ laser was used for laser cladding. Optical, SEM and STEM microanalysis techniques were employed to characterize the different phases produced during the cladding process. Microstructural studies showed a high degree of grain refinement, increased solid solubility of Hf in the matrix and the formation of Hf-rich precipitates. A thermogravimetric analysis was carried out to determine the oxidation properties of these clad alloys with an extended solid solution of Hf. Considerable improvement over the base metal was observed. This paper discusses microstructural development in this laser-clad alloy and its effect on oxidation.

  3. Gross Gamma Dose Rate Measurements for TRIGA Spent Nuclear Fuel Burnup Validation

    SciTech Connect

    Winston, Philip Lon; Sterbentz, James William

    2001-04-01

    Gross gamma-ray dose rates from six spent TRIGA fuel elements were measured and compared to calculated values as a means to validate the reported element burnups. A newly installed and functional gamma-ray detection subsystem of the In-Cell Examination System was used to perform the measurements and is described in some detail. The analytical methodology used to calculate the corresponding dose rates is presented along with the calculated values. Comparison of the measured and calculated dose rates for the TRIGA fuel elements indicates good agreement (less than a factor of 2 difference). The intent of the subsystem is to measure the gross gamma dose rate and correlate the measurement to a calculated dose rate based on the element s known burnup and other pertinent spent fuel information. Although validation of the TRIGA elements’ burnup is of primary concern in this paper, the measurement and calculational techniques can be used to either validate an element’s reported burnup or provide a burnup estimate for an element with an unknown burnup.

  4. 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.

  5. Laboratory and field evaluations of clad Christmas tree equipment

    SciTech Connect

    Sisak, W.J. ); Gordon, J.R. )

    1991-02-01

    In this paper a methodology is presented for evaluating the corrosion resistance of clad equipment in sour environments. This paper describes the laboratory and field test programs used to evaluate the suitability for sour service of gate valves clad with Alloy UNS NO6625 by weld overlay and by hot, isostatic pressing (HIP). The laboratory tests used to evaluate corrosion and cracking resistance of the clads are reviewed. Experience with actual field installations of both types of clad valves is presented, and the test/inspection methods used to evaluate the equipment during and after the service period are discussed. Finally, laboratory and field results are compared to highlight key findings related to the clad valve bodies and to the trim components.

  6. Wideband slow-light modes for time delay of ultrashort pulses in symmetrical metal-cladding optical waveguide.

    PubMed

    Zheng, Yuanlin; Yuan, Wen; Chen, Xianfeng; Cao, Zhuangqi

    2012-04-23

    A widebandwidth optical delay line is a useful device for various fascinating applications, such as optical buffering and processing of ultrafast signal. Here, we experimentally demonstrated effective slow light of sub-picosecond signal over 10 THz frequency range by employing the wide slow light modes in thick symmetrical metal-cladding optical waveguide (SMCOW). Ultrahigh-order guided modes travelling as slow light in waveguide together with strong confinement provided by metal-cladding makes this scheme nearly material dispersion independent and compatible with wide bandwidth operation.

  7. 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

    ... responses to its notice of institution (77 FR 5052, February 1, 2012) were adequate. A record of the... 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...

  8. 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.

  9. 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.).

  10. 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

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. Raman and XPS characterization of fuel cladding interactions using miniature specimens

    NASA Astrophysics Data System (ADS)

    Windisch, C. F., Jr.; Henager, C. H., Jr.; Engelhard, M. H.; Bennett, W. D.

    2009-01-01

    A combination of laser Raman spectroscopy and X-ray photoelectron spectroscopy was applied in a study of fuel-cladding chemical interactions on miniature oxide-coated HT-9 disks at elevated temperature. The experiments were intended as a preliminary step toward the development of a quick-screening technique for candidate alloys for cladding materials and actinide-based mixed oxide fuel mixtures. The results indicated that laser Raman spectroscopy was capable of determining the major oxides on HT-9 and how they changed in composition due to heating. However, X-ray photoelectron spectroscopy was necessary to identify the role of the metallic phases and provide depth resolution. Using the two techniques the kinetics of chromia growth were shown to be affected by the presence of an applied oxide coating. A single replacement reaction involving residual reduced metal within the coating was also identified.

  16. Graded-index thin-film stack for cladding and coupling.

    PubMed

    Lim, Kim Peng; Ng, Doris Keh Ting; Pu, Jing; Toh, Yeow Teck; Febiana, Tjiptoharsono; Vivek, Krishnamurthy; Wang, Qian

    2016-08-20

    A graded-index multilayer thin-film stack is optimized to act as a cladding layer on top of a silicon (Si) nanowaveguide and also a collimator for chip coupling where the waveguide ends. The numerical example shows an optimized graded-index profile from 2.35 to 1.45 provides an optical coupling to the standard single-mode fiber with efficiency close to 90% while retaining tight light confinement for the Si nanowaveguide. The corresponding material realization of a graded-index profile with a Si-rich nitride SiNx/SiON/SiO2 system is explored using inductively coupled plasma chemical vapor deposition, and a SiNx cladded Si waveguide is demonstrated. PMID:27556999

  17. Reliable fabrication with alloy 59 hot roll-clad plates in a major FGD-project

    SciTech Connect

    Kirchheiner, R.; Herda, W.; Schupp, N.

    1996-08-01

    Flue gas desulfurization (FGD) units require large components to accommodate the flow volumes and short contact times between raw gas and absorbent which are typical of these processes. For over two decades, increasing use has been made of nickel base alloys to protect flue gas ducts, absorbers, chimneys and internals against corrosion attack. In order to take advantage of their excellent corrosion resistance while keeping investment costs low, these NiCrMo-materials are frequently used in the form of a cladding applied in thin layers on the carbon steel structural load-bearing component. In view of the thin corrosion protection layer, the quality of the welded joint on the nickel alloy side assumes critical importance. This paper describes the development of tests and welding procedures for hot roll-clad plates used in large FGD components in a modern lignite-fired 3,500 MW power plant in Germany.

  18. Analysis of microstructure and properties of multilayer coatings produced by laser cladding

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

    Purpose of the work is to prepare multilayer coatings corresponding to specified requirements to recovery and improvement of surface details. Requirements to coatings: providing durable and reliable adhesion base and filler materials, absence of pores, cracks, delaminations, reducing mixing metal base and cladding. We used iron-based PR-10R6M5 and tungsten carbide Hoganas 44712 powders. Experimental determination of the optimal technological mode of application of the single track, the coefficient of overlapping tracks to create a full layer, the angle of the second cladding layer, relative to the first one and, finally, the determination of the optimal additive tungsten carbide to achieve increased durability were produced to fulfill these requirements.

  19. TiN-based coatings on fuel cladding tubes for advanced nuclear reactors

    SciTech Connect

    Ickchan Kim; Fauzia Khatkhatay; Liang Jiao; Greg Swadener; James Cole; Jian Gan; Haiyan Wang

    2012-10-01

    Titanium nitride (TiN) thin films are coated on HT-9 and MA957 fuel cladding tubes and bars to explore their mechanical strength, thermal stability, diffusion barrier properties, and thermal conductivity properties. The ultimate goal is to implement TiN as an effective diffusion barrier to prevent the inter-diffusion between the nuclear fuel and the cladding material, and thus lead to a longer life time of the cladding tubes. Mechanical tests including hardness and scratch tests for the samples before and after thermal cycle tests show that the films have a high hardness of 28 GPa and excellent adhesion properties despite the thermal treatment. Thermal conductivity measurements demonstrate that the thin TiN films have very minimal impact on the overall thermal conductivity of the MA957 and HT9 substrates, i.e., the thermal conductivity of the uncoated HT-9 and MA957 substrates was 26.25 and 28.44 W m-1K-1 , and that of the coated ones was 26.21 and 28.38 W m-1K-1, respectively. A preliminary Ce diffusion test on the couple of Ce/TiN/HT-9 suggests that TiN has excellent material compatibility and good diffusion barrier properties.

  20. 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.

  1. Hollow core anti-resonant fiber with split cladding.

    PubMed

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

    2016-04-01

    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.

  2. Double clad tapered fiber for high power applications.

    PubMed

    Filippov, V; Chamorovskii, Yu; Kerttula, J; Golant, K; Pessa, M; Okhotnikov, O G

    2008-02-01

    We report a novel type of active fiber - tapered double clad fiber suitable for pumping by low brightness sources with large beam parameter product of 50/300 mm x mrad. Ytterbium double clad all-silica fiber (core/1(st) clad/2(nd) clad diameters 27/834/890 mum, NA(core)=0.11, NA(clad)=0.21), tapered down by a factor 4.8 for a length of 10.5 m was drawn from a preform fabricated by plasma chemical technologies. At a moderate Yb-ion concentration and 1:31 core/cladding ratio, the tapered double clad fiber demonstrates 0.9 dB/m pump absorption at 976 nm and excellent lasing slope efficiency. An ytterbium fiber laser with 84 W of output power and 92% slope efficiency, a 74 W superfluorescent source with 85% slope efficiency and amplifiers operating both in CW and pulsed regimes have been realized. All devices demonstrated robust single mode operation with a beam quality factor of M(2)=1.07. PMID:18542272

  3. 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.

  4. 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.

  5. Increasing corrosion resistance of carbon steels by surface laser cladding

    NASA Astrophysics Data System (ADS)

    Polsky, V. I.; Yakushin, V. L.; Dzhumaev, P. S.; Petrovsky, V. N.; Safonov, D. V.

    2016-04-01

    This paper presents results of investigation of the microstructure, elemental composition and corrosion resistance of the samples of low-alloy steel widely used in the engineering, after the application of laser cladding. The level of corrosion damage and the corrosion mechanism of cladded steel samples were established. The corrosion rate and installed discharge observed at the total destruction of cladding were obtained. The regularities of structure formation in the application of different powder compositions were obtained. The optimal powder composition that prevents corrosion of samples of low-carbon low-alloy steel was established.

  6. 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.

  7. 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.

  8. 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).

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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

  14. Chalcogenide optical microwires cladded with fluorine-based CYTOP.

    PubMed

    Li, Lizhu; Abdukerim, Nurmemet; Rochette, Martin

    2016-08-22

    We demonstrate optical transmission results of highly nonlinear As2Se3 optical microwires cladded with fluorine-based CYTOP, and compare them with microwires cladded with typical hydrogen-based polymers. In the linear optics regime, the CYTOP-cladded microwire transmits light in the spectral range from 1.3 µm up to >2.5 µm without trace of absorption peaks such as those observed using hydrogen-based polymer claddings. The microwire is also pumped in the nonlinear optics regime, showing multiple-orders of four-wave mixing and supercontinuum generation spanning from 1.0 µm to >4.3 µm. We conclude that with such a broadband transparency and high nonlinearity, the As2Se3-CYTOP microwire is an appealing solution for nonlinear optical processing in the mid-infrared. PMID:27557174

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Supercontinuum Generation in a Microstructured Fiber with an Irregular Cladding

    NASA Astrophysics Data System (ADS)

    Minkovich, V. P.; Sotsky, A. B.; Vaca Pereira G., M.; Dzen, I. S.; Sotskaya, L. I.

    2016-05-01

    A broad-band supercontinuum generation was obtained at excitation of a microstructured optical fiber with an irregular cladding by femtosecond laser pulses. To explain the experimental data, calculations of the mode characteristics of microstructured fibers were performed. It was shown that the creation of air channels with different radii in the fiber cladding makes it possible to involve both the fundamental and high fiber modes in the supercontinuum generation that helps to increase the width of the generation spectrum.

  1. Zircoloy Cladding Oxidation Simulation for LWR under LOCA Conditions

    2003-04-25

    PRECIP-2 simulates zircaloy cladding oxidation under LOCA conditions of LWR’s. The code calculates oxygen concentration distribution across the cladding wall by solving the diffusion equation with moving boundary conditions, taking into account the structure change of the beta— phase, i.e. alpha precipitation during the cooling period. The code also predicts total oxygen uptake, thicknesses of alpha, beta and oxide layers.

  2. 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.

  3. 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.

  4. HYDRIDE-RELATED DEGRADATION OF SNF CLADDING UNDER REPOSITORY CONDITIONS

    SciTech Connect

    K. McCoy

    2000-12-12

    The purpose and scope of this analysis/model report is to analyze the degradation of commercial spent nuclear fuel (CSNF) cladding under repository conditions by the hydride-related metallurgical processes, such as delayed hydride cracking (DHC), hydride reorientation and hydrogen embrittlement, thereby providing a better understanding of the degradation process and clarifying which aspects of the process are known and which need further evaluation and investigation. The intended use is as an input to a more general analysis of cladding degradation.

  5. Analysis of the magnetic corrosion product deposits on a boiling water reactor cladding

    SciTech Connect

    Orlov, Andrey; Degueldre, Claude; Kaufmann, Wilfried

    2013-01-15

    The buildup of corrosion product deposits (CRUD) on the fuel cladding of the boiling water reactor (BWR) before and after zinc injection has been investigated by applying local experimental analytical techniques. Under the BWR water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}] spinel solid solutions. X-ray absorption spectroscopy (XAS) revealed inversion ratios of cation distribution in spinels deposited from the solid solution. Based on this information, a two-site ferrite spinel solid solution model is proposed. Electron probe microanalysis (EPMA) and extended X-ray absorption fine structure (EXAFS) findings suggest the zinc-rich ferrite spinels formation on BWR fuel cladding mainly at lower pin. - Graphical Abstract: Analysis of spinels in corrosion product deposits on boiling water reactor fuel rod. Combining EPMA and XAFS results: schematic representation of the ferrite spinels in terms of the end members and their extent of inversion. Note that the ferrites are represented as a surface between the normal (upper plane, M[Fe{sub 2}]O{sub 4}) and the inverse (lower plane, Fe[MFe]O{sub 4}). Actual compositions red Black-Small-Square for the specimen at low elevation (810 mm), blue Black-Small-Square for the specimen at mid elevation (1800 mm). The results have an impact on the properties of the CRUD material. Highlights: Black-Right-Pointing-Pointer Buildup of corrosion product deposits on fuel claddings of a boiling water reactor (BWR) are investigated. Black-Right-Pointing-Pointer Under BWR water conditions, Zn addition with Ni and Mn induced formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}]. Black-Right-Pointing-Pointer X-Ray Adsorption Spectroscopy (XAS) revealed inversion of cations in spinel solid solutions. Black-Right-Pointing-Pointer Zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations.

  6. 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.

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

    NASA Astrophysics Data System (ADS)

    Cabellos, O.; Sanz, J.; Rodríguez, A.; González, E.; Embid, M.; Alvarez, F.; Reyes, S.

    2005-05-01

    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.

  8. Sensitivity and Uncertainty Analysis to Burn-up Estimates on ADS Using ACAB Code

    SciTech Connect

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

    2005-02-11

    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 reevaluation of some uncertainty XSs for ADS.

  9. 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.

  10. 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.

  11. 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.

  12. Residual Stress Measurements of Explosively Clad Cylindrical Pressure Vessels

    SciTech Connect

    Taylor, Douglas J; Watkins, Thomas R; Hubbard, Camden R; Hill, M. R.; Meith, W. A.

    2012-01-01

    Tantalum refractory liners were explosively clad into cylindrical pressure vessels, some of which had been previously autofrettaged. Using explosive cladding, the refractory liner formed a metallurgical bond with the steel of the pressure vessel at a cost of induced strain. Two techniques were employed to determine the residual stress state of the clad steel cylinders: neutron diffraction and mechanical slitting. Neutron diffraction is typically nondestructive; however, due to attenuation along the beam path, the cylinders had to be sectioned into rings that were nominally 25 mm thick. Slitting is a destructive method, requiring the sectioning of the cylindrical samples. Both techniques provided triaxial stress data and useful information on the effects of explosive cladding. The stress profiles in the hoop and radial directions were similar for an autofrettaged, nonclad vessel and a clad, nonautofrettaged vessel. The stress profiles in the axial direction appeared to be different. Further, the data suggested that residual stresses from the autofrettage and explosive cladding processes were not additive, in part due to evidence of reverse yielding. The residual stress data are presented, compared and discussed.

  13. Phase transformations at steel/IN626 clad interfaces

    NASA Astrophysics Data System (ADS)

    Ayer, Raghavan; Mueller, R. R.; Leta, D. P.; Sisak, W. J.

    1989-04-01

    The microstructures of 4130 and 2.25Cr-1Mo steels clad to nickel base IN625 by welding and HIPing were examined by Analytical Electron Microscopy (AEM) and Secondary Ion Mass Spectroscopy (SIMS) to determine the interfacial microstructural characteristics which could affect their mechanical properties and corrosion resistance. The interface microstructures of the clads produced by the two methods were considerably different. The clad produced by welding was characterized by a low density of carbide precipitates confined to a very narrow region (˜1 μm) at the interface of ferrite and austenite. In addition, a thin region of untempered martensite was present at the interface which could affect its resistance to hydrogen embrittlement as well as other mechanical properties. The interface of the HIP clad composite contained several regions of distinct microstructural characteristics with widely varying densities of carbide precipitates. Relative to the clad produced by welding, extensive precipitation was observed both in the steel and in the IN625 at the interface, separated by a region free from precipitation. The extent of precipitation at the interface regions appears to be controlled essentially by the extent of carbon transport across the interface. The article describes the detailed analysis of the interface characteristics, and models are proposed to explain the microstructural evolution at the interface of the HIP and weld clad composites.

  14. 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).

  15. Corrosion of aluminum cladding under optimized water conditions

    SciTech Connect

    Gibbs, A.

    1992-07-08

    Experience at SRS, ORNL, BNL, and Georgia Institute of Technology involving irradiated aluminum clad fuel and target elements, as well as studies of non-irradiated aluminum indicate that some types of aluminum assemblies can be kept in a continually well-deionized water atmosphere for up to 25 years without problems. SRS experience ranges from 2.75 years for the L-1.1 charge kept in deionized D[sub 2]O[sup 1] to greater than 10 years for assemblies stored in the Receiving Basin for Off-site Fuel (RBOF)[sup 2]. Experience at Georgia Institute of Technology reactor in Atlanta yielded the longest value of 25 years without problems. The common denominators in all of the reports is that the water is continually deionized to approximately 2 M[Omega] (2 [times] 10[sup 6]ohms) resistivity and the containers for the water are stainless steel or other non-porous material. This resistivity value is equivalent to a value of 0.5 micromhos or microSiemens conductivity and is reagent grade II quality water.[sup 3] 4 tabs, 26 refs.

  16. Corrosion of aluminum cladding under optimized water conditions

    SciTech Connect

    Gibbs, A.

    1992-07-08

    Experience at SRS, ORNL, BNL, and Georgia Institute of Technology involving irradiated aluminum clad fuel and target elements, as well as studies of non-irradiated aluminum indicate that some types of aluminum assemblies can be kept in a continually well-deionized water atmosphere for up to 25 years without problems. SRS experience ranges from 2.75 years for the L-1.1 charge kept in deionized D{sub 2}O{sup 1} to greater than 10 years for assemblies stored in the Receiving Basin for Off-site Fuel (RBOF){sup 2}. Experience at Georgia Institute of Technology reactor in Atlanta yielded the longest value of 25 years without problems. The common denominators in all of the reports is that the water is continually deionized to approximately 2 M{Omega} (2 {times} 10{sup 6}ohms) resistivity and the containers for the water are stainless steel or other non-porous material. This resistivity value is equivalent to a value of 0.5 micromhos or microSiemens conductivity and is reagent grade II quality water.{sup 3} 4 tabs, 26 refs.

  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. Revised Burnup Code System SWAT: Description and Validation Using Postirradiation Examination Data

    SciTech Connect

    Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide

    2002-05-15

    The burnup code system Step-Wise Burnup Analysis Code System (SWAT) is revised for use in a burnup credit analysis. An important feature of the revised SWAT is that its functions are achieved by calling validated neutronics codes without any changes to the original codes. This feature is realized with a system function of the operating system, which allows the revised SWAT to be independent of the development status of each code.A package of the revised SWAT contains the latest libraries based on JENDL-3.2 and the second version of the JNDC FP library. These libraries allow us to analyze burnup problems, such as an analysis of postirradiation examination (PIE), using the latest evaluated data of not only cross sections but also fission yield and decay constants.Another function of the revised SWAT is a library generator for the ORIGEN2 code, which is one of the most reliable burnup codes. ORIGEN2 users can obtain almost the same results with the revised SWAT using the library prepared by this function.The validation of the revised SWAT is conducted by calculation of the Organization for Economic Cooperation and Development/Nuclear Energy Agency burnup credit criticality safety benchmark Phase I-B and analyses of PIE data for spent fuel from Takahama Unit 3. The analysis of PIE data shows that the revised SWAT can predict the isotopic composition of main uranium and plutonium with a deviation of 5% from experimental results taken from UO{sub 2} fuels of 17 x 17 fuel assemblies. Many results of fission products including samarium are within a deviation of 10%. This means that the revised SWAT has high reliability to predict the isotopic composition for pressurized water reactor spent fuel.

  19. 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?

  20. 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?

  1. 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.

  2. 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

    ... the structures, systems, and components of the dry cask storage systems for high burnup spent nuclear... of dry storage cask systems and experiments would be beneficial. A draft test plan for the High... Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research...

  3. International studies on burnup credit criticality safety by an OECD/NEA working group

    SciTech Connect

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

    1998-11-01

    The results and conclusions from a six-year study by an international benchmarking group in the comparison of computational methods for evaluating burnup credit in criticality safety analyses is presented. Approximately 20 participants from 12 countries have provided results for most problems. Four detailed benchmark problems for pressurized-water-reactor fuel have been completed. Results from work being finalized, addressing burnup credit for boiling-water-reactor fuel, are discussed, as well as planned activities for additional benchmarks, including mixed-oxide fuels, and other activities.

  4. 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.

  5. 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.

  6. METHOD OF MANUFACTURE OF METAL ENCASED CORE MATERIAL

    DOEpatents

    Peters, E.J.

    1963-06-01

    A method of making reactor fuel elements in which the fissionable material is encased and sealed in steel or aluminum cladding having an enclosed channel from the fissionable material to the surface of the cladding at one end is described. Heat and pressure sufficient to bond the assembled fuel element are applied in a nonoxidizing atmosphere. (AEC)

  7. 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.

  8. Vanadium diffusion coating on HT-9 cladding for mitigating the fuel cladding chemical interactions

    NASA Astrophysics Data System (ADS)

    Lo, Wei-Yang; Yang, Yong

    2014-08-01

    Fuel cladding chemical interaction (FCCI) has been identified as one of the crucial issues for developing Ferritic/Martensitic (F/M) stainless steel claddings for metallic fuels in a fast reactor. The anticipated elevated temperature and high neutron flux can significantly aggravate the FCCI, in terms of formation of inter-diffusion and lower melting point eutectic phases. To mitigate the FCCI, vanadium carbide coating as a diffusion barrier was deposited on the HT-9 substrate using a pack cementation diffusion coating (PCDC) method, and the processing temperature was optimized down to 730 °C. A solid metallurgical bonding between the coating layer and substrate was achieved, and the coating is free from through depth cracks. The microstructural characterizations using SEM and TEM show a nanostructured grain structure. EDS/WDS and XRD analysis confirm the phase of coating layer as V2C. Diffusion couple tests at 660 °C for 100 h demonstrate that V2C layer with a thickness of less than 5 μm can effectively eliminate the inter-diffusion between the lanthanide cerium and HT-9 steel.

  9. Influence of texture on iodine-induced stress corrosion cracking of Zircaloy-4 cladding tubes

    NASA Astrophysics Data System (ADS)

    Schuster, I.; Lemaignan, C.

    1992-07-01

    A specific study was carried out to measure the influence of texture on the behaviour of Zircaloy-4 under iodine-induced stress corrosion cracking. The aim was to determine the relative effects of various metallurgical parameters involved in fuel rod fracture by pellet-clad interaction (PCI). Cladding tubes of different geometries were manufactured from a given Zircaloy-4 ingot. In this way tubes with different textures were obtained. Rings from these tubes were then subjected to slow tensile tests in an inert atmosphere and in an iodine vapour atmosphere. The sensitivity of the tubes to stress corrosion cracking is quantified by the loss of ductility on fracture between the tests in each atmosphere. Combined with the findings of other studies, the results showed that: (a) texture has a strong effect on the stress corrosion cracking behaviour of Zircaloy-4, (b) the mechanical properties do not have any bearing on the material behaviour under stress corrosion cracking, and that the better behaviour of a recrystallized material — compared to the same material in a stress-relieved state — can be explained solely by the texture effect, (c) texture is a more important parameter than chemical composition of Zircaloy-4, on condition that this composition remains within the ASTM specification. The conflict between the various mechanisms involved in stress corrosion crack propagation may explain these observations. Preliminary extrapolation of these conclusions to the irradiated material shows that a more specific study is needed using appropriate parameters.

  10. 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.

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

    DOE PAGESBeta

    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 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

  12. BOXER: Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations

    2010-02-01

    Neutron transport, calculation of multiplication factor and neutron fluxes in 2-D configurations: cell calculations, 2-D diffusion and transport, and burnup. Preparation of a cross section library for the code BOXER from a basic library in ENDF/B format (ETOBOX).

  13. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-03-01

    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. 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.

  15. 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.

  16. Microhardness and Young's modulus of high burn-up UO2 fuel

    NASA Astrophysics Data System (ADS)

    Cappia, F.; Pizzocri, D.; Marchetti, M.; Schubert, A.; Van Uffelen, P.; Luzzi, L.; Papaioannou, D.; Macián-Juan, R.; Rondinella, V. V.

    2016-10-01

    Vickers microhardness (HV0.1) and Young's modulus (E) measurements of LWR UO2 fuel at burn-up ≥60 GWd/tHM are presented. Their ratio HV0.1/E was found constant in the range 60-110 GWd/tHM. From the ratio and the microhardness values vs porosity, the Young's modulus dependence on porosity was derived and extended to the full radial profile, including the high burn-up structure (HBS). The dependence is well represented by a linear correlation. The data were compared to fuel performance codes correlations. A burn-up dependent factor was introduced in the Young's modulus expression. The modifications extend the experimental validation range of the TRANSURANUS correlation from un-irradiated to irradiated UO2 and up to 20% porosity. First simulations of LWR fuel rod irradiations were performed in order to illustrate the impact on fuel performance. In the specific cases selected, the simulations suggest a limited effect of the Young's modulus decrease due to burn-up on integral fuel performance.

  17. 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.

  18. Regularities of crystallographic texture formation in cladding tubes from Zr-based alloys during their production

    NASA Astrophysics Data System (ADS)

    Isaenkova, M.; Perlovich, Yu; Fesenko, V.

    2016-04-01

    This paper summarizes researches of the authors, which are directed on the development of the methodological basis of X-ray studies in the materials science of zirconium and on the systematization of new experimental results obtained using developed methods. The paper describes regularities of the formation of the crystallographic texture and the substructure inhomogeneity of cladding tubes from zirconium alloys at various stages of their manufacturing, i.e. during hot and cold deformation, recrystallization, phase transformations and interactions of the above processes.

  19. 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.

  20. 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

  1. Preparation of Plasma Cladding Gradient Wear-Resistant Layer and Study on Its Impact Fatigue Properties

    NASA Astrophysics Data System (ADS)

    Zhang, Dekun; Liu, Yuan; Yin, Yan

    2016-02-01

    Plasma cladding technology is used to prepare plasma cladding gradient wear-resistant specimens, and the performance of these specimens is analyzed and compared with those of single cladding specimens. The results indicate that plasma cladding gradient wear-resistant layers implement the gradient changes in microstructure and hardness from the surface of the outer cladding layer to the fusion line and that the outer and inner cladding layers are well combined, the inner cladding layer can improve rapid decreases in hardness of single wear-resistant samples from the cladding layer to the matrix, changes in hardness from the outer to inner cladding layer are buffered, and the inner cladding layer performs important functions in the transition between the outer cladding layer and substrate. The highest hardness of the outer layer, which reaches 735 HV0.1, is approximately 3.9 times that of the matrix. The impact fatigue resistance performance of the plasma gradient cladding specimens is superior to that of single cladding specimens, and fatigue cracks begin to form only after 1 × 105 cyclical impacts.

  2. 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)

  3. Parametric Study of Control Rod Exposure for PWR Burnup Credit Criticality Safety Analyses

    SciTech Connect

    Sanders, C.E.

    2001-07-20

    The Interim Staff Guidance on burnup credit (ISG-8) for pressurized water reactor (PWR) spent nuclear fuel (SNF), issued by the Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office, recommends the use of analyses that provide an ''adequate representation of the physics'' and notes particular concern with the ''need to consider the more reactive actinide compositions of fuels burned with fixed absorbers or with control rods fully or partly inserted.'' In the absence of readily available information on the extent of control rod (CR) usage in U.S. PWRs and the subsequent reactivity effect of CR exposure on discharged SNF, NRC staff have indicated a need for greater understanding in these areas. In response, this paper presents results of a parametric study of the effect of CR exposure on the reactivity of discharged SNF for various CR designs (including Axial Power Shaping Rods), fuel enrichments, and exposure conditions (i.e., burnup and axial insertion). The study is performed in two parts. In the first part, two-dimensional calculations are performed, effectively assuming full axial CR insertion. These calculations are intended to bound the effect of CR exposure and facilitate comparisons of the various CR designs. In the second part, three-dimensional calculations are performed to determine the effect of various axial insertion conditions and gain a better understanding of reality. The results from the study demonstrate that the reactivity effect increases with increasing CR exposure (e.g., burnup) and decreasing initial fuel enrichment (for a fixed burnup). Additionally, the results show that even for significant burnup exposures, minor axial CR insertions (e.g., < 20 cm) result in an insignificant effect on the k{sub eff} of a spent fuel cask.

  4. 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.

  5. 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).

  6. 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.

  7. 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.

  8. Determination of curie content and {sup 134/137}cesium ratios by gamma spectroscopy of high burnup plutonium-aluminum fuel assemblies

    SciTech Connect

    Haggard, D.L.; Tanner, J.E.

    1997-06-01

    Nondestructive assay (NDA) gamma spectroscopy techniques were used to measure {sup 134/137}Cs ratios on nine PuAl Mark 42 fuel assemblies. The purpose of the ratio measurement was to confirm theoretical burnup calculations. {sup 134/137}Cs ratios were determined from the measured activity based on corrected net peak area counts for the 605 keV peak from {sup 134}Cs and the 662 keV peak from {sup 137}Cs/{sup 137m}Ba. Assembly No. 2 {sup 134/137}Cs ratio measured on 4-15-92 was 0.19. The measured {sup 134/137}Cs ratio was decay corrected to be 2.11 on 8-1-84 based on the half lives of {sup 134}Cs and {sup 137}Cs. The measured {sup 134/137}Cs ratio range was 1.90--2.14 for all nine assemblies. These measured values compare to a theoretical ratio of 1.7 on 8-1-84 determined by burnup calculations. Total cesium curie content was also requested and determined using the NDA direct measurements. Gamma spectral data were measured on the nine sectioned Mark 42 fuel assemblies. Measured cesium curie content, decay corrected to 8-1-84, ranged from 18170--24480 curies of {sup 134}Cs and 8620--11646 curies of {sup 137}Cs. Theoretical cesium curie content of 8-1-84 was 15200 curies {sup 134}Cs and 8973 curies {sup 137}Cs. Direct assay cesium ratio is 12% to 26% higher than the predicted ratio of 1.7. The measured {sup 134}Cs data indicate between 20%--61% more activity than that predicted by the burnup code, whereas the measured {sup 137}Cs activity is between 4% less to 30% more than the predicted activity. This information may be used to address issues concerning criticality safety, storage, and shipping of this type of material.

  9. 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.

  10. Nanoscale light-matter interactions in atomic cladding waveguides

    NASA Astrophysics Data System (ADS)

    Stern, Liron; Desiatov, Boris; Goykhman, Ilya; Levy, Uriel

    2013-03-01

    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.

  11. 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

  12. Optical Properties Of Low Loss Fluoride Glass-Cladded Fibers

    NASA Astrophysics Data System (ADS)

    Burk, M. J.; Tran, D. C.; Fisher, C. F.; Levin, K. H.; Hart, Patricia; Sigel, G. H.

    1984-12-01

    Recent improvements in preform processing and fiber drawing techniques have resulted in glass-cladded fluoride glass fibers having losses under 10 dB/km. Multicomponent zirconium fluoride glass was used, and care was taken to reduce impurities such as transition metals and water. The reduction of scattering centers was also a major concern. Preforms were made using the rotational casting approach, which resulted in glass-cladded preforms having no observable core-clad defects. The preforms were coated with teflon, and drawn into fibers using an r.f. induction furnace. The optical attenuation of the fibers was measured in the infrared region. The minimum loss occurred around 2.5 microns. The fiber scattering loss was also measured. A variety of lasers were used for this measurement, including an infrared color-center laser to obtain scattering data directly in the infrared region.

  13. 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.

  14. 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. PMID:23462991

  15. Antifriction properties of gas thermal coatings of clad carbide powders

    SciTech Connect

    Borisov, Y.S.; Geleyshvili, T.P.; Gorbatov, L.N.; Molyar, A.G.; Kulgavyy, E.A.; Polishchuk, I.Y.

    1984-12-01

    A study is made of the antifriction properties of atomized coatings on VI-22 titanium alloy. Atomization was performed with carbides of titanium and chromium clad with cobalt, copper or nickel. Cladding was performed by the autoclave method by reducing copper, cobalt, and nickel from a solution of their salts with hydrogen under pressure. The clad coating thickness was 3 to 5 micro m. Electrolytically chrome plated titanium alloy and coatings of oxides were also tested for comparison. The structure of the coatings after atomization consisted of carbide particles uniformly distributed in a metal matrix. In friction the coatings worked in well, at relatively low coefficient of friction and slight wear. The most effective antifriction properties, minimum wear, and minimum coefficient of friction were achieved in coatings of composite TiC-Co and TiC-Cu powders.

  16. 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.

  17. Behaviour of fission gas in the rim region of high burn-up UO 2 fuel pellets with particular reference to results from an XRF investigation

    NASA Astrophysics Data System (ADS)

    Mogensen, M.; Pearce, J. H.; Walker, C. T.

    1999-01-01

    XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8-54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.

  18. Results of NDE Technique Evaluation of Clad Hydrides

    SciTech Connect

    Dennis C. Kunerth

    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

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. Electromechanical characterization of silver-clad BSCCO tapes.

    SciTech Connect

    Salib, S.; Iyer, A. N.; Vipulanandan, C.; Salama, K.; Balachandran, U.; Energy Technology; Univ. of Houston

    1998-01-01

    During the fabrication of silver-clad BSCCO tapes they are subjected to stresses which could lead to degradation in their current transport property. In the present investigation, studies were made to evaluate the electromechanical characteristics of silver-clad BSCCO conductors. The tensile strain tolerance characteristics of the monofilament, multifilament and composite (15 and 30% of Ag powder by volume) tapes were evaluated at 77 K. The average irreversible strain of monofilament and composite tapes were 0.19 and 0.47%, respectively. No noticeable improvement in strain tolerance was observed with the multifilament tapes. Detailed phase and microstructural analysis have been conducted using scanning electron microscopy.

  4. Efficient microfluidic photocatalysis in a symmetrical metal-cladding waveguide.

    PubMed

    Zhu, Shu; Dai, Hailang; Jiang, Bei; Shen, Zhenhua; Chen, Xianfeng

    2016-02-14

    In this paper, a symmetrical metal-cladding optical waveguide based microfluidic chip with a self-organized and free-standing TiO2 nanotube membrane was utilized to perform efficient photocatalysis. The chip has a microchannel bonded with TiO2 nanotube coated glass. The employment of microfluidic chip for hydrolysis reaction can enable the transfer of mass and photons. Moreover, the incorporation of the double metal-cladding waveguide enhances the light-matter interaction and effectively improves the efficiency of photocatalysis.

  5. 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.

  6. 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.

  7. Microstructure and corrosion resistance of TC2 Ti alloy by laser cladding with Ti/TiC/TiB2 powders

    NASA Astrophysics Data System (ADS)

    Diao, Yunhua; Zhang, Kemin

    2015-10-01

    In the present work, a TiC/TiB2 composite coating was produced onto a TC2 Ti alloy by laser cladding with Ti/TiC/TiB2 powders. The surface microstructure, phase components and compositions were characterized with methods of optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffractometry (XRD), and energy dispersive spectrometry (EDS). The cladding layer is consisted of Ti, TiC and TiB2. And the surface microhardness was measured. After laser cladding, a maximum hardness of 1100 HV is achieved in the laser cladding surface layer, which is more three times higher than that of the TC2 substrate (∼300 HV). Due to the formation of TiC and TiB2 intermetallic compounds in the alloyed region and grain refinement, the microhardness of coating is higher than TC2 Ti alloy. In this paper, the corrosion property of matrix material and treated samples were both measured in NaCl (3.5 wt%) aqueous solution. From the result we can see that the laser cladding specimens' corrosion property is clearly becoming better than that of the substrate.

  8. Spot weld attachment of thermocouples to a fuel rod cladding interior surface

    SciTech Connect

    Page, R.E.; Bates, S.O.; Pilger, J.P.

    1984-08-01

    Research was conducted by Pacific Northwest Laboratory to weld 0.020-inch-diameter thermocouples to the interior surface of Zircaloy 4 light-water reactor fuel cladding. Inconel sheathed Type K thermocouples were attached to fuel cladding to register cladding temperatures during loss-of-coolant accident testing. This report describes the development of welding parameters and the effects of thermocouple attachment on the burst strength and integrity of the cladding at temperatures up to 1550/sup 0/F.

  9. Evaluation of transient fuel pin cladding failure criteria for application to inherently safe LMFBR designs

    SciTech Connect

    Kramer, J M; DiMelfi, R J

    1984-03-01

    Purpose of report is to evaluate the methods for determining time-temperature-stress limits for cladding failure under accident conditions for inherently safe LMFBR designs. The range of expected thermal-mechanical cladding loading conditions is outlined for generic accident events, and application of existing mechanistic and empirical cladding failure models to these conditions is evaluated. The study is restricted to reference oxide fuel pins with austenitic stainless steel cladding.

  10. 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.

  11. A simple in-line fiber polarizer based on tapered flat-clad microfiber with a liquid cladding overlay

    NASA Astrophysics Data System (ADS)

    Lee, C.-L.; Chuang, H.-P.; Lai, K.-C.; Hsieh, I.-S.; Chang, T.-F.; Fan, J.-X.; Tsao, H.-H.; Hsu, K.-C.; Chen, N.-K.

    2008-11-01

    A novel and simple in-line fiber polarizer is presented. The proposed device is fabricated by tapering an anisotropic flat-cladding birefringent micro-fiber surrounding with low-dispersion optical-liquid cladding. We also presented a theoretical analysis for dispersive birefringence of flat-clad micro-fiber with liquid overlay. The proposed device can be useful as all-fiber polarizer for optical communications. Simulation results show the birefringence of the device can be enhanced when the aspect is larger. In the experimental measurement for polarization extinction ratio (PER) of the proposed device, a fiber-pigtailed 1549.25 nm DFB laser light was used as light source. The PER about 30 dB was demonstrated when the liquid with refractive index nD = 1.45 was used.

  12. Characteristics of a long-period fiber grating with reduced cladding for refractive index sensing

    NASA Astrophysics Data System (ADS)

    Chen, Haiyun; Gu, Zhengtian

    2011-10-01

    The sensitivity to surrounding refractive index (SRI) of a long-period fiber grating (LPFG) can be effectively improved by decreasing the cladding radius. When the cladding is reduced, a three-layer model is necessary to evaluate the effective refractive index (ERI) of the core mode. A variation of SRI can induce a greater resonant wavelength shift when the core mode is coupled to a higher-order cladding mode. However, as the cladding is reduced further, the highest-order cladding mode would be cut off, i.e. the number of cladding modes that a given fiber structure can support would be less; thus, the higher-order cladding modes that can be used for higher sensitivity are limited. Hence, the implementation of high sensitivity for SRI sensing with cladding-reduced LPFGs is dependent on the proper combination of cladding radius and cladding mode order. Based on the vector coupled-mode theory, the transmission spectrum and sensitivity are numerically analyzed with respect to the cladding radius, which shows that the SRI sensitivity of the HE12 mode with cladding radius a 2 = 20 µm is 32 times as high as that with a 2 = 62.5 µm and the SRI resolution is available to the order of 10-7.

  13. Using NDA Techniques to Improve Safeguards Metrics on Burnup Quantification and Plutonium Content in LWR SNF

    SciTech Connect

    Saavedra, Steven F; Charlton, William S; Solodov, Alexander A; Ehinger, Michael H

    2010-01-01

    Globally, there exists a long history in reprocessing in evaluation of the shipper/receiver difference (SRD) on spent nuclear fuel (SNF) received and processed. Typically, the declared shipper s values for uranium and plutonium in SNF (based on calculations involving the initial manufacturer s data and reactor operating history) are used as the input quantities to the head-end process of the facility. Problems have been encountered when comparing these values with measured results of the input accountability tank contents. A typical comparison yields a systematic bias indicated as a loss of 5 7 percent of the plutonium (Pu) and approximately 1 percent for the uranium (U). Studies suggest that such deviation can be attributed to the non-linear nature of the axial burnup values of the SNF. Oak Ridge National Laboratory and Texas A&M University are co-investigating the development of a new method, via Nondestructive Assay (NDA) techniques, to improve the accuracy in burnup and Pu content quantification. Two major components have been identified to achieve this objective. The first component calculates a measurement-based burnup profile along the axis of a fuel rod. Gamma-ray data is collected at numerous locations along the axis of the fuel rod using a High Purity Germanium (HPGe) detector designed for a wide range of gamma-ray energies. Using two fission products, 137Cs and 134Cs, the burnup is calculated at each measurement location and a profile created along the axis of the rod based on the individual measurement locations. The second component measures the U/Pu ratio using an HPGe detector configured for relatively low-energy gamma-rays including x-rays. Fluorescence x-rays from U and Pu are measured and compared to the U/Pu ratio determined from a destructive analysis of the sample. This will be used to establish a relationship between the measured and actual values. This relationship will be combined with the burnup analysis results to establish a relationship

  14. 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.

  15. 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.

  16. 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).

  17. 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.

  18. 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.

  19. Optimization of Hydride Rim Formation in Unirradiated Zr 4 Cladding

    SciTech Connect

    Shimskey, Rick W.; Hanson, Brady D.; MacFarlan, Paul J.

    2013-09-30

    The purpose of this work is to build on the results reported in the M2 milestone M2FT 13PN0805051, document number FCRD-USED-2013-000151 (Hanson, 2013). In that work, it was demonstrated that unirradiated samples of zircaloy-4 cladding could be pre-hydrided at temperatures below 400°C in pure hydrogen gas and that the growth of hydrides on the surface could be controlled by changing the surface condition of the samples and form a desired hydride rim on the outside diameter of the cladding. The work performed at Pacific Northwest National Laboratory since the issuing of the M2 milestone has focused its efforts to optimize the formation of a hydride rim on available zircaloy-4 cladding samples by controlling temperature variation and gas flow control during pre-hydriding treatments. Surface conditioning of the outside surface was also examined as a variable. The results of test indicate that much of the variability in the hydride thickness is due to temperature variation occurring in the furnaces as well as how hydrogen gas flows across the sample surface. Efforts to examine other alloys, gas concentrations, and different surface conditioning plan to be pursed in the next FY as more cladding samples become available

  20. Reform in Teacher Education through the CLAD/BCLAD Policy.

    ERIC Educational Resources Information Center

    Barreto, Ramona Maile

    1997-01-01

    Analyzes obstacles facing multicultural/bilingual teacher education reform in the context of California's Crosscultural Language and Academic Development (CLAD) or Bilingual Crosscultural Language and Academic Development (BCLAD) programs, which try to translate theoretical frameworks concerned with cultural difference into credentialing policy.…

  1. PERSPECTIVE WITH WEST PORTAL. THE BRIDGE IS CLAD IN HORIZONTAL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PERSPECTIVE WITH WEST PORTAL. THE BRIDGE IS CLAD IN HORIZONTAL CLAPBOARD SIDING AND HAS A SHEET METAL ROOF. NOTE THE TWO OPENINGS THAT RUN THE LENGTH OF THE BRIDGE; ONE IS AT THE EAVES AND THE OTHER IS ABOUT 4’ ABOVE THE DECK. - Dreibelbis Station Bridge, Spanning Maiden Creek, Balthaser Road (TR 745), Lenhartsville, Berks County, PA

  2. Direct Laser Cladding , Current Status and Future Scope of Application

    NASA Astrophysics Data System (ADS)

    Weisheit, A.; Gasser, A.; Backes, G.; Jambor, T.; Pirch, N.; Wissenbach, K.

    During the last decades Direct Laser Cladding has become an established technique in many industrial fields for applying wear and corrosion protection layers on metallic surfaces as well as for the repair of high value-added components. The most important application fields are die and tool making, turbine components for aero engines and power generation, machine components such as axes and gears, and oil drilling components. Continuous wave (CW) lasers with a power up to 18 kW are used on automated machines with three or more axes, enabling 3D cladding . The outstanding feature of DLC is the high precision which leads to a minimum heat input into the work piece and a very low distortion. Due to the high cooling rates a fine grained microstructure is achieved during solidification. A new development in laser cladding is micro cladding in a size range below 50 \\upmum especially for electronic and medical applications. Furthermore, additive manufacturing is coming again into focus as a clean and resource-efficient method to manufacture and modify functional prototypes as well as unique and small lot parts.

  3. 78 FR 7451 - Clad Steel Plate From Japan; Determination

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-01

    ..., 2012 (77 FR 5052) and determined on May 7, 2012 that it would conduct a full review (77 FR 37439, June..., 2012 (77 FR 38825). The hearing was held in Washington, DC, on December 6, 2012, and all persons who... COMMISSION Clad Steel Plate From Japan; Determination On the basis of the record \\1\\ developed in the...

  4. 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.

  5. An Innovative Ceramic Corrosion Protection System for Zircaloy Cladding

    SciTech Connect

    Ronald H. Baney, Dr. D. Butt, Dr. P. Demkowicz, Dr. G. Fuchs Department of Materials Science; James S. Tulenko, Department of Nuclear and Radiological Engineering; University of Florida.

    2003-02-19

    Light Water reactor (LWR) fuel performance is currently limited by thermal, chemical and mechanical constraints associated with the design, fabrication, and operation of the fuel in incore operation. Corrosion of the zirconium based (Zircaloy-4) alloy cladding of the fuel is a primary limiting factor. Recent success at the University of Florida in developing thin ceramic films with great adhesive properties for metal substrates offers an innovative breakthrough for eliminating a major weakness of the Zircaloy clad. ?The University of Florida proposes to coat the existing Zircaloy clad tubes with a ceramic coating for corrosion protection. An added bonus of this approach would be the implementation of a boron-containing burnable poison outer layer will also be demonstrated as part of the ceramic coating development. In this proposed effort, emphasis will be on the ceramic coating with only demonstration of feasibility on the burnable outer coating approach. This proposed program i s expected to give a step change (approximately a doubling) in clad lifetime before failure due to corrosion. In the development of ceramic coatings for Zircaloy-4 clad, silicon carbide and zirconium carbide coatings will first be applied to Zircaloy-4 coupons and cladding samples by thermal assisted chemical vapor deposition, plasma assisted chemical vapor deposition or by laser ablation deposition. All of these processes are in use at the University of Florida and have shown great potential. The questions of adhesion and thermal expansion mismatch of the ceramic coating to the Zircaloy substrate will be addressed. Several solutions to these conditions will be examined, if needed. These solutions include the use of a zirconium oxide compliant layer, employment of a laser roughened surface and the use of a gradient composition interlayer. These solutions have already been shown to be effective for other high modulus coatings on metal substrates. Mechanical properties and adhesion of the

  6. Stress Analysis and Structural Optimization of a Three-Layer Composite Cladding Tube Under Thermo-Mechanical Loads

    SciTech Connect

    S.-S. Zhou; X.-L. Gao; G. W. Griffith

    2012-07-01

    A general solution for the stress and strain fields in a three-layer composite tube subjected to internal and external pressures and temperature changes is first derived using thermo-elasticity. The material in each layer is treated as orthotropic, and the composite tube is regarded to be in a generalized plane strain state. A three-layer ZRY4-SiCf/SiCSiC composite cladding tube under a combined pressure and thermal loading is then analyzed and optimized by applying the general solution. The effects of temperature changes, applied pressures, and layer thickness on the mechanical behavior of the tube are quantitatively studied. The von Mises failure criterion for isotropic materials and the Tsai-Wu's failure theory for composites are used, respectively, to predict the failure behavior of the monolithic ZRY4 (i.e., Zircaloy-4) inner layer and SiC outer layer and the composite SiCf/SiC core layer of the three-layer tube. The numerical results reveal that the maximum radial and circumferential stresses in each layer always occur on the bonding surfaces. By adjusting the thickness of each layer, the effective stress in the three-layer cladding tube under the prescribed thermal-mechanical loading can be changed, thereby making it possible to optimally design the cladding tube.

  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. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

    SciTech Connect

    Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

    2005-05-24

    The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

  13. Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

    NASA Astrophysics Data System (ADS)

    Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

    2005-05-01

    The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

  14. Orientation-Dependent Displacement Sensor Using an Inner Cladding Fiber Bragg Grating.

    PubMed

    Yang, Tingting; Qiao, Xueguang; Rong, Qiangzhou; Bao, Weijia

    2016-01-01

    An orientation-dependent displacement sensor based on grating inscription over a fiber core and inner cladding has been demonstrated. The device comprises a short piece of multi-cladding fiber sandwiched between two standard single-mode fibers (SMFs). The grating structure is fabricated by a femtosecond laser side-illumination technique. Two well-defined resonances are achieved by the downstream both core and cladding fiber Bragg gratings (FBGs). The cladding resonance presents fiber bending dependence, together with a strong orientation dependence because of asymmetrical distribution of the "cladding" FBG along the fiber cross-section. PMID:27626427

  15. Evaluation of burnup credit for fuel storage analysis -- Experience in Spain

    SciTech Connect

    Conde, J.M.; Recio, M.

    1995-04-01

    Several Spanish light water reactor commercial nuclear power plants are close to maximum spent-fuel pool storage capacity. The utilities are working on the implementation of state-of-the-art methods to increase the storage capacity, including both changes in the pool design (recracking) and the implementation of new analysis approaches with reduced conservation (burnup credit). Burnup credit criticality safety analyses have been approved for two pressurized water reactor plants (four units) and one boiling water reactor (BWR); an other BWR storage analysis is being developed at this moment. The elimination of the ``fresh fuel assumption`` increases the complexity of the criticality analysis to be performed, sometimes putting into question the capability of the analytic tools to properly describe this new situation and increasing the scope of the scenarios to be analyzed. From a regulatory perspective, the reactivity reduction associated with burnup of the fuel can be given credit only if the exposure of each fuel bundle can be known with enough accuracy. Subcriticality of spent-fuel storage depends mainly on the initial fuel enrichment, storage geometry, fuel exposure history, and cooling time. The last two aspects introduced new uncertainties in the criticality analysis that should be quantified in an adequate way. In addition, each and every fuel bundle has its own specific exposure history, so that strong assumptions and simplified calculational schemes have to be developed to undertake the analysis. The Consejo de Seguridad Nuclear (CSN), Spanish regulatory authority on the matter of nuclear safety and radiation protection, plays an active role in the development of analysis methods to support burnup credit, making proposals that may be beneficial in terms of risk and cost while keeping the widest safety margins possible.

  16. 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.

  17. Method for adding additional isotopes to actinide-only burnup credit

    SciTech Connect

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

    1998-01-01

    The Topical Report on Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel Packages requires computer code validation to be performed against a benchmark set of chemical assays for isotopic concentration and against a benchmark set of critical experiments for package criticality. Both sets contain all the isotopes included in the methodology. The chemical assays used include the uranium and plutonium isotopes, while the critical experiments were composed of UO{sub 2} or MOX rods, covering the isotopes in the actinide only approach. Since other isotopes are not included in the validation benchmark sets, it would be necessary to justify both the content and worth of any additional isotope for which burnup credit is to be taken (i.e., both the concentration and criticality effect of each particular isotope must be validated). A method is proposed here that can be used for any number of additional isotopes. As does the actinide-only burnup credit methodology, this method makes use of chemical assay data to establish the conservatism in the prediction of each isotope`s concentration. Criticality validation is also performed using a benchmark set of UO{sub 2} and MOX critical experiments, where the additional isotopes are validated using worth experiments to conservatively account for any uncertainty in their cross sections. The remaining requirements (analysis and modeling parameters, loading criteria generation, and physical implementation and controls) are performed exactly as described in the actinide-only burnup credit methodology. This report provides insight into each particular requirement in the new methodology.

  18. 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.

  19. A Genesis breakup and burnup analysis in off-nominal Earth return and atmospheric entry

    NASA Technical Reports Server (NTRS)

    Salama, Ahmed; Ling, Lisa; McRonald, Angus

    2005-01-01

    The Genesis project conducted a detailed breakup/burnup analysis before the Earth return to determine if any spacecraft component could survive and reach the ground intact in case of an off-nominal entry. In addition, an independent JPL team was chartered with the responsibility of analyzing several definitive breakup scenarios to verify the official project analysis. This paper presents the analysis and results of this independent team.

  20. 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.

  1. 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.

  2. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Asiah, Nur; Shafii, M. Ali

    2010-12-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio.

  3. Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations

    DOE PAGESBeta

    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

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

    SciTech Connect

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

    2002-02-15

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

  5. 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.

  6. 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.

  7. Three Dimensional Analysis of 3-Loop PWR RCCA Ejection Accident for High Burnup

    SciTech Connect

    Marciulescu, Cristian; Sung, Yixing; Beard, Charles L.

    2006-07-01

    The Rod Control Cluster Assembly (RCCA) ejection accident is a Condition IV design basis reactivity insertion event for Pressurized Water Reactors (PWR). The event is historically analyzed using a one-dimensional (1D) neutron kinetic code to meet the current licensing criteria for fuel rod burnup to 62,000 MWD/MTU. The Westinghouse USNRC-approved three-dimensional (3D) analysis methodology is based on the neutron kinetics version of the ANC code (SPNOVA) coupled with Westinghouse's version of the EPRI core thermal-hydraulic code VIPRE-01. The 3D methodology provides a more realistic yet conservative analysis approach to meet anticipated reduction in the licensing fuel enthalpy rise limit for high burnup fuel. A rod ejection analysis using the 3D methodology was recently performed for a Westinghouse 3-loop PWR at an up-rated core power of 3151 MWt with reload cores that allow large flexibility in assembly shuffling and a fuel hot rod burnup to 75,000 MWD/MTU. The analysis considered high enrichment fuel assemblies at the control rod locations as well as bounding rodded depletions in the end of life, zero power and full power conditions. The analysis results demonstrated that the peak fuel enthalpy rise is less than 100 cal/g for the transient initiated at the hot zero power condition. The maximum fuel enthalpy is less than 200 cal/g for the transient initiated from the full power condition. (authors)

  8. 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.

  9. Cladding hull decontamination and densification process. Part 1. The prototype cladding hull decontamination system

    SciTech Connect

    Lambright, T.M.; Montgomery, D.R.

    1980-04-01

    A prototype system for decontaminating Zircaloy-4 cladding hulls has been assembled and tested at Pacific Northwest Laboratory. The decontamination process consists of treatment with a gaseous mixture of hydrogen fluoride (HF) and argon (Ar) followed by a dilute aqueous etch of ammonium oxalate, ammonium citrate, ammonium fluoride, and hydrogen peroxide. The continuous cleaning process described in this report successfully descaled small portions of most charges, but was unable to handle the original design capacity of 4 kg/hr because of problems in the following areas: control of HF reactor temperatures, regulation of HF and argon mixtures and flows, isolation of the HF reactor atmosphere from the aqueous washer/rinser atmosphere, regulation of undesirable side reactions, and control over hull transport through the system. Due to the limited time available to solve these problems, the system did not attain fully operational status. The work was performed with unirradiated hulls that simulated irradiated hulls. The system was not built to be remotely operable. The process chemistry and system equipment are described in this report with particular emphasis on critical operating areas. Recommendations for improved system operation are included.

  10. A Novel Method of Modeling the Deformation Resistance for Clad Sheet

    SciTech Connect

    Hu Jianliang; Yi Youping; Xie Mantang

    2011-08-22

    Because of the excellent thermal conductivity, the clad sheet (3003/4004/3003) of aluminum alloy is extensively used in various heat exchangers, such as radiator, motorcar air conditioning, evaporator, and so on. The deformation resistance model plays an important role in designing the process parameters of hot continuous rolling. However, the complex behaviors of the plastic deformation of the clad sheet make the modeling very difficult. In this work, a novel method for modeling the deformation resistance of clad sheet was proposed by combining the finite element analysis with experiments. The deformation resistance model of aluminum 3003 and 4004 was proposed through hot compression test on the Gleeble-1500 thermo-simulation machine. And the deformation resistance model of clad sheet was proposed through finite element analysis using DEFORM-2D software. The relationship between cladding ratio and the deformation resistance was discussed in detail. The results of hot compression simulation demonstrate that the cladding ratio has great effects on the resistance of the clad sheet. Taking the cladding ratio into consideration, the mathematical model of the deformation resistance for clad sheet has been proved to have perfect forecasting precision of different cladding ratio. Therefore, the presented model can be used to predict the rolling force of clad sheet during the hot continuous rolling process.

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

    SciTech Connect

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

    1999-12-01

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

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

    SciTech Connect

    George W. Griffith

    2011-10-01

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

  13. Assessment of a new fabrication route for Fe-9Cr-1W ODS cladding tubes

    NASA Astrophysics Data System (ADS)

    Toualbi, L.; Cayron, C.; Olier, P.; Malaplate, J.; Praud, M.; Mathon, M.-H.; Bossu, D.; Rouesne, E.; Montani, A.; Logé, R.; de Carlan, Y.

    2012-09-01

    Oxide Dispersion Strengthened ferritic/martensitic steels are developed as future cladding materials for Generation IV Sodium-Cooled Fast Reactors. ODS alloys are elaborated by powder metallurgy, consolidated by hot extrusion and manufactured into tube cladding using cold rolling process. ODS steels present low ductility and high hardness at room temperature which complicate their manufacturing. Cold working leads to the hardening of the tube which needs to be softened by heat treatment. A new high temperature fabrication route performed on a Fe-9Cr-1W-Ti-Y2O3-ODS martensitic steel has been designed by following the hardness values, the morphological and crystallographic anisotropy and the nano-precipitation size evolution at each step of the fabrication route. Observations show that phase transformation from ferrite (α) to austenite (γ) is crucial to reduce the morphological and the crystallographic anisotropy induced by the manufacturing processes. The high temperature heat treatments permit to make the austenitic grain grow leading to an improvement of the cold workability. Ultimate Tensile Strength values obtained in the hoop direction remain about 315 MPa at 650 °C which is slightly lower compared to other Fe-9Cr ODS tubes but the new microstructure could be more favorable for creep properties.

  14. Initial and Long-Term Movement of Cladding Installed Over Exterior Rigid Insulation

    SciTech Connect

    Baker, Peter

    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 will be 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 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. This research was an extension on previous research conducted by Building Science Corporation 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 year’s worth of cladding movement data from assemblies constructed in an exposed outdoor environment.

  15. Degradation of mechanical properties of stainless steel cladding due to neutron irradiation and thermal aging

    SciTech Connect

    Haggag, F.M.

    1994-09-01

    Thermal aging of three-wire series-arc stainless steel weld overlay cladding at 288{degrees}C for 1605 h resulted in an appreciable decrease (16%) in the Charpy V-notch (CVN) upper-shelf energy (USE), but the effect on the 41-J transition temperature shift was very small (3{degrees}C). The combined effect following neutron irradiation at 288{degrees}C to a fluence of 5 X 10{sup 19} neutrons/cm{sup 2} (>1 MeV) was a 22% reduction in the USE and a 29{degrees}C shift in the 41-J transition temperature. The effect of thermal aging on tensile properties was very small. However, the combined effect of irradiation and aging was an increase in the yield strength (6 to 34% at test temperatures from 288 to -125{degrees}C) and no apparent change in ultimate tensile strength or total elongation. Neutron irradiation reduced the initiation fracture toughness (J{sub {kappa}}) much more than did thermal aging alone. However, irradiation slightly decreased the tearing modulus but no reduction was caused by thermal aging alone. The effects of long-term thermal exposure times (20,000 and 50,000 h) will be investigated when the specimens become available. Also, long-term thermal exposure of the three-wire cladding as well as type 308 stainless steel weld materials at 343{degrees}C is in progress.

  16. Investigation of semiconductor clad optical waveguides

    NASA Technical Reports Server (NTRS)

    Batchman, T. E.; Mcwright, G. M.

    1982-01-01

    Glass waveguides are studied because of the ease and economy of fabricating devices in glass. All calculations are based on the assumption of a glass guide and substrate, but the effects being studied will occur on other materials if the proper refractive indices are used in the calculations.

  17. Er/Yb co-doped oxy-fluoride glass-ceramics core/polymer cladding optical fibers

    NASA Astrophysics Data System (ADS)

    Czerska, E.; Świderska, M.

    2014-11-01

    Erbium/ytterbium co-doped glasses can be applied as NIR laser sources (1.55 μm) or optical amplifiers in this range. About hundred meters of Er/Yb co-doped oxy-fluoride glass-ceramics fibers have been drawn from a glass preform followed by controlled annealing. Processing temperatures (drawing and annealing) were selected upon thermal analysis results (DTA/DSC plots). Glass-ceramic structure was confirmed by the XRD measurements. Obtained fibers show good optical properties. As a cladding material polymer material (acrylic resin) is considered due to its low deposition temperature and suitable value of refractive index.

  18. Filtration and Leach Testing for PUREX Cladding Sludge and REDOX Cladding Sludge Actual Waste Sample Composites

    SciTech Connect

    Shimskey, Rick W.; Billing, Justin M.; Buck, Edgar C.; Casella, Amanda J.; Crum, Jarrod V.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Hallen, Richard T.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Swoboda, Robert G.

    2009-03-02

    A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan (Barnes and Voke 2006). The test program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Hanford Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Under test plan TP RPP WTP 467 (Fiskum et al. 2007), eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. Under this test plan, a waste testing program was implemented that included: • Homogenizing the archive samples by group as defined in the test plan. • Characterizing the homogenized sample groups. • Performing parametric leaching testing on each group for compounds of interest. • Performing bench-top filtration/leaching tests in the hot cell for each group to simulate filtration and leaching activities if they occurred in the UFP2 vessel of the WTP Pretreatment Facility. This report focuses on a filtration/leaching test performed using two of the eight waste composite samples. The sample groups examined in this report were the plutonium-uranium extraction (PUREX) cladding waste sludge (Group 3, or CWP) and reduction-oxidation (REDOX) cladding waste sludge (Group 4, or CWR). Both the Group 3 and 4 waste composites were anticipated to be high in gibbsite, thus requiring caustic leaching. WTP RPT 167 (Snow et al. 2008) describes the homogenization, characterization, and parametric leaching activities before benchtop filtration/leaching testing of these two waste groups. Characterization and initial parametric data in that report were used to plan a single filtration/leaching test using a blend of both wastes. The test focused on filtration testing of the waste and caustic leaching for aluminum, in the form

  19. 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

  20. Building America Case Study: Initial and Long-Term Movement of Cladding Installed Over Exterior Rigid Insulation (Fact Sheet)

    SciTech Connect

    Not Available

    2014-10-01

    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.