WorldWideScience

Sample records for fuel cladding corrosion

  1. Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, J.C.

    1985-08-01

    Twenty-one LWBR irradiation test rods containing ThO/sub 2/-UO/sub 2/ fuel and Zircaloy cladding with holes or cracks operated successfully. Zircaloy cladding corrosion on the inside and outside diameter surfaces and hydrogen pickup in the cladding were measured. The observed outer surface Zircaloy cladding corrosion oxide thicknesses of the test rods were similar to thicknesses measured for nondefected irradiation test rods. An analysis model, which was developed to calculate outer surface oxide thickness of non-defected rods, gave results which were in reasonable agreement with the outer surface oxide thicknesses of defected rods. When the analysis procedure was modified to account for additional corrosion proportional to fission rate and to time, the calculated values agreed well with measured inner oxide corrosion film values. Hydrogen pickup in the defected rods was not directly proportional to local corrosion oxide weight gain as was the case for non-defected rods. 16 refs., 6 figs., 8 tabs.

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

    Science.gov (United States)

    Carr, James Patrick, IV

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

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

    Directory of Open Access Journals (Sweden)

    de Souza Gomes Daniel

    2016-01-01

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

  4. Galvanic corrosion of Mg-Zr fuel cladding and steel immobilized in Portland cement and geopolymer at early ages

    Science.gov (United States)

    Rooses, Adrien; Lambertin, David; Chartier, David; Frizon, Fabien

    2013-04-01

    Galvanic corrosion behaviour of Mg-Zr alloy fuel cladding and steel has been studied in Ordinary Portland cement and Na-geopolymer. Portland cements implied the worse magnesium corrosion performances due to the negative effects of cement hydrates, grinding agents and gypsum on the galvanic corrosion. Galvanic corrosion in Na-geopolymer paste remains very low. Silicates and fluoride from the geopolymer activation solution significantly improve the corrosion resistance of the magnesium alloy while coupling with a cathode.

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

    OpenAIRE

    de Souza Gomes Daniel; Abe Alfredo; Silva Antonio Teixeira e; Giovedi Claudia; Martins Marcelo Ramos

    2016-01-01

    In nuclear reactors, the use of stainless steel (SS) as the cladding material offers some advantages such as good mechanical and corrosion resistance. However, its main advantage is the reduction in the amount of the hydrogen released during loss-of-coolant accident, as observed in the Fukushima Daiichi accident. Hence, research aimed at developing accident tolerant fuels should consider SS as an important alternative to existing materials. However, the available computational tools used to a...

  6. Potential corrosion and degradation mechanisms of Zircaloy cladding on spent nuclear fuel in a tuff repository

    Energy Technology Data Exchange (ETDEWEB)

    Rothman, A.J.

    1984-09-01

    A literature review and analysis were made of corrosion and degradation processes applicable to Zircaloy cladding on spent nuclear fuel in a tuff repository. In particular, lifetime sought for the Zircaloy is 10,000 years. Among the potential failure mechanisms examined were: oxidation by steam, air, and water, including the effects of ions whose presence is anticipated in the water; mechanical overload; stress (creep) rupture; stress-corrosion cracking (SCC); and delayed failure due to hydride cracking. The conclusion is that failure due to oxidation is not credible, although a few experiments are suggested to confirm the effect of aqueous fluoride on the Zircaloy cladding. Mechanical overload is not a problem, and failure from stress-rupture does not appear likely based on a modified Larson-Miller analysis. Analysis shows that delayed hydride cracking is not anticipated for the bulk of spent fuel pins. However, for a minority of pins under high stress, there is some uncertainty in the analysis as a result of: (1) uncertainty about crack depths in spent fuel claddings and (2) the effect of slow cooling on the formation of radially oriented hydride precipitates. Experimental resolution is called for. Finally, insufficient information is currently available on stress-corrosion cracking. While evidence is presented that SCC failure is not likely to occur, it is difficult to demonstrate this conclusively because the process is not clearly understood and data are limited. Further experimental work on SCC susceptibility is especially needed.

  7. EPRI fuel cladding integrity program

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-01-01

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

  8. Evaluation of zinc addition on fuel cladding corrosion at the Halden test reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Kolstad, E.; Symons, W.J.; Bryhn-Integrigtsen, K.; Oberlaender, B.C.

    1996-08-01

    Experimental studies have shown that addition of zinc to a PWR environment reduces the general corrosion rates of materials in the primary system and delays the initiation of primary water stress corrosion cracking (PWSCC) in Alloy 600. In order to provide an early warning of any potential adverse effects on the fuel cladding, corrosion studies were initiated at the Halden test reactor. These tests were carried out in a PWR rig inserted in the Halden reactor core. The rig simulated thermal hydraulic and coolant conditions typical of a MR. It had two flow channels where the fuel rod segments were exposed to the coolant under irradiation flux. Selected pre-characterized rodlets with fresh and pre-irradiated standard and low-tin Zircaloy-4 material were irradiated for three cycles. First cycle lasted for 110 effective full power days (EFPDs), the second for 95 EFPDs and the last 62 EFPDs. The cladding corrosion behavior was monitored by initial, interim and final oxide thickness measurements by eddy current lift-off probe. Crud sampling was performed in both channels after cycle 1 and 2. Destructive post-irradiation examinations (PIE) of two rodlets, irradiated during cycle 1 and 2, have also been completed at the conclusion of the in-pile testing. This report presents the results on oxide thickness measurements, irradiation history and water chemistry data, and the PIE.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Corrosion of the AlFeNi alloy used for the fuel cladding in the Jules Horowitz research reactor

    Science.gov (United States)

    Wintergerst, M.; Dacheux, N.; Datcharry, F.; Herms, E.; Kapusta, B.

    2009-09-01

    The AlFeNi aluminium alloy (1 wt% Fe, 1 wt% Ni, 1 wt% Mg) is expected to be used as nuclear fuel cladding for the Jules Horowitz experimental reactor. To guarantee a safe behaviour of the fuel, a good understanding of the fuel clad corrosion mechanisms is required. In this field, the experimental characterization of the selected alloy was performed. Then experimental studies of the aluminium alloy corrosion product obtained in autoclaves have shown an oxide film composed of two layers. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion parallel to a dissolution-precipitation process to form the outer zone. Dynamic experiments at 70 °C have demonstrated that a solid diffusion step controls the release kinetic. Then post-irradiation exams performed on irradiated fuel plates were used to investigate the effects of the irradiation on the corrosion behaviour in the reactor core.

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

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J.

    2014-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-31

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

  14. Evaluation of aluminum-clad spent fuel corrosion in Argentine basins

    Energy Technology Data Exchange (ETDEWEB)

    Haddad, R.; Loberse, A.N.; Semino, C.J.; Guasp, R. [CNEA, Buenos Aires (Argentina)

    2001-07-01

    An IAEA sponsored Coordinated Research Program was extended to study corrosion effects in several sites. Racks containing Aluminum samples were placed in different positions of each basin and periodic sampling of all the waters was performed to conduct chemical analysis. Different forms of corrosion have been encountered during the programme. In general, the degree of degradation is inversely proportional to the purity of the water. Maximum pit depths after 2 years of exposure are in the range of 100-200 {mu}m. However, sediments deposited on the coupon surfaces seem to be responsible for the developing of large pits (1-2 mm in diameter). In many cases, what appears to be iron oxide particles were found originated by the corrosion of carbon steel components present elsewhere in the basin. These results correlate with observations made on the fuel itself, during exhaustive visual inspection. (author)

  15. Nuclear fuel elements having a composite cladding

    Science.gov (United States)

    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.

  16. Fuel pin cladding

    Science.gov (United States)

    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.

  17. Fuel pin cladding

    Science.gov (United States)

    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.

  18. Evaluation of alternative descriptions of PWR cladding corrosion behavior

    Energy Technology Data Exchange (ETDEWEB)

    Quecedo, M.; Serna, J. J.; Weiner, R. A.; Kersting, P. J.

    1999-05-15

    A statistical procedure has been used to evaluate several alternative descriptions of pressurized water reactor (PWR) cladding corrosion behavior, using an extensive database of Improved (low tin) Zr-4 cladding corrosion measurements from fuel irradiated in commercial PWRs. The in-reactor corrosion enhancement factors considered in the model development are based on a comprehensive review of the current literature for PWR cladding corrosion phenomenology and models. In addition, because prediction of PWR cladding corrosion behavior is very sensitive to the values used for the oxide surface temperatures, several models for the forced convection and sub-cooled nucleate boiling (SNB) coolant heat transfer under PWR conditions have also been evaluated. This evaluation determined that the choice of the forced convection heat transfer has the greatest impact on the ability to fit the data. In addition, the SNB heat transfer model used must account for a continuous transition from forced convection conditions to fully developed SNB conditions. With these choices for the heat transfer models, the evaluation determined that the significant in-reactor corrosion enhancement factors are related to the formation of a hydride rim at the cladding outer diameter, the coolant lithium concentration, and the fast neutron fluence (author) (ml)

  19. Coupon Surveillance For Corrosion Monitoring In Nuclear Fuel Basin

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-01

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

  20. Fuel clad chemical interactions in fast reactor MOX fuels

    Science.gov (United States)

    Viswanathan, R.

    2014-01-01

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

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

    Science.gov (United States)

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

    2010-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  3. Development Status of Accident Tolerant Fuel Cladding for LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Hydrogen explosions and the release of radionuclides are caused by severe damage of current nuclear fuels, which are composed of fuel pellets and fuel cladding, during an accident. To reduce the damage to the public, the fuels have to enhance their integrity under an accident environment. Enhanced accident tolerance fuels (ATFs) can tolerate a loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations as well as operational transients, in comparison with the current UO{sub 2}-Zr alloy system used in the LWR. Surface modified Zr cladding as a new concept was suggested to apply an enhanced ATF cladding. The aim of the partial ODS treatment is to increase the high-temperature strength to suppress the ballooning/rupture behavior of fuel cladding during an accident event. The target of the surface coating is to increase the corrosion resistance during normal operation and increase the oxidation resistance during an accident event. The partial ODS treatment of Zircaloy-4 cladding can be produced using a laser beam scanning method with Y2O3 powder, and the surface Cr-alloy and Cr/FeCrAl coating on Zircaloy-4 cladding can be obtained after the development of 3D laser coating and arc ion plating technologies.

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

    Science.gov (United States)

    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

  5. Accident-tolerant oxide fuel and cladding

    Energy Technology Data Exchange (ETDEWEB)

    Mariani, Robert D.

    2017-05-30

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

  6. Preliminary Modeling of Corrosion/Oxidation Properties of CrAl Alloy-coated Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jong-Dae; Kim, Hyo Chan; Shin, Chang Hwan; Yang, Yong Sik; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Accident tolerant fuel (ATF) cladding has been being developed globally after the Fukushima accident with the demands for the nuclear fuel having higher safety at normal operation conditions as well as even in a severe accident conditions. Korea Atomic Energy Research Institute (KAERI) has been developed some of remarkable ATF cladding candidates. They showed a superior oxidation/corrosion resistance in water and steam conditions to the commercial Zr alloys and totally different behaviors from commercial Zr alloys. Prior to evaluate entire fuel performance of newly developed CrAl alloy cladding by KAERI collectively, preliminary model of water-side corrosion and high temperature oxidation model were proposed. They were highly consistent with experimental results. Also this model is useful for the quantitative analysis with given with relative superior characteristics to existing commercial fuel claddings.

  7. Advanced Fuels Campaign Cladding & Coatings Meeting Summary

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-03-01

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

  8. Development of advanced LWR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Corrosion Minimization for Research Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Eric Shaber; Gerard Hofman

    2005-06-01

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  11. Double-clad nuclear fuel safety rod

    Science.gov (United States)

    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.

  12. Protection of spent aluminum-clad research reactor fuels during extended wet storage

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, Stela M.C.; Correa, Olandir V.; Souza, Jose A.; Ramanathan, Lalgudi V., E-mail: lalgudi@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Antunes, Renato A. [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Ramanathan, Lalgudi V. [Electrocell Ind. Com. Equip. Elet. LTDA (CIETEC), Sao Paulo, SP (Brazil)

    2013-07-01

    Aluminum-clad spent nuclear fuel from research reactors (RR) is stored in light water filled pools or basins worldwide. Many incidences of pitting corrosion of the fuel cladding has been reported and attributed to synergism in the effect of certain water parameters. Protection of spent Al-clad RR fuel with a conversion coating was proposed in 2008. Preliminary results revealed increased pitting corrosion resistance of cerium oxide coated aluminum alloys AA 1050 and AA 6061, used as RR fuel plate cladding. Further development of conversion coatings for Al alloys was carried out and this paper presents: (a) the preparation and characterization of hydrotalcite (HTC) coatings; (b) the results of laboratory tests in which the corrosion behavior of coated Al alloys in NaCl solutions was determined; (c) the results of field tests in which un-coated, boehmite coated, HTC coated and cerium modified boehmite / HTC coated AA 1050 and AA 6061 coupons were exposed to the IEA-R1 reactor spent fuel basin for extended periods. In these field tests the coupons coated with HTC from a high temperature (HT) bath and subsequently modified with Ce were the most resistant to pitting corrosion. In laboratory tests also, HT- hydrotalcite + Ce coated specimens were the most corrosion resistant in 0.01 M NaCl. The role of cerium in increasing the corrosion resistance imparted by the different conversion coatings of spent Al-clad RR fuel elements is presented. (author)

  13. Improvement in PCI property of PWR fuel cladding by texture control

    Energy Technology Data Exchange (ETDEWEB)

    Inoue, S. (Kansai Electric Power Co., Inc., Osaka (Japan)); Abeta, S.; Ozawa, M.; Takahashi, T.

    1993-09-01

    Effects of texture on out-of-pile Stress Corrosion Cracking (SCC) resistance in Zircaloy fuel cladding tube and the Pellet-Clad Interaction (PCI) property of a fuel rod using texture controlled cladding tube under power ramp conditions are described. The cladding tube with radial texture, which means that the c-axis of hcp crystal of Zr is highly concentrated in the radial direction of the tube, showed excellent performance in out-of-pile SCC tests and power ramp tests. (author).

  14. Clad thickness variation N-Reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Smith, E.A.

    1966-05-12

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

  15. CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION

    Directory of Open Access Journals (Sweden)

    Dávid Halabuk

    2015-12-01

    Full Text Available The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.

  16. Preliminary study of mechanical behavior for Cr coated Zr-4 Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do-Hyoung; Kim, Hak-Sung [Hanyang Univ., Seoul (Korea, Republic of); Kim, Hyo-Chan; Yang, Yong-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    To decrease the oxidation rate of Zr-based alloy components, many concepts of accident tolerant fuel (ATF) such as Mo-Zr cladding, SiC/SiCf cladding and iron-based alloy cladding are under development. One of the promised concept is the coated cladding which can remarkably increase the corrosion and wear resistance. Recently, KAERI is developing the Cr coated Zircaloy cladding as accident tolerance cladding. To coat the Cr powder on the Zircaloy, 3D laser coating technology has been employed because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. Therefore, for this work, the mechanical integrity of Cr coated Zircaloy should be evaluated to predict the safety of fuel cladding during the operating or accident of nuclear reactor. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr, which were referred from the literatures and experimental reports. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr. The pellet-clad mechanical interaction (PCMI) properties of Cr coated Zr-4 cladding were investigated by thermo-mechanical finite element analysis (FEA) simulation. The mechanical properties of Zr-4 and Cr was validated by simulation of ring compression test (RCT) of fuel cladding.

  17. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    Energy Technology Data Exchange (ETDEWEB)

    A K Wertsching

    2012-09-01

    As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Flux Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to

  18. Experimental assessment of fuel-cladding interactions

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-29

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

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

    Science.gov (United States)

    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.

  20. Results from studies of surface deposits on the claddings of fuel rods used in RBMK-1000 reactors

    Science.gov (United States)

    Smirnova, I. M.; Markov, D. V.

    2010-07-01

    The results of studies on analyzing the element composition of deposits on the cladding surfaces of fuel rods used in a fuel assembly at the Leningrad nuclear power station are presented. The distribution of elements in deposits over the fuel rod height is analyzed, and the zones of their concentration are revealed. It is shown that deposits of copper penetrating into cracks in the surface layer of zirconium oxide introduce an essential contribution in the development of nodular corrosion of fuel rod claddings.

  1. PFR fuel cladding transient test results and analysis

    Science.gov (United States)

    Cannon, N. S.; Hunter, C. W.; Kear, K. L.; Wood, M. H.

    1986-05-01

    Fuel Cladding Transient Tests (FCTT) were performed on M316 cladding specimens obtained from mixed-oxide fuel pins irradiated in the Prototype Fast Reactor (PFR) to burnups of 4 and 9 atom percent. In these tests, specimens of fuel cladding were pressurized and heated until failure occurred. Samples of cladding from PFR fuel pins exhibited generally greater strength and ductility than specimens from Experimental Breeder Reactor-II (EBR-II) mixed-oxide fuel pins tested under similar conditions. Apparently, the PFR cladding properties were not degraded by a fuel adjacency effect (FAE) observed in fuel pin cladding from EBR-II irradiations. A recently developed model of grain boundary cavity growth was used to predict the results of the tests conducted on PFR cladding. It was found that the predicted failure temperatures for the relevant internal pressures were in good agreement with experimental failure temperatures.

  2. Wear resistance and hot corrosion behaviour of laser cladding Co-based alloy

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    2Cr13 stainless steel was surface cladded with Co-based alloy using a high power carbon dioxide laser. The microstructure, wear resistance and corrosion properties of the clad layer were investigated. It is found that the high temperature corrosion behavior and wearing resistant property of the clad layer are 3 and 2.5 times higher than those of the parent metal. Under the high temperature molten lead sulphate salt corrosion condition, the clad layer fails by spalling which is caused by intergrannular corrosion within the clad layer. The fine dendritic structure and the oxide help to retard the penetration of the sulphur ion that induces the intergrannular corrosion.

  3. Oxidation resistant chromium coating on Zircaloy-4 for accident tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jung-Hwan; Kim, Eui-Jung; Jung, Yang-Il; Park, Dong-Jun; Kim, Hyun-Gil; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The attributes of such a fuel are approved reaction kinetics with steam, a slower hydrogen generation rate, and good cladding thermo-mechanical properties. Many researchers have tried to modify zirconium alloys to improve their oxidation resistance in the early stages of the ATF development. Corrosion resistant coating on cladding is one of the candidate technologies to improve the oxidation resistance of zirconium cladding. By applying coating technology to zirconium cladding, it is easy to obtain corrosion resistance without a change in the base materials. Among the surface coating methods, arc ion plating (AIP) is a coating technology to improve the adhesion owing to good throwing power, and a dense deposit (Fig. 1). Owing to these advantages, AIP has been widely used to efficiently form protective coatings on cutting tools, dies, bearings, etc. In this study, The AIP technique for the protection of zirconium claddings from the oxidation in a high-temperature steam environment was studied. The homogeneous Cr film with a high adhesive ability to the cladding was deposited by AIP and acted as a protection layer to enhance the corrosion resistance of the zirconium cladding. It was concluded that the AIP technology is effective for coating a protective layer on claddings.

  4. Mechanical modelling of transient- to- failure SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L. E.

    2014-07-01

    The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)

  5. Water-moderated reactor fuel cladding reliability study

    OpenAIRE

    Бакутяк, Елена Викторовна; Пелых, Сергей Николаевич

    2014-01-01

    Considering the fuel element, averaged by fuel assembly (FA) of water-moderated reactor with the power of 1000 MW (VVER-1000), the number of fuel elements with the greatest cladding failure probability after 4 operation years at Khmelnitsky NPP-2 (KNPP-2) is found. This will allow to calculate the fuel cladding failure probability and determine the most likely cladding damages, which will enable to improve the performance and economic indexes of VVER.The novelty of the paper lies in calculati...

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

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, Andrey [Paul Scherrer Institut, Villigen (Switzerland); Degueldre, Claude, E-mail: claude.degueldre@psi.ch [Paul Scherrer Institut, Villigen (Switzerland); Kaufmann, Wilfried [Kernkraftwerk Leibstadt, Leibstadt (Switzerland)

    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.

  7. The Corrosion and Corrosion Fatigue Behavior of Nickel Based Alloy Weld Overlay and Coextruded Claddings

    Science.gov (United States)

    Stockdale, Andrew

    The use of low NOx boilers in coal fired power plants has resulted in sulfidizing corrosive conditions within the boilers and a reduction in the service lifetime of the waterwall tubes. As a solution to this problem, Ni-based weld overlays are used to provide the necessary corrosion resistance however; they are susceptible to corrosion fatigue. There are several metallurgical factors which give rise to corrosion fatigue that are associated with the localized melting and solidification of the weld overlay process. Coextruded coatings offer the potential for improved corrosion fatigue resistance since coextrusion is a solid state coating process. The corrosion and corrosion fatigue behavior of alloy 622 weld overlays and coextruded claddings was investigated using a Gleeble thermo-mechanical simulator retrofitted with a retort. The experiments were conducted at a constant temperature of 600°C using a simulated combustion gas of N2-10%CO-5%CO2-0.12%H 2S. An alternating stress profile was used with a minimum tensile stress of 0 MPa and a maximum tensile stress of 300 MPa (ten minute fatigue cycles). The results have demonstrated that the Gleeble can be used to successfully simulate the known corrosion fatigue cracking mechanism of Ni-based weld overlays in service. Multilayer corrosion scales developed on each of the claddings that consisted of inner and outer corrosion layers. The scales formed by the outward diffusion of cations and the inward diffusion of sulfur and oxygen anions. The corrosion fatigue behavior was influenced by the surface finish and the crack interactions. The initiation of a large number of corrosion fatigue cracks was not necessarily detrimental to the corrosion fatigue resistance. Finally, the as-received coextruded cladding exhibited the best corrosion fatigue resistance.

  8. Corrosion resistant PEM fuel cell

    Science.gov (United States)

    Fronk, Matthew Howard; Borup, Rodney Lynn; Hulett, Jay S.; Brady, Brian K.; Cunningham, Kevin M.

    2002-01-01

    A PEM fuel cell having electrical contact elements comprising a corrosion-susceptible substrate metal coated with an electrically conductive, corrosion-resistant polymer containing a plurality of electrically conductive, corrosion-resistant filler particles. The substrate may have an oxidizable metal first layer (e.g., stainless steel) underlying the polymer coating.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gibert, C

    1999-07-01

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

  10. ZrC COATING ON FUEL ELEMENT CLADDING ZIRCALOY-2

    Directory of Open Access Journals (Sweden)

    Etty Mutiara

    2017-02-01

    Full Text Available ZrC COATING ON FUEL ELEMENT ZIRCALOY-2 CLADDING. The intensive researchs on high discharge burn-up of Light Water Reactor (LWR fuel element were performed due to the extension of fuel element’s utility life. One of these researches was allowing for alteration of the existing zirconium-based clad system through coating. This technique is supposed to improve the corrosion resistance of cladding without changing the dimension of fuel cladding. In current research, the ZrC film was coated on the zircaloy-2 cladding surface by dipping process of zircaloy-2 specimens in colloidal graphite at room temperature. The dip-coated specimens then undergone heating process at 700oC, 900oC and 1100oC respectively in Argon gas atmosphere for 1 hour. The microstructure and crystal structure of the coated cladding were characterized by optical microscope and XRD respectively. The optical microscope showed the growth of the grains with increasing temperature. XRD examination on the specimens revealed that the ZrC crystal structure on the cladding surface occurred only at 1100oC, but it did not appear at 700oC and 900oC. It can be concluded that dipping process of specimen in colloidal graphite with subsequent heating at 1100oC provided ZrC film coated on zircaloy-2 cladding. The heating process at this temperature allowed carbon atoms to diffuse into zircaloy surface to form ZrC film. PELAPISAN ZrC PADA KELONGSONG ELEMEN BAKAR NUKLIR ZIRKALOI-2. Riset yang intensif pada elemen bakar reaktor berpendingin air dengan fraksi bakar tinggi terus dilakukan dalam rangka memperpanjang umur operasi elemen bakar. Salah satu riset tersebut berupa proses untuk mengubah kelongsong berbasis zirkonium yang ada saat ini dengan cara pelapisan. Cara ini diharapkan akan memperbaiki ketahanan korosi kelongsong tanpa mengubah dimensi kelongsong tersebut. Pada riset ini, lapisan tipis ZrC dilapiskan pada permukaan kelongsong zirkaloi-2 melalui proses pencelupan (dipping spesimen

  11. Pellet-clad interaction in water reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

  12. Siemens advance PWR fuel assemblies (HTP) and cladding

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R. B.; Woods, K. N. [Siemens Nuclear Power Corp., Richland, WA (United States)

    1997-04-01

    This paper describes the key features of the Siemens HTP (High Thermal Performance) fuel design, the current in-reactor performance of this advanced fuel assembly design, and the advanced cladding types available.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  14. Nuclear reactor fuel element with vanadium getter on cladding

    Science.gov (United States)

    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.

  15. Tensile Hoop Behavior of Irradiated Zircaloy-4 Nuclear Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jaramillo, Roger A [ORNL; Hendrich, WILLIAM R [ORNL; Packan, Nicolas H [ORNL

    2007-03-01

    A method for evaluating the room temperature ductility behavior of irradiated Zircaloy-4 nuclear fuel cladding has been developed and applied to evaluate tensile hoop strength of material irradiated to different levels. The test utilizes a polyurethane plug fitted within a tubular cladding specimen. A cylindrical punch is used to compress the plug axially, which generates a radial displacement that acts upon the inner diameter of the specimen. Position sensors track the radial displacement of the specimen outer diameter as the compression proceeds. These measurements coupled with ram force data provide a load-displacement characterization of the cladding response to internal pressurization. The development of this simple, cost-effective, highly reproducible test for evaluating tensile hoop strain as a function of internal pressure for irradiated specimens represents a significant advance in the mechanical characterization of irradiated cladding. In this project, nuclear fuel rod assemblies using Zircaloy-4 cladding and two types of mixed uranium-plutonium oxide (MOX) fuel pellets were irradiated to varying levels of burnup. Fuel pellets were manufactured with and without thermally induced gallium removal (TIGR) processing. Fuel pellets manufactured by both methods were contained in fuel rod assemblies and irradiated to burnup levels of 9, 21, 30, 40, and 50 GWd/MT. These levels of fuel burnup correspond to fast (E > 1 MeV) fluences of 0.27, 0.68, 0.98, 1.4 and 1.7 1021 neutrons/cm2, respectively. Following irradiation, fuel rod assemblies were disassembled; fuel pellets were removed from the cladding; and the inner diameter of cladding was cleaned to remove residue materials. Tensile hoop strength of this cladding material was tested using the newly developed method. Unirradiated Zircaloy-4 cladding was also tested. With the goal of determining the effect of the two fuel types and different neutron fluences on clad ductility, tensile hoop strength tests were

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Dr. John Garnier; Dr. Kevin McHugh

    2012-09-01

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

  18. Double-clad nuclear-fuel safety rod

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Jason D. Hales; Various

    2014-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-15

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

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

    Science.gov (United States)

    Courty, Olivier; Motta, Arthur T.; Hales, Jason D.

    2014-09-01

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

  2. SPENT NUCLEAR FUEL STORAGE BASIN WATER CHEMISTRY: ELECTROCHEMICAL EVALUATION OF ALUMINUM CORROSION

    Energy Technology Data Exchange (ETDEWEB)

    Hathcock, D

    2007-10-30

    The factors affecting the optimal water chemistry of the Savannah River Site spent fuel storage basin must be determines in order to optimize facility efficiency, minimize fuel corrosion, and reduce overall environmental impact from long term spent nuclear fuel storage at the Savannah River Site. The Savannah River National Laboratory is using statistically designed experiments to study the effects of NO{sub 3}{sup -}, SO{sub 4}{sup 2-}, and Cl{sup -} concentrations on alloys commonly used not only as fuel cladding, but also as rack construction materials The results of cyclic polarization pitting and corrosion experiments on samples of Al 6061 and 1100 alloys will be used to construct a predictive model of the basin corrosion and its dependence on the species in the basin. The basin chemistry model and corrosion will be discussed in terms of optimized water chemistry envelope and minimization of cladding corrosion.

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

    Directory of Open Access Journals (Sweden)

    WEON-JU KIM

    2013-08-01

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

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

    Science.gov (United States)

    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.

  5. Application of Coating Technology for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To commercialize the ATF cladding concepts, various factors are considered, such as safety under normal and accident conditions, economy for the fuel cycle, and developing development challenges, and schedule. From the proposed concepts, it is known that the cladding coating, FeCrAl alloy, and Zr-Mo claddings are considered as a near/mid-term application, whereas the SiC material is considered as a long-term application. Among them, the benefit of cladding coating on Zr-based alloys is the fuel cycle economy regarding the manufacturing, neutron cross section, and high tritium permeation characteristics. However, the challenge of cladding coating on Zr-based alloys is the lower oxidation resistance and mechanical strength at high-temperature than other concepts. Another important point is the adhesion property between the Zr-based alloy and coating materials. As an improved coating technology compared to a previous study, a 3D laser coating technology supplied with Cr powders is considered to make a coated cladding because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. We are systematically studying the laser beam power, inert gas flow, cooling of the cladding tube, and powder control as key points to develop 3D laser coating technology. After Cr-coating on the Zr-based cladding, ring compression and ring tensile tests were performed to evaluate the adhesion property between a coated layer and Zr-based alloy tube at room temperature (RT), and a high-temperature oxidation test was conducted to evaluate the oxidation behavior at 1200 .deg. C of the coated tube samples. A 3D laser coating method supplied with Cr powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a Cr-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured.

  6. Microstructure stability of candidate stainless steels for Gen-IV SCWR fuel cladding application

    Science.gov (United States)

    Li, Jian; Zheng, W.; Penttilä, S.; Liu, P.; Woo, O. T.; Guzonas, D.

    2014-11-01

    In the past few years, significant progress has been made in materials selection for Gen-IV SCWR fuel cladding applications. Current studies indicate that austenite stainless steels such as 310H are promising candidates for in-core applications. Alloys in this group are promising for their corrosion resistance, SCC resistance, high temperature mechanical properties and creep resistance at temperatures up to 700 °C. However, one under-studied area of this alloy is the long-term microstructure stability under the proposed reactor operating condition. Unstable microstructure not only results in embrittlement but also has the potential to reduce their resistance to corrosion or stress-corrosion cracking. In this study, stainless steels 310H and 304H were tested for their SCWR corrosion resistance and microstructure stability.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  8. The state of the art report on the development of advanced nuclear fuel cladding tube

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-04-01

    Since the operating conditions of modern PWR trend toward long-term operation, high burn-up, high coolant temperature and high pH, the need to develop a new advanced nuclear fuel cladding as an alternative to Zircaloy-4 increased. To overcome this problem, a number of researches to develop a advanced nuclear fuel cladding tube with superior corrosion resistance and creep resistance have been performed in many advanced nations in the field of nuclear power. Especially, some advanced cladding tubes are already confirmed to have an excellent in-pile properties from the test results in commercial reactor. Also in Korea, KAERI has been researching extensively to develop a high burn-up nuclear fuel cladding Zr alloy since 1990. To design new alloys, it is necessary to study the state of the art on the development of advanced alloys in other countries. In this report, as a part of development of advanced Zr alloy, we studied the state of the art on the development of ZIRLO in U.S.A., E635 in Russia, M5 in France, and MDA and NDA in Japan, which will be applied as basic data to develop an advanced Zr alloy.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

  10. Optimization of N18 Zirconium Alloy for Fuel Cladding of Water Reactors

    Institute of Scientific and Technical Information of China (English)

    B.X. Zhou; M. Y. Yao; Z.K. Li; X.M. Wang; J. Zhoua; C.S. Long; Q. Liu; B.F. Luan

    2012-01-01

    In order to optimize the microstructure and composition of N18 zirconium alloy (Zr-1Sn-0.35Nb-0.35Fe-0.1Cr, in mass fraction, %), which was developed in China in 1990s, the effect of microstructure and composition variation on the corrosion resistance of the N18 alloy has been investigated. The autoclave corrosion tests were carried out in super heated steam at 400 ~C/10.3 MPa, in deionized water or lithiated water with 0.01 mol/L LiOH at 360 ~C/18.6 MPa. The exposure time lasted for 300-550 days according to the test temperature. The results show that the microstructure with a fine and uniform distribution of second phase particles (SPPs), and the decrease of Sn content from 1% (in mass fraction, the same as follows) to 0.8% are of benefit to improving the corrosion resistance; It is detrimental to the corrosion resistance if no Cr addition. The addition of Nb content with upper limit (0.35%) is beneficial to improving the corrosion resistance. The addition of Cu less than 0.1% shows no remarkable influence upon the corrosion resistance for N18 alloy. Comparing the corrosion resistance of the optimized N18 with other commercial zirconium alloys, such as Zircaloy-4, ZIRLO, E635 and Ell0, the former shows superior corrosion resistance in all autoclave testing conditions mentioned above. Although the data of the corrosion resistance as fuel cladding for high burn-up has not been obtained yet, it is believed that the optimized N18 alloy is promising for the candidate of fuel cladding materials as high burn-up fuel assemblies. Based on the theory that the microstructural evolution of oxide layer during corrosion process will affect the corrosion resistance of zirconium alloys, the improvement of corrosion resistance of the N18 alloy by obtaining the microstructure with nano-size and uniform distribution of SPPs, and by decreasing the content of Sn and maintaining the content of Cr is discussed.

  11. Modeling the mechanical behaviour of CANDU fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Holt, R.A. [Queen' s Univ., Dept. of Mechanical Engineering, Kingston, Ontario (Canada)

    2003-07-01

    Models for the mechanical behaviour of fuel cladding were developed in the period 1973-1983 by staff at AECL CRNL. The models for the mechanical properties of fuel cladding during normal operation were a by-product of programs during the period 1970-1975 to understand the origin of fuel-cladding defects caused by power ramps at Douglas Point and Pickering A. Models for accident conditions were, initially, based heavily on mechanical properties data generated by McGill University and Westinghouse Canada under contract to AECL in the late 1960's and early 1970's and attempts to interpret the data in terms of the underlying deformation mechanisms. The model for normal operating conditions was embodied in the ELESTRES/ELESIM series of codes, and the models for accident conditions were embodied in NIRVANA. (author)

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

    Directory of Open Access Journals (Sweden)

    Bo Cheng

    2016-02-01

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

  13. Integrated Computational Modeling of Water Side Corrosion in Zirconium Metal Clad Under Nominal LWR Operating Conditions

    Science.gov (United States)

    Aryanfar, Asghar; Thomas, John; Van der Ven, Anton; Xu, Donghua; Youssef, Mostafa; Yang, Jing; Yildiz, Bilge; Marian, Jaime

    2016-10-01

    A mesoscopic chemical reaction kinetics model to predict the formation of zirconium oxide and hydride accumulation light-water reactor (LWR) fuel clad is presented. The model is designed to include thermodynamic information from ab initio electronic structure methods as well as parametric information in terms of diffusion coefficients, thermal conductivities and reaction constants. In contrast to approaches where the experimentally observed time exponents are captured by the models by design, our approach is designed to be predictive and to provide an improved understanding of the corrosion process. We calculate the time evolution of the oxide/metal interface and evaluate the order of the chemical reactions that are conducive to a t 1/3 dependence. We also show calculations of hydrogen cluster accumulation as a function of temperature and depth using spatially dependent cluster dynamics. Strategies to further cohesively integrate the different elements of the model are provided.

  14. Integrated Computational Modeling of Water Side Corrosion in Zirconium Metal Clad Under Nominal LWR Operating Conditions

    Science.gov (United States)

    Aryanfar, Asghar; Thomas, John; Van der Ven, Anton; Xu, Donghua; Youssef, Mostafa; Yang, Jing; Yildiz, Bilge; Marian, Jaime

    2016-11-01

    A mesoscopic chemical reaction kinetics model to predict the formation of zirconium oxide and hydride accumulation light-water reactor (LWR) fuel clad is presented. The model is designed to include thermodynamic information from ab initio electronic structure methods as well as parametric information in terms of diffusion coefficients, thermal conductivities and reaction constants. In contrast to approaches where the experimentally observed time exponents are captured by the models by design, our approach is designed to be predictive and to provide an improved understanding of the corrosion process. We calculate the time evolution of the oxide/metal interface and evaluate the order of the chemical reactions that are conducive to a t 1/3 dependence. We also show calculations of hydrogen cluster accumulation as a function of temperature and depth using spatially dependent cluster dynamics. Strategies to further cohesively integrate the different elements of the model are provided.

  15. CLAD CARBIDE NUCLEAR FUEL, THERMIONIC POWER, MODULES.

    Science.gov (United States)

    The general objective is to evaluate a clad carbide emitter, thermionic power module which simulates nuclear reactor installation, design, and...performance. The module is an assembly of two series-connected converters with a single common cesium reservoir. The program goal is 500 hours

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-15

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

  17. A method for limitation of probability of accumulation of fuel elements claddings damage in WWER

    OpenAIRE

    Sergey N. Pelykh; Mark V. Nikolsky; S. D. Ryabchikov

    2014-01-01

    The aim is to reduce the probability of accumulation of fuel elements claddings damage by developing a method to control the properties of the fuel elements on stages of design and operation of WWER. An averaged over the fuel assembly WWER-1000 fuel element is considered. The probability of depressurization of fuel elements claddings is found. The ability to predict the reliability of claddings by controlling the factors that determine the properties of the fuel elements is proved. The expedi...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Impact of thicker cladding on the nuclear parameters of the NPP Krsko fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kromar, Marjan, E-mail: marjan.kromar@ijs.s [Jozef Stefan Institute, Reactor Physics Department, Jamova 39, 1001 Ljubljana (Slovenia); Kurincic, Bojan [Nuclear Power Plant Krsko, Engineering Division, Nuclear Fuel and Reactor Core, Vrbina 12, 8270 Krsko (Slovenia)

    2011-04-15

    To make fuel rods more resistant to grid-to-rod fretting or other cladding penetration failures, the cladding thickness could be increased or strengthened. Implementation of thicker fuel rod cladding was evaluated for the NPP Krsko that uses 16 x 16 fuel design. Cladding thickness of the Westinghouse standard fuel design (STD) and optimized fuel design (OFA) is increased. The reactivity effect during the fuel burnup is determined. To obtain a complete realistic view of the fuel behaviour a typical, near equilibrium, 18-month fuel cycle is investigated. The most important nuclear core parameters such as critical boron concentrations, isothermal temperature coefficient and rod worth are determined and compared.

  20. Corrosion Surveillance for Research Reactor Spent Nuclear Fuel in Wet Basin Storage

    Energy Technology Data Exchange (ETDEWEB)

    Howell, J.P.

    1998-10-16

    Foreign and domestic test and research reactor fuel is currently being shipped from locations over the world for storage in water filled basins at the Savannah River Site (SRS). The fuel was provided to many of the foreign countries as a part of the "Atoms for Peace" program in the early 1950's. In support of the wet storage of this fuel at the research reactor sites and at SRS, corrosion surveillance programs have been initiated. The International Atomic Energy Agency (IAEA) established a Coordinated Research Program (CRP) in 1996 on "Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water" and scientists from ten countries worldwide were invited to participate. This paper presents a detailed discussion of the IAEA sponsored CRP and provides the updated results from corrosion surveillance activities at SRS. In May 1998, a number of news articles around the world reported stories that microbiologically influenced corrosion (MIC) was active on the aluminum-clad spent fuel stored in the RBOF basin at SRS. This assessment was found to be in error with details presented in this paper. A biofilm was found on aluminum coupons, but resulted in no corrosion. Cracks seen on the surface were not caused by corrosion, but by stresses from the volume expansion of the oxide formed during pre-conditioning autoclaving. There has been no pitting caused by MIC or any other corrosion mechanism seen in the RBOF basin since initiation of the SRS Corrosion Surveillance Program in 1993.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-12

    In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermal swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO₂ ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U₃Si₅ fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.

  2. Novel Accident-Tolerant Fuel Meat and Cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-01

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

  3. Impact of the use of the ferritic/martensitic ODS steels cladding on the fuel reprocessing PUREX process

    Science.gov (United States)

    Gwinner, B.; Auroy, M.; Mas, D.; Saint-Jevin, A.; Pasquier-Tilliette, S.

    2012-09-01

    Some ferritic/martensitic oxide dispersed strengthened (F/M ODS) steels are presently developed at CEA for the fuel cladding of the next generation of sodium fast nuclear reactors. The objective of this work is to study if this change of cladding could have any consequences on the spent fuel reprocessing PUREX process. During the fuel dissolution stage the cladding can actually be corroded by nitric acid. But some process specifications impose not to exceed a limit concentration of the corrosion products such as iron and chromium in the dissolution medium. For that purpose the corrosion behavior of these F/M ODS steels is studied in hot and concentrated nitric acid. The influence of some metallurgical parameters such as the chromium content, the elaboration process and the presence of the yttrium oxides is first discussed. The influence of environmental parameters such as the nitric acid concentration, the temperature and the presence of oxidizing species coming from the fuel is then analyzed. The corrosion rate is characterized by mass loss measurements and electrochemical tests. Analyses of the corroded surface are carried out by X-ray photoelectron spectroscopy.

  4. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings - phase II

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Stanko, G.J. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1996-08-01

    In Phase I a variety of developmental and commercial tubing alloys and claddings were exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase II (in situ testing) has exposed samples of 347, RA-8511, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 modified, 800HT, NF 709, 690 clad, and 671 clad for over 10,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were installed on an air-cooled, retractable corrosion probe, installed in the reheater cavity, and controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. Samples of each alloy will be exposed for 4000, 12,000, and 16,000 hours of operation. The results will be presented for the metallurgical examination of the corrosion probe samples after 4000 hours of exposure.

  5. Material Selection for Accident Tolerant Fuel Cladding

    Science.gov (United States)

    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.

  6. Material Selection for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥ 1200°C for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO2, and therefore Ti2AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation-assisted α´ formation. The composition effects and critical limits to retaining protective scale formation at > 1400°C are still being evaluated.

  7. Erosion and Corrosion Behavior of Laser Cladded Stainless Steels with Tungsten Carbide

    Science.gov (United States)

    Singh, Raghuvir; Kumar, Mukesh; Kumar, Deepak; Mishra, Suman K.

    2012-11-01

    Laser cladding of tungsten carbide (WC) on stainless steels 13Cr-4Ni and AISI 304 substrates has been performed using high power diode laser. The cladded stainless steels were characterized for microstructural changes, hardness, solid particle erosion resistance and corrosion behavior. Resistance of the clad to solid particle erosion was evaluated using alumina particles according to ASTM G76 and corrosion behavior was studied by employing the anodic polarization and open circuit potential measurement in 3.5% NaCl solution and tap water. The hardness of laser cladded AISI 304 and 13Cr-4Ni stainless steel was increased up to 815 and 725Hv100 g, respectively. The erosion resistance of the modified surface was improved significantly such that the erosion rate of cladded AISI 304 (at 114 W/mm2) was observed ~0.74 mg/cm2/h as compared to ~1.16 and 0.97 mg/cm2/h for untreated AISI 304 and 13Cr-4Ni, respectively. Laser cladding of both the stainless steels, however, reduced the corrosion resistance in both NaCl and tap water.

  8. Fuel corrosion processes under waste disposal conditions

    Energy Technology Data Exchange (ETDEWEB)

    Shoesmith, D.W. [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada)

    1999-09-01

    Under the oxidizing conditions likely to be encountered in the Yucca Mountain Repository, fuel dissolution is a corrosion process involving the coupling of the anodic dissolution of the fuel with the cathodic reduction of oxidants available within the repository. The oxidants potentially available to drive fuel corrosion are environmental oxygen, supplied by the transport through the permeable rock of the mountain and molecular and radical species produced by the radiolysis of available aerated water. The mechanism of these coupled anodic and cathodic reactions is reviewed in detail. While gaps in understanding remain, many kinetic features of these reactions have been studied in considerable detail, and a reasonably justified mechanism for fuel corrosion is available. The corrosion rate is determined primarily by environmental factors rather than the properties of the fuel. Thus, with the exception of increase in rate due to an increase in surface area, pre-oxidation of the fuel has little effect on the corrosion rate.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

    2014-01-09

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

  10. 78 FR 3853 - Retrievability, Cladding Integrity and Safe Handling of Spent Fuel at an Independent Spent Fuel...

    Science.gov (United States)

    2013-01-17

    ... COMMISSION 10 CFR Parts 71 and 72 Retrievability, Cladding Integrity and Safe Handling of Spent Fuel at an... several key areas, such as: retrievability, cladding integrity, and safe handling of spent fuel... potential policy issues and requirements related to retrievability, cladding integrity, and safe handling...

  11. Optical-Based Sensors for Monitoring Corrosion of Reinforcement Rebar via an Etched Cladding Bragg Grating

    Directory of Open Access Journals (Sweden)

    Faisal Rafiq Mahamd Adikan

    2012-11-01

    Full Text Available In this paper, we present the development and testing of an optical-based sensor for monitoring the corrosion of reinforcement rebar. The testing was carried out using an 80% etched-cladding Fibre Bragg grating sensor to monitor the production of corrosion waste in a localized region of the rebar. Progression of corrosion can be sensed by observing the reflected wavelength shift of the FBG sensor. With the presence of corrosion, the etched-FBG reflected spectrum was shifted by 1.0 nm. In addition, with an increase in fringe pattern and continuously, step-like drop in power of the Bragg reflected spectrum was also displayed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-09-01

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

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

    Directory of Open Access Journals (Sweden)

    Cheng Bo

    2016-01-01

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

  17. Model of fracture for the Zry cladding of nuclear fuel rods included in the code DIONISIO 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Av. del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: soba@cnea.gov.ar; Denis, Alicia [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Av. del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: denis@cnea.gov.ar

    2008-12-15

    The DIONISIO code describes most of the main phenomena occurring in a fuel rod during normal operation of a nuclear power reactor. Starting from the irradiation history, the code predicts the temperature distribution, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the internal rod volume, gas mixing, pressure increase, irradiation growth of the cladding, development of an oxide layer on its surface and hydrogen uptake, restructuring and grain growth in the pellet. This work presents the model of Zircaloy fracture included in the code DIONISIO 1.0. The model of pellet-cladding mechanical interaction (PCMI) provides the forces caused by the solid-solid contact which add to the changing internal pressure and to the constant external pressure. Besides, the program evaluates the effects of a corrosive atmosphere (stress corrosion cracking, SCC) internal or external. With these data, the code calculates the J integral around the tip of an initiated crack, and proceeds to analyze, according to the quantity of corrosive substance dissolved and the cladding stress field, if the crack remains unchanged, if it grows due to the I-SCC mechanism, or if propagation is ductile, following the R curve of the material. Results corresponding to different PHWR and PWR reactors are presented and compared with code results. In particular, good agreement is obtained in the simulation of MOX experiments, where the cladding failed due to propagation of cracks originated in SCC.

  18. Erosion and corrosion resistance of laser cladded AISI 420 stainless steel reinforced with VC

    Science.gov (United States)

    Zhang, Zhe; Yu, Ting; Kovacevic, Radovan

    2017-07-01

    Metal Matrix Composites (MMC) fabricated by the laser cladding process have been widely applied as protective coatings in industries to improve the wear, erosion, and corrosion resistance of components and prolong their service life. In this study, the AISI 420/VC metal matrix composites with different weight percentage (0 wt.%-40 wt.%) of Vanadium Carbide (VC) were fabricated on a mild steel A36 by a high power direct diode laser. An induction heater was used to preheat the substrate in order to avoid cracks during the cladding process. The effect of carbide content on the microstructure, elements distribution, phases, and microhardness was investigated in detail. The erosion resistance of the coatings was tested by using the abrasive waterjet (AWJ) cutting machine. The corrosion resistance of the coatings was studied utilizing potentiodynamic polarization. The results showed that the surface roughness and crack susceptibility of the laser cladded layer were increased with the increase in VC fraction. The volume fraction of the precipitated carbides was increased with the increase in the VC content. The phases of the coating without VC consisted of martensite and austenite. New phases such as precipitated VC, V8C7, M7C3, and M23C6 were formed when the primary VC was added. The microhardness of the clads was increased with the increase in VC. The erosion resistance of the cladded layer was improved after the introduction of VC. The erosion resistance was increased with the increase in the VC content. No obvious improvement of erosion resistance was observed when the VC fraction was above 30 wt.%. The corrosion resistance of the clads was decreased with the increase in the VC content, demonstrating the negative effect of VC on the corrosion resistance of AISI 420 stainless steel

  19. Experimental Setup for Reflood Quench of Accident Tolerant Fuel Claddings

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan; Lee, Kwan Geun; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The concept of accident tolerant fuel (ATF) is a solution to suppress the hydrogen generation in loss of coolant accident (LOCA) situation without safety injection, which was the critical incident in the severe accident in the Fukushima. The changes in fuel and cladding materials may cause a significant difference in reactor performance in long term operation. Properties in terms of material science and engineering have been tested and showed promising results. However, numerous tests are still required to ensure the design performance and safety. Thermal hydraulic tests including boiling and quenching are partly confirmed, but not yet complete. We have been establishing the experimental setup to confirm the properties in the terms of thermal hydraulics. Design considerations and preliminary tests are introduced in this paper. An experimental setup to test thermal hydraulic characteristics of new ATF claddings are established and tested. The W heater set inside the cladding is working properly, exceeding 690 W/m linear power with thermocouples and insulating ceramic sheaths inside. The coolant injection control was also working in good conditions. The setup is about to complete and going to simulate quenching behavior of the ATF in the LOCA situation.

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

    Science.gov (United States)

    Farahmand, Parisa

    In oil and gas industry, soil particles, crude oil, natural gas, particle-laden liquids, and seawater can carry various highly aggressive elements, which accelerate the material degradation of component surfaces by combination of slurry erosion, corrosion, and wear mechanisms. This material degradation results into the loss of mechanical properties such as strength, ductility, and impact strength; leading to detachment, delamination, cracking, and ultimately premature failure of components. Since the failure of high valued equipment needs considerable cost and time to be repaired or replaced, minimizing the tribological failure of equipment under aggressive environment has been gaining increased interest. It is widely recognized that effective management of degradation mechanisms will contribute towards the optimization of maintenance, monitoring, and inspection costs. The hardfacing techniques have been widely used to enhance the resistance of surfaces against degradation mechanisms. Applying a surface coating improves wear and corrosion resistance and ensures reliability and long-term performance of coated parts. A protective layer or barrier on the components avoids the direct mechanical and chemical contacts of tool surfaces with process media and will reduce the material loss and ultimately its failure. Laser cladding as an advanced hardfacing technique has been widely used for industrial applications in order to develop a protective coating with desired material properties. During the laser cladding, coating material is fused into the base material by means of a laser beam in order to rebuild a damaged part's surface or to enhance its surface function. In the hardfacing techniques such as atmospheric plasma spraying (APS), high velocity oxygen-fuel (HVOF), and laser cladding, mixing of coating materials with underneath surface has to be minimized in order to utilize the properties of the coating material most effectively. In this regard, laser cladding offers

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lo, Wei-Yang; Yang, Yong, E-mail: yongyang@ufl.edu

    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 V{sub 2}C. Diffusion couple tests at 660 °C for 100 h demonstrate that V{sub 2}C layer with a thickness of less than 5 μm can effectively eliminate the inter-diffusion between the lanthanide cerium and HT-9 steel.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-30

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

  6. Aircraft Integral Fuel Tank Corrosion Study

    Science.gov (United States)

    2007-11-02

    biology of Amorphoteca resinae . Materials und Organismen, 6, (3), p. 161, (1971). 8. D. Cabral. Corrosion by microorganisms of jet aircraft integral fuel...the mycelium of the fungus Hormoconis resinae in the MIC of Al alloys. Proc. XI Int. Corrosion Congress, Houston, USA, 5B, p. 3773, (1993). 14. M

  7. Obtention of fracture properties of unirradiated fuel cladding from ring compression tests

    Energy Technology Data Exchange (ETDEWEB)

    Martin-Rengel, M.A. [Departamento de Ciencia de Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos Profesor Aranguren s/n, E-28040 Madrid (Spain); Consejo de Seguridad Nuclear (CSN), Justo Dorado 11, E-28040 Madrid (Spain); Gomez, F.J.; Ruiz-Hervias, J.; Caballero, L.; Valiente, A. [Departamento de Ciencia de Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos Profesor Aranguren s/n, E-28040 Madrid (Spain)

    2009-06-15

    Zirconium alloy cladding is used as the first structural barrier to contain the nuclear fuel and the fission products. In addition to its neutron transparency, this material has a good corrosion resistance and remarkable mechanical properties at operational temperatures. Consequently, it is or paramount importance to precisely characterize the mechanical behaviour and fracture properties of irradiated cladding to ensure a safe operation. It is known that the mechanical behaviour of unirradiated zirconium alloy cladding is anisotropic. The elastoplastic response depends on the direction, namely radial, hoop or longitudinal. For this reason, different fracture properties should be expected in each direction. From the various tests employed to characterize the mechanical behaviour along the hoop direction in nuclear fuel cladding, the ring compression test is particularly useful to study material fracture. With this test it is possible to determine the moment when a real crack is formed, due to a sudden decrease in the applied load at a given displacement value. The aim of this research is to determine as precisely as possible the value of the fracture energy from the ring compression test load vs. displacement curves. To this end, a finite element calculation incorporating the cohesive zone model was performed. In this case, the cohesive zone theory is applied in its simplest form. It is considered that the cohesive crack transfers a constant stress until the displacement of this cohesive crack reaches a critical value. At this precise moment a real crack is generated. The properties of the softening curve of the cohesive zone model can be obtained by directly comparing the experimental load vs. displacement records with the finite element calculations. The area under the softening curve is the fracture energy, which is directly related with the material fracture toughness. The experimental data used in this work have been obtained on unirradiated Zirlo cladding

  8. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings - Phase II

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Krawchuk, M.T.; Van Weele, S.F. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1995-08-01

    A number of developmental and commercial tubing alloys and claddings have previously been exposed in Phase I to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. This program is exposing samples of TP 347, RA-85H, HR-3C, 253MA, Fe{sub 3}Al + 5Cr, 310 modified, NF-709, 690 clad, and 671 clad, which showed good corrosion resistance from Phase 1, to the actual operating conditions of a 250-MW, coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, and are being controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. The exposure will continue for 4000, 12,000, and 16,000 hours of operation. After the three exposure times, the samples will be metallurgically examined to determine the wastage rates and mode of attack. The probes were commissioned November 16, 1994. The temperatures are being recorded every 15 minutes, and the weighted average temperature calculated for each sample. Each of the alloys is being exposed to a temperature in each of two temperature bands-1150 to 1260{degrees}F and 1260 to 1325{degrees}F. After 2000 hours of exposure, one of the corrosion probes was cleaned and the wall thicknesses were ultrasonically measured. The alloy performance from the field probes will be discussed.

  9. DISSOLUTION OF ZIRCALOY 2 CLAD UO2 COMMERCIAL REACTOR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Sharon M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chattin, Marc Rhea [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giaquinto, Joseph [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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

  11. Corrosion Behavior of Copper-Clad Steel Bars with Unclad Two-End Faces for Grounding Grids in the Red Clay Soil

    Science.gov (United States)

    Shao, Yupei; Mu, Miaomiao; Zhang, Bing; Nie, Kaibin; Liao, Qiangqiang

    2017-02-01

    Iron-aluminum oxides in the red soil have a significant impact on the corrosion behavior of the metal for grounding grids. Effects of iron-aluminum oxides on the corrosion behavior of the cross section of copper-clad steel in the red soil have been investigated using electrochemical impedance spectroscopy and Tafel polarization. All the data indicate that the iron-aluminum oxides can promote the corrosion of copper-clad steel in the red soil. The corrosivity of the red soil greatly increases after iron-aluminum oxides are added into the soil. Iron-aluminum oxides promote galvanic corrosion of copper-clad steel and increase the corrosion degree of the center steel layer. The iron-aluminum oxides stimulate corrosion process of copper-clad steel acting as a cathodic depolarizing agent. XRD results further validate that the corrosion products of the copper-clad steel bar mainly consist of Fe3O4 and Cu2O.

  12. Corrosion Behavior of Copper-Clad Steel Bars with Unclad Two-End Faces for Grounding Grids in the Red Clay Soil

    Science.gov (United States)

    Shao, Yupei; Mu, Miaomiao; Zhang, Bing; Nie, Kaibin; Liao, Qiangqiang

    2017-04-01

    Iron-aluminum oxides in the red soil have a significant impact on the corrosion behavior of the metal for grounding grids. Effects of iron-aluminum oxides on the corrosion behavior of the cross section of copper-clad steel in the red soil have been investigated using electrochemical impedance spectroscopy and Tafel polarization. All the data indicate that the iron-aluminum oxides can promote the corrosion of copper-clad steel in the red soil. The corrosivity of the red soil greatly increases after iron-aluminum oxides are added into the soil. Iron-aluminum oxides promote galvanic corrosion of copper-clad steel and increase the corrosion degree of the center steel layer. The iron-aluminum oxides stimulate corrosion process of copper-clad steel acting as a cathodic depolarizing agent. XRD results further validate that the corrosion products of the copper-clad steel bar mainly consist of Fe3O4 and Cu2O.

  13. Cr plating technology for preventing Fuel Cladding Chemical Interaction

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Ryu, Ho Jin; Jee, Seung Hyun; Cheon, Jin Sik; Lee, Byoung Oon; Lee, Chan Bock; Yang, Seong Woo [KAERI, Daejeon (Korea, Republic of)

    2010-11-15

    The objectives of the report are to analyze chrome electroplating technology in order to apply in the field of diffusion barrier to suppress Fuel-Cladding Chemical Interaction (FCCI). This report consists of the principle of the chrome electroplating, plating parameter and possibility of the barrier application. Chrome plating has been considered as one of the probable candidates in the field of barrier tube because of its simpleness, superior FCCI resistance, and effective coating performance at relatively low cost. However, cracks can be generate at the surface of the coating surface which reduces the coating performance. To minimize such a crack, controlling plating parameter like bath composition and bath temperature, current profile, and post-heat treatment has been reviewed. Concept for the application at the inner surface of the cladding has been also described. Based on the technology that suggested at the present report, optimizing plating parameter will be carried out. After the performance test like diffusion couple test of the metallic fuel, final barrier condition will be concluded and the fabrication of the prototype barrier tube will be conducted in the near future

  14. Corrosion free phosphoric acid fuel cell

    Science.gov (United States)

    Wright, Maynard K.

    1990-01-01

    A phosphoric acid fuel cell with an electrolyte fuel system which supplies electrolyte via a wick disposed adjacent a cathode to an absorbent matrix which transports the electrolyte to portions of the cathode and an anode which overlaps the cathode on all sides to prevent corrosion within the cell.

  15. In-core measurements of fuel-clad interactions in the Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bennett Peter

    2008-10-15

    A combination of on-line measurement techniques was used to demonstrate AOA in the Halden reactor: - Diameter gauge to demonstrate crud deposition - Coolant flow and temperature measurements to show effect of crud on thermal-hydraulic conditions - Neutron detectors to show power depression caused by boron in the crud. - Coolant chemistry analyses provided supporting evidence of AOA - Lithium return during shutdown - PIE showed that the type of crud observed in US plants suffering severe AOA can be reproduced. Loop systems allow testing under LWR thermal-hydraulic and water chemistry conditions. - A combination of on-line instrumentation allows measurements of complicated phenomena, egPWR AOA. - Techniques are under development to allow on-line measurements of fuel clad corrosion

  16. Enhancing surface integrity and corrosion resistance of laser cladded Cr-Ni alloys by hard turning and low plasticity burnishing

    Science.gov (United States)

    Zhang, Peirong; Liu, Zhanqiang

    2017-07-01

    In this research, the enhancements of surface integrity and corrosion resistance of the laser cladded parts by combined hard turning with low plasticity burnishing (LPB) were presented by both potentiodynamic polarization and electrochemical impedance spectroscopy (EIS) methods. The investigated results indicated that the corrosion resistance of the laser cladded parts could be improved by combined hard turning with LPB than by sole hard turning. An innovative model was proposed to explain the corrosion mechanism of the laser cladded parts after hybrid machining. Both surface adsorption and passive film were observed to dominate the corrosion resistance of the hybrid machined Cr-Ni alloys by laser cladding. The surface integrity led to the inhomogeneity of passive film, and then altered the corrosion resistance of the machined samples. In terms of the surface integrity factors, residual compressive stresses and surface finish were found to play more important roles in improving the corrosion resistance than the grain refinement and microhardness of the machined surface layer materials did. Based on the research results, anti-corrosion parts with laser cladded alloys could be fabricated by hybrid machining using the combination of hard turning and LPB.

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

    Science.gov (United States)

    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

  18. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Seitz, W.W. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1997-12-01

    In Phase 1 a variety of developmental and commercial tubing alloys and claddings were exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347 RA-85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 Ta modified, NF 709, 690 clad, and 671 clad for approximately 4,000, 12,000, and 16,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were assembled on an air-cooled, retractable corrosion probe, the probe was installed in the reheater activity of the boiler and controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. The results will be presented for the preliminary metallurgical examination of the corrosion probe samples after 16,000 hours of exposure. Continued metallurgical and interpretive analysis is still on going.

  19. Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2 field testing

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Seitz, W.W.; Girshik, A. [Foster Wheeler Development Corp., Livingston, NJ (United States)

    1998-06-01

    In Phase 1 of this project, laboratory experiments were performed on a variety of developmental and commercial tubing alloys and claddings by exposing them to fireside corrosion tests which simulated a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347, RA85H, HR3C, RA253MA, Fe{sub 3}Al + 5Cr, Ta-modified 310, NF 709, 690 clad, 671 clad, and 800HT for up to approximately 16,000 hours to the actual operating conditions of a 250-MW, coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, and controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy were exposed for 4,483, 11,348, and 15,883 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after the full 15,883 hours of exposure. A previous topical report has been issued for the 4,483 hours of exposure.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-09-01

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

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

    Science.gov (United States)

    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

  2. DUPIC fuel irradiation test and performance evaluation; the performance analysis of pellet-cladding contact fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ho, K. I.; Kim, H. M.; Yang, K. B.; Choi, S. J. [Suwon University, Whasung (Korea)

    2002-04-01

    Thermal and mechanical models were reviewed, and selected for the analysis of nuclear fuel performance in reactor. 2 dimensional FEM software was developed. Thermal models-gap conductances, thermal conductivity of pellets, fission gas release, temperature distribution-were set and packaged into a software. Both thermal and mechanical models were interrelated to each other, and the final results, fuel performance during irradiation is obtained by iteration calculation. Also, the contact phenomena between pellet and cladding was analysed by mechanical computer software which was developed during this work. dimensional FEM program was developed which estimate the mechanical behavior and the thermal behaviors of nuclear fuel during irradiation. Since there is a importance during the mechanical deformation analysis in describing pellet-cladding contact phenomena, simplified 2 dimensional calculation method is used after the contact. The estimation of thermal fuel behavior during irradiation was compared with the results of other. 8 refs., 17 figs. (Author)

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

    Science.gov (United States)

    Metzger, Kathryn E.

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

  4. Post-irradiation examination of AlFeNi cladded U 3Si 2 fuel plates irradiated under severe conditions

    Science.gov (United States)

    Leenaers, A.; Koonen, E.; Parthoens, Y.; Lemoine, P.; Van den Berghe, S.

    2008-04-01

    Three full size AlFeNi cladded U 3Si 2 fuel plates were irradiated in the BR2 reactor of the Belgian Nuclear Research Centre (SCK·CEN) under relatively severe, but well defined conditions. The irradiation was part of the qualification campaign for the fuel to be used in the future Jules Horowitz reactor in Cadarache, France. After the irradiation, the fuel plates were submitted to an extensive post-irradiation campaign in the hot cell laboratory of SCK·CEN. The PIE shows that the fuel plates withstood the irradiation successfully, as no detrimental defects have been found. At the cladding surface, a multilayered corrosion oxide film has formed. The U-Al-Si layer resulting from the interaction between the U 3Si 2 fuel and the Al matrix, has been quantified as U(Al,Si) 4.6. It is found that the composition of the fuel particles is not homogenous; zones of USi and U 3Si 2 are observed and measured. The fission gas-related bubbles generated in both phases show a different morphology. In the USi fuel, the bubbles are small and numerous while in U 3Si 2 the bubbles are larger but there are fewer.

  5. Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

    Science.gov (United States)

    Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

    2016-10-01

    Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

  6. Neutronic evaluation of coating and cladding materials for accident tolerant fuels

    OpenAIRE

    Younker, I; Fratoni, M

    2016-01-01

    © 2015 Elsevier Ltd. All rights reserved. In severe accident conditions with loss of active cooling in the core, zirconium alloys, used as fuel cladding materials for current light water reactors (LWR), undergo a rapid oxidation by high temperature steam with consequent hydrogen generation. Novel fuel technologies, named accident tolerant fuels (ATF), seek to improve the endurance of severe accident conditions in LWRs by eliminating or at least mitigating such detrimental steam-cladding inter...

  7. Corrosion product deposits on boiling-water reactor cladding: Experimental and theoretical investigation of magnetic properties

    Science.gov (United States)

    Orlov, A.; Degueldre, C.; Wiese, H.; Ledergerber, G.; Valizadeh, S.

    2011-09-01

    Recent Eddy current investigations on the cladding of nuclear fuel pins have shown that the apparent oxide layers are falsified due to unexpected magnetic properties of corrosion product deposits. Analyses by Scanning Electron Microscopy (SEM) or Electron Probe Micro Analysis (EPMA) demonstrated that the deposit layer consists of complex 3-d element oxides (Ni, Mn, Fe) along with Zn, since the reactor operates with a Zn addition procedure to reduce buildup of radiation fields on the recirculation system surfaces. The oxides crystallise in ferritic spinel structures. These spinels are well-known for their magnetic behaviour. Since non-magnetic zinc ferrite (ZnFe 2O 4) may become magnetic when doped with even small amounts of Ni and/or Mn, their occurrence in the deposit layer has been analyzed. The magnetic permeability of zinc ferrite, trevorite and jacobsite and their solid solutions are estimated by magnetic moment additivity. From the void history examination, the low elevation sample (810 mm) did not face significant boiling during the irradiation cycles suggesting growth of (Mn0.092+Zn0.752+Fe0.293+)[(Fe1.713+Mn0.032+Ni0.132+)O] crystals with theoretical value of the magnetic permeability for the averaged heterogeneous CRUD layer of 9.5 ± 3. Meanwhile, (Mn0.162+Zn0.552+Fe0.293+)[(Fe1.713+Mn0.042+Ni0.252+)O] crystallizes at the mid elevation (1810 mm) with theoretical magnetic permeability for the CRUD layer of 4.2 ± 1.5 at the investigated azimuthal location. These theoretical data are compared with the magnetic permeability of the corrosion product deposited layers gained from reactor pool side Eddy current (EC) analyses (9.0 ± 1.0 for low and 3.5 ± 1.0 for high elevation). The calculated thicknesses and magnetic permeability values of the deposition layers (estimated by MAGNACROX multifrequency EC method) match together with these estimated using an "ion magnetic moment additivity" model.

  8. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Rochkhudson B. de; Cardoso, Fabiano; Salome, Jean A.D.; Pereira, Claubia; Fortini, Angela, E-mail: rochkhudson@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of raising the burning and maintaining the safety of nuclear plants is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and low neutron affinity. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with boron, burnable poison rods, and temperature variations. Sensitivity calculation and the impact in multiplication factor to both claddings, zircaloy and silicon carbide, were performed during the burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. (author)

  9. Development of experimental apparatus for evaluating corrosion resistance of cladding materials applied for advanced power reactor. 1

    Energy Technology Data Exchange (ETDEWEB)

    Inohara, Yasuto; Ioka, Ikuo; Fukaya, Kiyoshi; Tachibana, Katsumi; Suzuki, Tomio; Kiuchi, Kiyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kuroda, Yuji; Miyamoto, Satoshi [Japan Atomic Power Co., Tokyo (Japan)

    2001-03-01

    On the development of cladding materials for advanced power reactors, it is important to clarify long performance and to control the compatibility to high temperature water at heat conducting surfaces under heavy irradiation. On the present study, the high temperature water loop with an autoclave was made for examining the corrosion behavior up to the super critical water range and for developing the simulation testing technique under irradiation in the hot cell. The loop is applicable to immersion tests in the temperature and pressure ranges up to 450degC and 25 MPa that are covered the surface temperature range of fuel claddings. One of the characteristics of this apparatus is a pair of sapphire windows of autoclave for in-situ observations, and a phase transition from water to super critical water conditions was clearly verified through these windows. In this apparatus, it is possible to control the temperature, pressure and Dissolved Oxygen (DO) within a fluctuations of few % on three phases, namely, water, steam and super critical water. (author)

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

    Science.gov (United States)

    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.

  11. Potential for fuel melting and cladding thermal failure during a PCM event in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, M.S.; Croucher, D.W.

    1979-01-01

    The primary concern in nuclear reactor safety is to ensure that no conceivable accident, whether initiated by a failure of the reactor system or by incorrect operation, will lead to a dangerous release of radiation to the environment. A number of hypothesized off-normal power or cooling conditions, generally termed as power-cooling-mismatch (PCM) accidents, are considered in the safety analysis of light water reactors (LWRs). During a PCM accident, film boiling may occur at the cladding surface and cause a rapid temperature increase in the fuel and the cladding, perhaps producing embrittlement of the zircaloy cladding by oxidation. Molten fuel may be produced at the center of the pellets, extrude radially through open cracks in the outer, unmelted portion of the pellet and relocate in the fuel-cladding gap. If the amount of extruded molten fuel is sufficient to establish contact with the cladding, which is at a high temperature during film boiling, the zircaloy cladding may melt. The present work assesses the potential for central fuel melting and thermal failure of the zircaloy cladding due to melting upon being contacted by extruded molten UO/sub 2/-fuel during a PCM event.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

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

  14. Raman Spectroscopy Analysis of Oxide Film on Spent Fuel Rod Cladding from Qinshan PhaseⅠNPP

    Institute of Scientific and Technical Information of China (English)

    WANG; Hua-cai; TANG; Qi; FU; Cheng; LIANG; Zheng-qiang

    2015-01-01

    The outside surface of cladding is one of the important factors limiting the service life of the fuel rods.Studying the structure of oxide film under reactor operating conditions has great significance in study of the cause of different appearances of cladding,establishing the relationship between oxide film thickness and oxide structure

  15. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    Science.gov (United States)

    Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

    2008-12-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

  16. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vaibhaw, Kumar [Nuclear Fuel Complex, ECIL Post, Hyderabad 500 062 (India)], E-mail: krvaibhaw@yahoo.co.in; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N. [Nuclear Fuel Complex, ECIL Post, Hyderabad 500 062 (India)

    2008-12-15

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition ({approx}300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F{sub n}) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

  17. U-Mo Foil/Cladding Interactions in Friction Stir Welded Monolithic RERTR Fuel Plates

    Energy Technology Data Exchange (ETDEWEB)

    D.D. Keiser; J.F. Jue; C.R. Clark

    2006-10-01

    Interaction between U-Mo fuel and Al has proven to dramatically impact the overall irradiation performance of RERTR dispersion fuels. It is of interest to better understand how similar interactions may affect the performance of monolithic fuel plates, where a uranium alloy fuel is sandwiched between aluminum alloy cladding. The monolithic fuel plate removes the fuel matrix entirely, which reduces the total surface area of the fuel that is available to react with the aluminum and moves the interface between the fuel and cladding to a colder region of the fuel plate. One of the major fabrication techniques for producing monolithic fuel plates is friction stir welding. This paper will discuss the interactions that can occur between the U-Mo foil and 6061 Al cladding when applying this fabrication technique. It has been determined that the time at high temperatures should be limited as much as is possible during fabrication or any post-fabrication treatment to reduce as much as possible the interactions between the foil and cladding. Without careful control of the fabrication process, significant interaction between the U-Mo foil and Al alloy cladding can result. The reaction layers produced from such interactions can exhibit notably different morphologies vis-à-vis those typically observed for dispersion fuels.

  18. Screening of advanced cladding materials and UN–U{sub 3}Si{sub 5} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R., E-mail: nbrown@bnl.gov; Todosow, Michael; Cuadra, Arantxa

    2015-07-15

    Highlights: • Screening methodology for advanced fuel and cladding. • Cladding candidates, except for silicon carbide, exhibit reactivity penalty versus zirconium alloy. • UN–U{sub 3}Si{sub 5} fuels have the potential to exhibit reactor physics and fuel management performance similar to UO{sub 2}. • Harder spectrum in the UN ceramic composite fuel increases transuranic build-up. • Fuel and cladding properties assumed in these assessments are preliminary. - Abstract: 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 (UO{sub 2}) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO{sub 2} 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 UO{sub 2}–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–U{sub 3}Si{sub 5} fuels with Kanthal AF or APMT cladding. The objective of the U{sub 3}Si{sub 5} phase in the UN–U{sub 3}Si{sub 5} fuel concept is to shield the nitride phase from water. It was shown that UN–U{sub 3}Si{sub 5} 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 UO{sub 2}–Zr fuel–cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to {sup 14}N content in UN ceramic composites is high

  19. Characterization of Cassini GPHS Fueled-Clad Production Girth Welds

    Energy Technology Data Exchange (ETDEWEB)

    Franco-Ferreira, E.A.

    2000-03-23

    Fueled clads for radioisotope power systems are produced by encapsulating {sup 238}PuO{sub 2} in iridium alloy cups, which are joined at their equators by gas tungsten arc welding. Cracking problems at the girth weld tie-in area during production of the Galileo/Ulysses GPHS capsules led to the development of a first-generation ultrasonic test for girth weld inspection at the Savannah River Plant. A second-generation test and equipment with significantly improved sensitivity and accuracy were jointly developed by the Oak Ridge Y-12 Plant and Westinghouse Savannah River Company for use during the production of Cassini GPHS capsules by the Los Alamos National Laboratory. The test consisted of Lamb wave ultrasonic scanning of the entire girth weld from each end of the capsule combined with a time-of-flight evaluation to aid in characterizing nonrelevant indications. Tangential radiography was also used as a supplementary test for further evaluation of reflector geometry. Each of the 317 fueled GPHS capsules, which were girth welded for the Cassini Program, was subjected to a series of nondestructive tests that included visual, dimensional, helium leak rate, and ultrasonic testing. Thirty-three capsules were rejected prior to ultrasonic testing. Of the 44 capsules rejected by the standard ultrasonic test, 22 were upgraded to flight quality through supplementary testing for an overall process acceptance rate of 82.6%. No confirmed instances of weld cracking were found.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-15

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

  1. Characterization of SiC-SiC composites for accident tolerant fuel cladding

    Science.gov (United States)

    Deck, C. P.; Jacobsen, G. M.; Sheeder, J.; Gutierrez, O.; Zhang, J.; Stone, J.; Khalifa, H. E.; Back, C. A.

    2015-11-01

    Silicon carbide (SiC) is being investigated for accident tolerant fuel cladding applications due to its high temperature strength, exceptional stability under irradiation, and reduced oxidation compared to Zircaloy under accident conditions. An engineered cladding design combining monolithic SiC and SiC-SiC composite layers could offer a tough, hermetic structure to provide improved performance and safety, with a failure rate comparable to current Zircaloy cladding. Modeling and design efforts require a thorough understanding of the properties and structure of SiC-based cladding. Furthermore, both fabrication and characterization of long, thin-walled SiC-SiC tubes to meet application requirements are challenging. In this work, mechanical and thermal properties of unirradiated, as-fabricated SiC-based cladding structures were measured, and permeability and dimensional control were assessed. In order to account for the tubular geometry of the cladding designs, development and modification of several characterization methods were required.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-01

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

  3. Temperature limits for LMFBR fuel cladding under upset and emergency operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Govindarajan, S.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam Tamilnadu (India). Nuclear Systems Division

    1996-07-01

    LMFBR fuel pin cladding tube is subjected to high transient temperatures during incidents such as pump trip, pump to grid plate pipe rupture etc. It is required to know temperature limits under such transient operating conditions for components involved while analyzing such incidents. A methodology for deriving such limits for fuel clad tube is worked out in this paper by making use of the transient damage correlation proposed by W.F. Brizes et. al.

  4. Cold spray deposition of Ti2AlC coatings for improved nuclear fuel cladding

    Science.gov (United States)

    Maier, Benjamin R.; Garcia-Diaz, Brenda L.; Hauch, Benjamin; Olson, Luke C.; Sindelar, Robert L.; Sridharan, Kumar

    2015-11-01

    Coatings of Ti2AlC MAX phase compound have been successfully deposited on Zircaloy-4 (Zry-4) test flats, with the goal of enhancing the accident tolerance of LWR fuel cladding. Low temperature powder spray process, also known as cold spray, has been used to deposit coatings ∼90 μm in thickness using powder particles of accident tolerance to nuclear fuel cladding.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Feinroth, Herbert

    2008-03-03

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

  8. Corrosion of graphite composites in phosphoric acid fuel cells

    Science.gov (United States)

    Christner, L. G.; Dhar, H. P.; Farooque, M.; Kush, A. K.

    1986-01-01

    Polymers, polymer-graphite composites and different carbon materials are being considered for many of the fuel cell stack components. Exposure to concentrated phosphoric acid in the fuel cell environment and to high anodic potential results in corrosion. Relative corrosion rates of these materials, failure modes, plausible mechanisms of corrosion and methods for improvement of these materials are investigated.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar; Allen, Todd; Cole, James

    2013-02-27

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

  10. Direct observation of fuel-cladding mechanical interaction (FCMI) in mixed-oxide fast reactor fuel pins

    Science.gov (United States)

    Foster, J. P.; Nayak, U. P.

    1981-10-01

    The WSA-1 and WSA-2 fuel pins exhibit experimental evidence of fuel-cladding mechanical interaction (FCMI) as a result of steady-state irradiation. The direct FCMI evidence involves a comparison of local axial and hoop mechanical strain profiles. The determination of the local axial mechanical strain was possible because of the placement of axial hardness marks 12.7 mm apart along a line parallel to the tubing axis spanning the fuel column. The measured cladding local axial and hoop mechanical deformations were the same within experimental error. The experimental results are in contrast to gas pressurized tube data which exhibit no axial mechanical deformation. A substantial amount of indirect evidence further illustrating the influence of FCMI on the cladding mechanical strain profile is also discussed. The conditions leading to steady-state FCMI are: high fuel smear density (i.e. low fuel-cladding gaps and/or high fuel pellet density), thin wall cladding, low cladding swelling and low fission gas pressure.

  11. Microstructure and Corrosion Resistance of Cr7C3/γ-Fe Ceramal Composite Coating Fabricated by Plasma Cladding

    Institute of Scientific and Technical Information of China (English)

    LIU Junbo

    2007-01-01

    A new type in situ Cr7C3/γ-Fe ceramal composite coating was fabricated on substrate of hardened and tempered grade C steel by plasma cladding with Fe-Cr-C alloy powders. The ceramal composite coating has a rapidly solidified microstructure consisting of primary Cr7C3 and the Cr7C3/γ-Fe eutectics, and is metallurgically bonded to the degree C steel substrate. The corrosion resistances of the coating in water solutions of 0.5 mol/L H2SO4 and 3.5% NaCl were evaluated utilizing the electrochemical polarization corrosion-test method. Because of the inherent excellent corrosion-resisting properties of the constituting phase and the rapidly solidified homogeneous microstructure, the plasma clad ceramal composite coating exhibits excellent corrosion resistance in the water solutions of 0.5 mol/L H2SO4 and 3.5% NaCl.

  12. Fuel Performance Calculations for FeCrAl Cladding in BWRs

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  13. Simulation of accident and normal fuel rod work with Zr-cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tutnov, Anton A.; Tutnov, Alexander A. [Russian Research Centre, Moscow (Russian Federation). Kurchatov Inst.

    1995-12-31

    The technique of simulation of heat-physics, strength and safety characteristics of reactor RBMK and WWER rods under steady-state, transient and accident conditions is presented. That technique is used in mechanic and heat physics codes PULSAR-2 and STALACTITE. Simulation in both full scale and the most stress-loading part of cladding statement under accident conditions are considered. In this zone local swelling and cladding failure are possible. The accident simulation is based on the mechanical creep-plasticity problem solution in three-dimensional approach. The local cladding swelling is initiated with determining of little hot spot on the clad with several degrees temperature departure from average value. Mechanical problem is solved by finite elements method. Interaction of Zr with steam is taken in to account. Fuel and cladding melting, shortness and dispersion formation processes are simulated under subsequent rods warming up. (author). 2 refs., 6 figs.

  14. Feasibility Study on the Sodium Compatibility Test for Fuel Cladding of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Shin, Sang Hun; Park, Sang Gyu; Ryu, Woo Seog; Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A Sodium-cooled Fast Reactor (SFR), a reactor that uses fast neutrons as a fission process, is considered one of the most probable candidates in next-generation reactors because it can maximize the uranium utilization when compared to conventional water reactor. Liquid sodium is used as a coolant in a SFR, because it has superior efficiency of fast neutron economy and high thermal conductivity, which enables a high power core design. However, previous research reported that fuel cladding materials like austenitic and ferritic-martensitic steel (FMS) react sodium coolant so that it results in the loss of the thickness, intergranular attack, and carburization or decarburization process to induce the change of the mechanical property. Fuel cladding, a seamless tube which has approximately 0.5mm in thickness and 3m in length is the component which covers fuel to protect radioactive species from being released. Because of its smaller thickness, the mechanical properties of the cladding are easily affected by the small changes of material property. This paper summarizes the status of sodium-material compatibility facility and proposes the optimal option in the case of the SFR fuel cladding. Previous researches revealed that assessing in-situ mechanical property is important in the case of cladding material owing to its dimensional characteristic. Optimal test method for assessing sodium compatibility of the cladding tube can be proposed that pressurized creep test under the controlled liquid sodium environment.

  15. A micromechanical model for predicting hydride embrittlement in nuclear fuel cladding material

    Science.gov (United States)

    Chan, K. S.

    1996-01-01

    A major concern about nuclear fuel cladding under waste repository conditions is that the slow cooling rate anticipated in the repository may lead to the formation of excessive radial hydrides, and cause embrittlement of the cladding materials. In this paper, the development of a micromechanical model for predicting hydride-induced embrittlement in nuclear fuel cladding is presented. The important features of the proposed model are: (1) the capability to predict the orientation, morphology, and types of hydrides under the influence of key variables such as cooling rate, internal pressure, and time, and (2) the ability to predict the influence of hydride orientation and morphology on the tensile ductility and fracture toughness of the cladding material. Various model calculations are presented to illustrate the characteristics and utilities of the proposed methodology. A series of experiments was also performed to check assumptions used and to verify some of the model predictions.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-18

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

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

    OpenAIRE

    Bo Cheng; Young-Jin Kim; Peter Chou

    2016-01-01

    In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident managem...

  18. Corrosion assessment of dry fuel storage containers

    Energy Technology Data Exchange (ETDEWEB)

    Graves, C.E.

    1994-09-01

    The structural stability as a function of expected corrosion degradation of 75 dry fuel storage containers located in the 200 Area Low-Level Waste Burial Grounds was evaluated. These containers include 22 concrete burial containers, 13 55-gal (208-l) drums, and 40 Experimental Breeder Reactor II (EBR-II) transport/storage casks. All containers are buried beneath at least 48 in. of soil and a heavy plastic tarp with the exception of 35 of the EBR-II casks which are exposed to atmosphere. A literature review revealed that little general corrosion is expected and pitting corrosion of the carbon steel used as the exterior shell for all containers (with the exception of the concrete containers) will occur at a maximum rate of 3.5 mil/yr. Penetration from pitting of the exterior shell of the 208-l drums and EBR-II casks is calculated to occur after 18 and 71 years of burial, respectively. The internal construction beneath the shell would be expected to preclude containment breach, however, for the drums and casks. The estimates for structural failure of the external shells, large-scale shell deterioration due to corrosion, are considerably longer, 39 and 150 years respectively for the drums and casks. The concrete burial containers are expected to withstand a service life of 50 years.

  19. Impact of reactor water chemistry on cladding performance

    Energy Technology Data Exchange (ETDEWEB)

    Cox, B. [University of Toronto, Centre for Nuclear Engineering, Toronto, Ontario (Canada)

    1997-07-01

    Water chemistry may have a major impact on fuel cladding performance in PWRs. If the saturation temperature on the surface of fuel cladding is exceeded, either because of the thermal hydraulics of the system, or because of crud deposition, then LiOH concentration can occur within thick porous oxide films on the cladding. This can degrade the protective film and accelerate the corrosion rate of the cladding. If sufficient boric acid is also present in the coolant then these effects may be mitigated. This is normally the case through most of any reactor fuel cycle. Extensive surface boiling may disrupt this equilibrium because of the volatility of boric acid in steam. Under such conditions severe cladding corrosion can ensue. The potential for such effects on high burnup cladding in CANDU reactors, where bone acid is not present in the primary coolant, is discussed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [ORNL

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [ORNL

    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.

  2. New Low-Sn Zr Cladding Alloys with Excellent Autoclave Corrosion Resistance and High Strength

    Directory of Open Access Journals (Sweden)

    Ruiqian Zhang

    2017-04-01

    Full Text Available It is expected that low-Sn Zr alloys are a good candidate to improve the corrosion resistance of Zr cladding alloys in nuclear reactors, presenting excellent corrosion resistance and high strength. The present work developed a new alloy series of Zr-0.25Sn-0.36Fe-0.11Cr-xNb (x = 0.4~1.2 wt % to investigate the effect of Nb on autoclave corrosion resistance. Alloy ingots were prepared by non-consumable arc-melting, solid-solutioned, and then rolled into thin plates with a thickness of 0.7 mm. It was found that the designed low-Sn Zr alloys exhibit excellent corrosion resistances in three out of pile autoclave environments (distilled water at 633 K/18.6 MPa, 70 ppm LiOH solution at 633 K/18.6 MPa, and superheated water steam at 673 K/10.3 MPa, as demonstrated by the fact of the Zr-0.25Sn-0.36Fe-0.11Cr-0.6Nb alloy shows a corrosion weight gain ΔG = 46.3 mg/dm2 and a tensile strength of σUTS = 461 MPa following 100 days of exposure in water steam. The strength of the low-Sn Zr alloy with a higher Nb content (x = 1.2 wt % is enhanced up to 499 MPa, comparable to that of the reference high-Sn N36 alloy (Zr-1.0Sn-1.0Nb-0.25Fe, wt %. Although the strength improvement is at a slight expense of corrosion resistance with the increase of Nb, the corrosion resistance of the high-Nb alloy with x = 1.2 (ΔG = 90.4 mg/dm2 for 100-day exposure in the water steam is still better than that of N36 (ΔG = 103.4 mg/dm2.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Qiushi, E-mail: qiushi@clemson.edu [Glenn Department of Civil Engineering, Clemson University, Clemson, SC 29634 (United States); Ostien, Jakob T., E-mail: jtostie@sandia.gov [Mechanics of Materials Dept. 8256, Sandia National Laboratories, P.O. Box 969, Livermore, CA 94551-0969 (United States); Hansen, Glen, E-mail: gahanse@sandia.gov [Computational Multiphysics Dept. 1443, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185-1321 (United States)

    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 J{sub 2} 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.

  5. Stability increase of fuel clad with zirconium oxynitride thin film by metalorganic chemical vapor deposition

    Energy Technology Data Exchange (ETDEWEB)

    Jee, Seung Hyun [Department of Materials Science and Engineering, Yonsei University, 134 Sinchon Dong, Seoul 120-749 (Korea, Republic of); Materials Research and Education Center, Dept. of Mechanical Engineering, Auburn University, 275 Wilmore Labs, AL 36849-5341 (United States); Kim, Jun Hwan; Baek, Jong Hyuk [Recycled Fuel Development Division, Korea Atomic Energy Research Institute, P.O. Box 105, Yuseong, Daejeon, 305-600 (Korea, Republic of); Kim, Dong-Joo [Materials Research and Education Center, Dept. of Mechanical Engineering, Auburn University, 275 Wilmore Labs, AL 36849-5341 (United States); Kang, Seong Sik [Regulatory Research Division, Korea Institute of Nuclear Safety, 19, Guseong-Dong, Yuseong-Gu, Daejeon, 305-338 (Korea, Republic of); Yoon, Young Soo, E-mail: yoonys@yonsei.ac.kr [Department of Materials Science and Engineering, Yonsei University, 134 Sinchon Dong, Seoul 120-749 (Korea, Republic of)

    2012-06-01

    A zirconium oxynitride (ZON) thin film was deposited onto HT9 steel as a cladding material by a metalorganic chemical vapor deposition (MOCVD) in order to prevent a fuel-clad chemical interaction (FCCI) between a U-10 wt% Zr metal fuel and a clad material. X-ray diffraction spectrums indicated that the mixture of structures of zirconium nitride, oxide and carbide in the MOCVD grown ZON thin films. Also, typical equiaxial grain structures were found in plane and cross sectional images of the as-deposited ZON thin films with a thickness range of 250-500 nm. A depth profile using auger electron microscopy revealed that carbon and oxygen atoms were decreased in the ZON thin film deposited with hydrogen gas flow. Diffusion couple tests at 800 Degree-Sign C for 25 hours showed that the as-deposited ZON thin films had low carbon and oxygen content, confirmed by the Energy Dispersive X-ray Spectroscopy, which showed a barrier behavior for FCCI between the metal fuel and the clad. This result suggested that ZON thin film cladding by MOCVD, even with the thickness below the micro-meter level, has a high possibility as an effective FCCI barrier. - Highlights: Black-Right-Pointing-Pointer Zirconium oxynitride (ZON) deposited by metal organic chemical vapor deposition. Black-Right-Pointing-Pointer Prevention of fuel cladding chemical interaction (FCCI) investigated. Black-Right-Pointing-Pointer Interfusion reduced by between metal fuel (U-10 wt% Zr) and a HT9 cladding material. Black-Right-Pointing-Pointer Hydrogenation of the ZON during growth improved the FCCI barrier performance.

  6. Demonstration of fuel resistant to pellet-cladding interaction. Phase 2. First semiannual report, January-June 1979. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbaum, H.S. (comp.)

    1979-08-01

    This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress and reactive fission products during reactor service. This is the first semiannual progress report for Phase 2 of this program (January-June 1979). Progress in the irradiation testing of barrier fuel and of unfueled barrier cladding specimens is reported.

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

    Science.gov (United States)

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

    2017-04-01

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

  8. Effects of CeO2 on microstructure and corrosion resistance of TiC-VC reinforced Fe-based laser cladding layers

    Institute of Scientific and Technical Information of China (English)

    张辉; 邹勇; 邹增大; 史传伟

    2014-01-01

    The effects of CeO2 on microstructure and corrosion resistance of TiC-VC reinforced Fe-based laser cladding layers were investigated. The results showed that carbides presented in cladding layers were TiVC2 and VC. A small quantity of CeC appeared with 2.0 wt.%CeO2 addition. The amount of lamellar pearlite increased while the amount of residual austenite decreased with in-creasing CeO2 addition. The corrosion resistance of cladding layers increased firstly and then decreased with the addition of CeO2 in-creasing. The EIS spectrum of the cladding layer without CeO2 was composed of an inductive arc at low frequency and a capacitive arc at high frequency. The cladding layer with 0.5 wt.%CeO2 addition showed the best corrosion resistance, and then the inductive arc at low frequency transformed into a capacitive arc.

  9. Laser Cladding of an Al-11.7Wt% Si Alloy on ZM5 Magnesium Alloy to Enhance the Corrosion Resistance

    Institute of Scientific and Technical Information of China (English)

    CHEN Chang-jun; WANG Mao-cai; WANG Dong-sheng

    2004-01-01

    Magnesium alloy is an important engineering materials, but the wider application is restricted by poor corrosion resistance. An attempt was made to enhance the corrosion resistance and microhardness of a Mg-Al base ZM5 alloy by laser cladding of Al-11.7Wt%Si alloy powder with thickness 1.1mm and 1.7mm. The microstructure, phase and corrosion properties were analyzed by scanning electron micrographic (SEM), electron probe microanalysis(EPMA), vicker hardness tester and corrosion measurement system, respectively. Microhardness of the cladding layer was enhanced to 150-375Hv as compared to 60-99Hv of the substrate. The corrosion potential (Ecorr) of the cladding sample was 80mv higher than the substrate, while the corrosion current (Icorr) was lower than the substrate.

  10. Laser Cladding of an Al-11.7Wt% Si Alloy on ZM5 Magnesium Alloy to Enhance the Corrosion Resistance

    Institute of Scientific and Technical Information of China (English)

    CHENChang-jun; WANGMao-cai; WANGDong-sheng

    2004-01-01

    Magnesium alloy is an important engineering materials, but the wider application is restricted by poor corrosion resistance. An attempt was made to enhance the corrosion resistance and microhardness of a Mg-Al base ZM5 alloy by laser cladding of A1-11.7Wt%Si alloy powder with thickness 1.1 mm and 1.7inm. The microstructure, phase and corrosion properties were analyzed by scanning electron micrographic (SEM), electron probe microanalysis(EPMA), vicker hardness tester and corrosion measurement system, respectively. Microhardness of the cladding layer was enhanced to 150-375Hv as compared to 60-99Hv of the substrate. The corrosion potential (Ecorr) of the cladding sample was 80mv higher than the substrate, while the corrosion current (lcorr) was lower than the substrate.

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

    Science.gov (United States)

    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.

  12. Two dimensional structural analysis of reactor fuel element claddings due to local effects

    Energy Technology Data Exchange (ETDEWEB)

    Karimi, R; Wolf, L

    1978-04-01

    Two dimensional thermoelastic and inelastic stresses and deformation of typical LWR (PWR) and LMFBR (CRBR) claddings are evaluated by utilizing the following codes, for (1) Thermoelastic analysis (a) STRESS Code (b) SEGPIPE Code (2) Thermoinelastic analysis (a) Modified version of the GOGO code (b) One dimensional GRO-II code. The primary objective of this study is to analyze the effect of various local perturbations in the clad temperature field, namely eccentrically mounted fuel pellet, clad ovality, power tilt across the fuel and clad-coolant heat transfer variation on the cladding stress and deformation. In view of the fact that the thermoelastic analysis is always the first logical choice entering the structural field, it was decided to start the analysis with the two dimensional codes such as STRESS and SEGPIPE. Later, in order to assess the validity and compare the thermoelastic results to those obtained for actual reactor conditions, a two dimensional code, namely a modified version of the GOGO code, was used to account for inelastic effects such as irradiation and thermal creep and swelling in the evaluation. The comparison of thermoelastic and inelastic results shows that the former can be used effectively to analyze LWR fuel pin over 350 hours of lifetime under the most adverse condition and 500 hours of lifetime for an LMFBR fuel pin. Beyond that the inelastic solution must be used. The impact of the individual thermal perturbation and combinations thereof upon the structural quantity is also shown. Finally, the effect of rod displacement on the two dimensional thermal and structural quantities of the LMFBR fuel pin cladding is analyzed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-12-12

    Our work involved the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important parameters, and the development of a code that can be used by reactor operators to actively manage the accumulation of corrosion damage to the fuel cladding and other components in the heat transport circuits in both BWRs and PWRs. In addition, the modified boiling crevice model has been further developed to describe the accumulation of solutes in porous deposits (CRUD) on fuel under boiling (BWRs) and nucleate boiling (PWRs) conditions, in order to accurately describe the environment that is contact with the Zircaloy cladding. In the current report, we have derived expressions for the total steady-state current density and the partial anodic and cathodic current densities to establish a deterministic basis for describing Zircaloy oxidation. The models are “deterministic” because the relevant natural laws are satisfied explicitly, most importantly the conversation of mass and charge and the equivalence of mass and charge (Faraday’s law). Cathodic reactions (oxygen reduction and hydrogen evolution) are also included in the models, because there is evidence that they control the rate of the overall passive film formation process. Under open circuit conditions, the cathodic reactions, which must occur at the same rate as the zirconium oxidation reaction, are instrumental in determining the corrosion potential and hence the thickness of the barrier and outer layers of the passive film. Controlled hydrodynamic methods have been used to measure important parameters in the modified Point Defect Model (PDM), which is now being used to describe the growth and breakdown of the passive film on zirconium and on Zircaloy fuel sheathing in BWRs and PWRs coolant environments. The modified PDMs recognize the existence of a

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  17. Microstructure, Wear, and Corrosion Characteristics of TiC-Laser Surface Cladding on Low-Carbon Steel

    Science.gov (United States)

    El-Labban, Hashem F.; Mahmoud, Essam Rabea Ibrahim; Algahtani, Ali

    2016-04-01

    Laser cladding was used to produce surface composite layer reinforced with TiC particles on low-carbon steel alloy for improving the wear and corrosion resistances. The cladding process was carried out at powers of 2800, 2000, 1500, and 1000 W, and a fixed traveling speed of 4 mm/s. The produced layers are free from any cracks. Some of the TiC particles were melted and then re-solidified in the form of fine acicular dendrites. The amount of the melted TiC was increased by increasing the laser power. The hardness of the produced layers was improved by about 19 times of the base metal. Decreasing laser power led to hardness increment at the free surface. The improvement in wear resistance was reached to about 25 times (in case of 1500 W) of the base metal. Moreover, the corrosion resistance shows remarkable improvement after the laser treatment.

  18. Laser cladding of Zr-based coating on AZ91D magnesium alloy for improvement of wear and corrosion resistance

    Indian Academy of Sciences (India)

    Kaijin Huang; Xin Lin; Changsheng Xie; T M Yue

    2013-02-01

    To improve the wear and corrosion resistance of AZ91D magnesium alloy, Zr-based coating made of Zr powder was fabricated on AZ91D magnesium alloy by laser cladding. The microstructure of the coating was characterized by XRD, SEM and TEM techniques. The wear resistance of the coating was evaluated under dry sliding wear test condition at room temperature. The corrosion resistance of the coating was tested in simulated body fluid. The results show that the coating mainly consists of Zr, zirconium oxides and Zr aluminides. The coating exhibits excellent wear resistance due to the high microhardness of the coating. The main wear mechanism of the coating and the AZ91D sample are different, the former is abrasive wear and the latter is adhesive wear. The coating compared to AZ91D magnesium alloy exhibits good corrosion resistance because of the good corrosion resistance of Zr, zirconium oxides and Zr aluminides in the coating.

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

    Energy Technology Data Exchange (ETDEWEB)

    Isabella J van Rooyen

    2012-09-01

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

  20. Examination of the chemical composition of irradiated zirconium based fuel claddings at the metal/oxide interface by TEM

    Science.gov (United States)

    Abolhassani, S.; Bart, G.; Jakob, A.

    2010-04-01

    Detailed post-irradiation examinations have been performed at PSI on three fuel rods with differing cladding materials revealing different corrosion behaviour. The rods had been irradiated for 3-5 cycles at Gösgen nuclear power plant (pressurised water reactor), Switzerland. As zirconium corrosion is proceeding at the metal/oxide interface, extended micro-structural analyses were performed by transmission electron microscopy (TEM), expecting to possibly reveal phenomena explaining the varying corrosion resistance. This paper reports on the distribution of oxygen at the metal/oxide interface examined by energy dispersive X-ray spectroscopy (EDS) in TEM, while other micro-structural investigations have been published earlier [1]. In order to get some statistical confidence in the analyses, three neighbouring TEM samples of each cladding variant were studied. The oxygen concentration profiles of the three alloys (i.e. low-tin Zircaloy-4, Zr2.5%Nb and extra low-tin (Sn 0.56%)) both in the oxide and metal close to the metal/oxide interface are compared. The results of the examinations show the composition of the oxide in the vicinity of the interface to be sub-stoichiometric for all three materials, indicating an oxide layer adjacent to the interface, with diffusion-controlled access of oxygen to the metal/oxide interface. The metallic parts show highest oxygen concentrations at the metal/oxide interface which are reduced towards the bulk metal, pointing towards the expected second diffusion-controlled process leading to α-Zr (O). Based on the experimental results values for the diffusion coefficients in the range of 0.8-6.0 × 10 -20 m 2 s -1 are estimated for the oxygen dissolution process, the diffusion coefficient in Zircaloy-4 being six times higher than for the other two less corroding alloys. This finding is in contradiction with the present assumptions about the corrosion mechanism, and confirms the expected but not so far reported diffusion controlled

  1. Non-destructive control of cladding thickness of fuel elements for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karlov, Y.; Zhukov, Y.; Chashchin, S

    1997-07-01

    The control method of fuel elements for research reactors by means of measuring beta particles back scattering made it possible to perform complete automatic non-destructive control of internal and external claddings at our plant. This control gives high guarantees of the fuel element correspondence to the requirements. The method can be used to control the three-layer items of different geometry, including plates. (author)

  2. The oxidation and hydriding of zircaloy fuel cladding in high temperature aqueous solutions

    Science.gov (United States)

    Chen, Yingzi

    Nearly 90% of today's fission reactors use Zr based fuel cladding materials. The Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) are the two most common water-cooled nuclear reactors. Corrosion is the principal threat to the failure of the fuel in these reactors, resulting in the release of fission products to the coolant and hence to the establishment of radiation fields in out-of-core regions of the coolant circuit (e.g., steam generators in PWRs and turbines in BWRs). As is well known, corrosion is an electrochemical phenomenon; however, electrochemical effects are often neglected in corrosion studies on zirconium and its alloys, because of the difficulty in performing well-defined experiments under the appropriate conditions (high temperatures and pressures). In-situ studies have been carried out to examine the electrochemistry of passive zirconium under simulated BWR and PWR coolant conditions by using a controlled hydrodynamic, high temperature/high pressure test cell. The oxidation/hydriding mechanisms are elucidated by measuring the current, impedance, and capacitance of passive zirconium as a function of formation potential. The data are interpreted in terms of a modified point defect model (PDM) that recognize the existence of a passive film comprising a thick oxide outer layer over a thin barrier layer. From the composition of the zirconium passive film and thermodynamic analysis, it is postulated that a hydride barrier layer forms under PWR coolant conditions whereas an oxide barrier layer forms under BWR primary coolant conditions. Transients in current density and the thickness of the passive film formed on zirconium, when stepping the potential in either the positive or negative directions, have confirmed that the rate law afforded by the PDM adequately describes the growth and thinning of the passive film at high temperatures. The experimental results demonstrate that the kinetics of either oxygen or hydrogen vacancy generation

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

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Denis, Alicia [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: denis@cnea.gov.ar

    2008-02-29

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

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

    Science.gov (United States)

    Soba, Alejandro; Denis, Alicia

    2008-02-01

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

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

    Science.gov (United States)

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

    2017-04-01

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

  6. Increased local corrosion of SVEA-96 fuel assemblies in KKL. Final report; Erhoehte lokale Korrosion von SVEA-96-Brennelementen im KKL. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-11-15

    In February 1997 it was noted that the cladding surface below the spacers of SVEA-96 fuel assemblies showed increased local corrosion. This phenomenon called 'Enhanced Spacer Shadow Corrosion' (ESSC) by the fuel supplier was carefully monitored during the following 4 years. Several measures have been taken in order to counteract this ESSC. In this report of the Swiss Federal Agency for the Safety of Nuclear Installations (HSK) a summary is given of the technical and licensing aspects of ESSC. Although the fundamental mechanisms for the occurrence of ESSC are not yet sufficiently understood, short-term modification to water chemistry and the increasing use of improved cladding materials have effectively reduced this phenomenon. For the justification of the use of ESSC-damaged SVEA-96 fuel assemblies, HSK established temporary criteria which are based on technical investigations by the fuel assembly supplier. Among these, a special mention can be made of the more restrictive thermo-mechanical operation limit (TMOL) curve. As proof with respect of the HSK criteria, the plant operator conducted extended inspections on fuel assemblies during serving periods in 1997-2001 (measurement of oxide thickness). The conservative aspect of the measurements was assured through destructive examinations carried out at the Hot Laboratory of the Paul Scherrer Institute (PSI). Based on the modified water chemistry and the design of the core loading for cycle 18 (2001/2002) which contains only ESSC resistant cladding materials (LK2+, LK3), the original licence basis concerning the tolerable oxide thickness on the cladding could be guarantied. This has been verified by the results of a fuel assembly examination in August 2001. Therefore, the problem of the increased corrosion of the cladding of the SVEA-96 fuel assemblies is considered as being solved

  7. Corrosion Studies of Platinum Nano-Particles for Fuel Cells

    DEFF Research Database (Denmark)

    Shim, Signe Sarah

    The main focus of the present thesis is on corrosion and prevention of corrosion of platinum particles supported on carbon. This is important for instance in connection with start up and shutdown of fuel cells. The degradation mechanism of platinum particles supported on carbon has been...

  8. Modelling anelastic contribution to nuclear fuel cladding creep and stress relaxation

    Energy Technology Data Exchange (ETDEWEB)

    Tulkki, Ville, E-mail: ville.tulkki@vtt.fi; Ikonen, Timo

    2015-10-15

    In fuel behaviour modelling accurate description of the cladding mechanical response is important for both operational and safety considerations. While accuracy is desired, a certain level of simplicity is needed as both computational resources and detailed information on properties of particular cladding may be limited. Most models currently used in the integral codes divide the mechanical response into elastic and viscoplastic contributions. These have difficulties in describing both creep and stress relaxation, and often separate models for the two phenomena are used. In this paper we implement anelastic contribution to the cladding mechanical model, thus enabling consistent modelling of both creep and stress relaxation. We show that the model based on assumption of viscoelastic behaviour can be used to explain several experimental observations in transient situations and compare the model to published set of creep and stress relaxation experiments performed on similar samples. Based on the analysis presented we argue that the inclusion of anelastic contribution to the cladding mechanical models provides a way to improve the simulation of cladding behaviour during operational transients.

  9. Effect of Copper and Bronze Addition on Corrosion Resistance of Alloyed 316L Stainless Steel Cladded on Plain Carbon Steel by Powder Metallurgy

    Institute of Scientific and Technical Information of China (English)

    Wenjue CHEN; Yueying WU; Jianian SHEN

    2004-01-01

    A sandwich structure with cladding alloyed 316L stainless steel on plain carbon steel was prepared by means of powder metallurgy (PM) processing. Electrolytic Cu and prealloyed bronze (95Cu wt pct, 5Sn wt pct) were added in different contents up to 15% into the surface cladded 316L layers and the effect of alloying concentrations on the corrosion resistance of the 316L cladding layers was studied. The corrosion performances of the cladding samples were studied by immersion tests and potentio-dynamic anodic polarization tests in H2SO4 and FeCl3 solutions. Both 316L and alloyed 316L surface layers with 1.0 mm depth produced by PM cladding had an effect to improve corrosion resistance in H2SO4 and FeCl3 solutions. Small Cu and bronze addition (4%) had a positive effect in H2SO4 and FeCl3 solutions. 4% Cu alloyed 316L surface layer produced by PM cladding showed similar anodic polarization behaviour to the 316L cladding layer in H2SO4 and FeCl3 solutions.

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

    Science.gov (United States)

    2013-06-01

    Loss of Coolant Accident LWR Light Water Reactor MOX Mixed Oxide Fuel MTC Moderator Temperature Coefficient MWd/kgIHM Megawatt days per...working only with UO2 and UO2/PuO2 mixed oxide ( MOX ) fuels. 3.1 Studsvik Core Management Software CASMO-4E and SIMULATE-3 are the primary computational

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

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B.

    1989-10-01

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

  12. Measurement of Nucleate Pool Boiling Heat Transfer Limit using Fuel Cladding Material

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chi Young; Shin, Chang Hwan; Oh, Dong Seok; Chun, Tae Hyun; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    Zircaloy has been widely used as a fuel cladding material of light water reactor for more than three decades because it has a lower neutron absorption cross section and cracking rate. Recently, HANA-6 has been developed in KAERI (Korea Atomic Energy Research Institute) as the advanced fuel cladding for high burn-up fuel. Generally, under the normal and accident operating conditions of a nuclear reactor, the nuclear fuel cladding of zirconium based alloys undergoes the surface change, and the oxide layer can be formed. In such a case, the previous CHF correlations should be assessed and examined using the experimental results for not a fresh zircaloy surface but an oxidized one, to predict and examine the thermal margin and safety of a nuclear reactor core. Therefore, the experimental data using the oxidized zircaloy surface need to be provided quantitatively. In this paper, the CHF in saturated water pool boiling is measured and discussed using the specimens of zircaloy-4, HANA-6, and oxidized zircaloy-4 in high temperature air environment. The CHF of zircaloy-4, HANA-6, and oxidized surface was tested. Zircaloy-4 and HANA-6 had a similar CHF performance. This is because both are the zirconium based alloys, and appear the almost same water contact angle. On the other hands, the oxidized specimen became to be higher CHF than plain zircaloy-4 and HANA-6 specimens, due to smaller water contact angle (i. e., good hydrophilicity of specimen). The Kandlikar's (2001) correlation reasonably predicted the present experimental data.

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

    Science.gov (United States)

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

    2017-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

  15. Protective Coatings for Wet Storage of Aluminium-Clad Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, S.M.C.; Correa, O.V.; Souza, J.A. De; Ramanathan, L.V. [Materials science and Technology Center, Instituto de Pesquisas Energeticas e Nucleares - IPEN, Av. Prof. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)

    2011-07-01

    Corrosion protection of spent RR fuel for long term wet storage was considered important, primarily from the safety standpoint and the use of conversion coatings was proposed in 2008. This paper presents the results of: (a) on-going field tests in which un-coated and lanthanide-based conversion coated Al alloy coupons were exposed to the IEA-R1 reactor spent fuel basin for durations of up to a year; (b) preparation of cerium modified hydrotalcite coatings and cerium sealed boehmite coatings on AA 6061 alloy; (c) corrosion resistance of coated specimens in NaCl solutions. The field studies indicated that the oxidized and cerium dioxide coated coupons were the most corrosion resistant. The cerium modified hydrotalcite and cerium sealed boehmite coated specimens showed marked increase in pitting corrosion resistance. (author)

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

    Science.gov (United States)

    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).

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

    Science.gov (United States)

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

    2017-07-25

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

  18. Aluminum cladding oxidation of prefilmed in-pile fueled experiments

    Science.gov (United States)

    Marcum, W. R.; Wachs, D. M.; Robinson, A. B.; Lillo, M. A.

    2016-04-01

    A series of fueled irradiation experiments were recently completed within the Advanced Test Reactor Full size plate In center flux trap Position (AFIP) and Gas Test Loop (GTL) campaigns. The conduct of the AFIP experiments supports ongoing efforts within the global threat reduction initiative (GTRI) to qualify a new ultra-high loading density low enriched uranium-molybdenum fuel. This study details the characterization of oxide growth on the fueled AFIP experiments and cross-correlates the empirically measured oxide thickness values to existing oxide growth correlations and convective heat transfer correlations that have traditionally been utilized for such an application. This study adds new and valuable empirical data to the scientific community with respect to oxide growth measurements of highly irradiated experiments, of which there is presently very limited data. Additionally, the predicted oxide thickness values are reconstructed to produce an oxide thickness distribution across the length of each fueled experiment (a new application and presentation of information that has not previously been obtainable in open literature); the predicted distributions are compared against experimental data and in general agree well with the exception of select outliers.

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

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Sharon M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chattin, Marc Rhea [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giaquinto, Joseph M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-09-01

    To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified.To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. A sample of Surry-2 pressurized water reactor (PWR) cladding was heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. The tritium content was measured to be ~240 µCi/g. Cladding samples were heated to 500ºC, which is within the temperature range (480 - 600ºC) expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. Heating at 500°C for 24 hr removes ~0.2% of the tritium from the cladding, and heating at 700°C for 24 hr removes ~9%. Thus, a significant fraction of the tritium remains bound in the cladding and must be considered in operations involving cladding recycle.

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

    Science.gov (United States)

    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.

  1. KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Gyun; Kim, Young Il

    2006-12-15

    Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006.

  2. Microstructures and properties of low-chromium high corrosion-resistant TiC-VC reinforced Fe-based laser cladding layer

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hui; Zou, Yong, E-mail: yzou@sdu.edu.cn; Zou, Zengda; Wu, Dongting

    2015-02-15

    Highlights: • The cladding layer with 3.0%Cr and 0.25%CeO{sub 2} showed a good corrosion resistance. • Passive film formed on the cladding layer without Cr and CeO{sub 2} was Fe{sub 3}O{sub 4}. • Fe{sub 3}O{sub 4} displayed p type semiconductivity. • Passive film formed on the cladding layer with Cr and CeO{sub 2} was Fe(OH){sub 3} and Cr(OH){sub 3}. • Fe(OH){sub 3} displayed n type while Cr(OH){sub 3} displayed p type semiconductivity. - Abstract: Effects of 3.0 wt.%Cr and/or 0.25 wt.%CeO{sub 2} on microstructures and properties of TiC-VC reinforced Fe-based cladding layer were investigated by using X-ray diffractometry (XRD), scanning electron microscopy (SEM), and electrochemical impedance spectroscopy (EIS). Passive films formed on cladding layers surface were investigated by using X-ray photoelectron spectroscopy (XPS) and Mott-Schottky analysis. Results showed that phases of cladding layers were α-Fe, γ-Fe, TiC, VC and TiVC{sub 2}. There were no obvious effects of adding 3.0 wt.%Cr and/or 0.25 wt.%CeO{sub 2} on cladding layers phases. The microstructure of the cladding layer with 3.0 wt.%Cr and 0.25 wt.%CeO{sub 2} was lath martensite and retained austenite. Microhardness of the cladding layer with 0.25 wt.%CeO{sub 2} decreased slightly. Microhardness and corrosion resistance of the cladding layer with 3.0 wt.%Cr and 0.25 wt.%CeO{sub 2} both increased, the corrosion resistance increased 7.33 times while the EIS Nyquist spectrum transformed into a capacitive arc. The passive film formed on the cladding layer without Cr and CeO{sub 2} was Fe{sub 3}O{sub 4} which displayed p type semiconductivity. The passive film formed on the cladding layer with 3.0 wt.%Cr and 0.25 wt.%CeO{sub 2} was composed of Fe(OH){sub 3} and Cr(OH){sub 3}, which displayed n and p type semiconductivity respectively.

  3. New cladding materials and evolution of nuclear fuel components for PWR; Nouveaux materiaux de gainage et evolution des produits de combustible REP

    Energy Technology Data Exchange (ETDEWEB)

    Aubry, S. [Electricite de France (EDF), EDF Div. Combustible Nucleaire, 92 - Clamart (France); Francillon, E. [FRAMATOME ANP, Secteur Combustible, 92 - Paris-La-Defence (France); Guillet, J.L. [CEA Saclay, Dir. du Soutien Nucleaire Industriel, 91 - Gif-sur-Yvette (France)

    2004-07-01

    This paper presents recent improvements in the field of nuclear fuels made by Framatome-ANP. The first one is the use of the M5 (trade mark) alloy for the fuel cladding and guide tubes. This alloys is composed of zirconium, niobium and oxygen, it follows an optimized industrial fabrication process, it can bear combustion rates over 70 GWd/t even in harsh conditions and is strongly resistant to corrosion. Other improvements have been made in the design of the fuel assembly structure, it concerns the lower part of the one-piece tube guide for control rods and the bi-grid device whose purpose is to hold better the fuel assembly in order to reduce the fretting wear on the lower part of fuel rods. Another improvement is the doping of fuel pellets with chromium that allows, combined with an optimized micro-structure, the reduction of the volume of the gaseous fission products released in the fuel. (A.C.)

  4. Mechanical Property and Oxidation Behavior of ATF cladding developed in KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    To realize the coating cladding, coating material (Cr-based alloy) as well as coating technology (3D laser coating and arc ion plating combined with vacuum annealing) can be developed to meet the fuel cladding criteria. The coated Zr cladding can be produced after the optimization of coating technologies. The coated cladding sample showed the good oxidation/corrosion and adhesion properties without the spalling and/or severe interaction with the Zr alloy cladding from the various tests. Thus, it is known that the mechanical property and oxidation behavior of coated cladding concept developed in KAERI is reasonable for applying the ATF cladding in LWRs. At the present time various ATF concepts have been proposed and developing in many countries. The ATF concepts with potentially improved accident performance can be summarized to the coating cladding, Mo-Zr cladding, FeCrAl cladding, and SiCf/SiC cladding. Regarding the cladding performance, ATF cladding concepts will be evaluated with respect to the accident scenarios and normal operations of LWRs as well as to the fuel cladding fabrication.

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

    OpenAIRE

    Shengli Chen; Cenxi Yuan

    2017-01-01

    Neutronic performance is investigated for a potential accident tolerant fuel (ATF), which consists of U3Si2 fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2 has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into con...

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

    Science.gov (United States)

    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.

  7. Fabrication and measurement of hoop strength of SiC triplex tube for nuclear fuel cladding applications

    Science.gov (United States)

    Kim, Daejong; Lee, Hyun-Geun; Park, Ji Yeon; Kim, Weon-Ju

    2015-03-01

    The SiC ceramics are under investigation for the fuel cladding in the light water nuclear reactors because of its excellent high temperature strength and corrosion resistance against hot steam under the severe accident conditions. In this study, the SiC triplex tubes consisting of a SiC inner layer, a SiC/PyC/SiC intermediate layer, and a SiC outer layer were fabricated by the chemical vapor processes. The hoop strength and fracture behaviors of the SiC triplex tube were investigated. The SiC triplex tubes fabricated at the high ratio of H2/MTS had a quite high average strength with a relatively small standard deviation. The hoop strength of the composite tubes tends to increase with the volume fraction of the reinforced fibers. The highest fiber volume fraction was obtained using Tyranno SA3-0.8k with the dense winding patterns such as bamboo-like mosaic pattern, which resulted in the high hoop strength compared to other fibers of Tyranno SA3-1.6k and Hi-Nicalon Type S. Hoop strength also increased slightly as the winding angle increased from 45° to 65°. Fracture behaviors of the SiC triplex tube were investigated via the observation of microstructure of the failed samples.

  8. Fabrication and measurement of hoop strength of SiC triplex tube for nuclear fuel cladding applications

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Daejong, E-mail: dkim@kaeri.re.kr; Lee, Hyun-Geun; Park, Ji Yeon; Kim, Weon-Ju

    2015-03-15

    The SiC ceramics are under investigation for the fuel cladding in the light water nuclear reactors because of its excellent high temperature strength and corrosion resistance against hot steam under the severe accident conditions. In this study, the SiC triplex tubes consisting of a SiC inner layer, a SiC/PyC/SiC intermediate layer, and a SiC outer layer were fabricated by the chemical vapor processes. The hoop strength and fracture behaviors of the SiC triplex tube were investigated. The SiC triplex tubes fabricated at the high ratio of H{sub 2}/MTS had a quite high average strength with a relatively small standard deviation. The hoop strength of the composite tubes tends to increase with the volume fraction of the reinforced fibers. The highest fiber volume fraction was obtained using Tyranno SA3-0.8k with the dense winding patterns such as bamboo-like mosaic pattern, which resulted in the high hoop strength compared to other fibers of Tyranno SA3-1.6k and Hi-Nicalon Type S. Hoop strength also increased slightly as the winding angle increased from 45° to 65°. Fracture behaviors of the SiC triplex tube were investigated via the observation of microstructure of the failed samples.

  9. Corrosion report for the U-Mo fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Bennett, Wendy D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Fuller, E. S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hardy, John S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2014-08-28

    The Fuel Cycle Research and Development (FCRD) program of the Office of Nuclear Energy (NE) has implemented a program to develop a Uranium-Molybdenum (U-Mo) metal fuel for Light Water Reactors (LWR)s. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties, which includes high thermal conductivity for less stored heat energy. With sufficient development, it may be able to provide the Light Water industry with a melt-resistant accident tolerant fuel with improved safety response. However, the corrosion of this fuel in reactor water environments needs to be further explored and optimized by additional alloying. The Pacific Northwest National Laboratory has been tasked with performing ex-reactor corrosion testing to characterize the performance of U-Mo fuel. This report documents the results of the effort to characterize and develop the U-Mo metal fuel concept for LWRs with regard to corrosion testing. The results of a simple screening test in buffered water at 30°C using surface alloyed U-10Mo is documented and discussed. The screening test was used to guide the selection of several potential alloy improvements that were found and are recommended for further testing in autoclaves to simulate PWR water conditions more closely.

  10. GEH-4-63, 64: Proposal for irradiation of production brazed Zircaloy-2 clad fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Tverberg, J.C.

    1961-05-18

    A brazed end closure is currently being used on prototypical NPR fuel elements. The production closure will use a braze alloy composed of 5% Be + 95% Zry-2 to braze the Zircaloy-2 cap to the jacket and to the metallic uranium core. A similar MTR test, a GEH-4-57, 58, used a braze alloy of the composition 4% Be + 12% Fe + 84% Zry-2 which melts at a lower temperature. In this previous test, element GEH-4-57 failed through a cladding defect located at the base of the braze heat affected zone. Because of this failure it would be desirable to subject a fuel element, which had been subjected to more severe brazing conditions, to the same conditions as GEH-4-57, 58. For this reason the thermal conditions of this test essentially match those of GEH-4-57, 58. This irradiation test consists of two identical fuel elements. The fuel material is normal metallic uranium, Zircaloy-2 clad of the tubular geometry, NPR inner size. The fuel was coextruded at Hanford by General Electric`s Fuels Preparation Department. Each element is 10.8 inches in length with flat Zircaloy-2 end caps brazed to the jacket and uranium core with the 5 Be + 95 Zry-2 brazing alloy, then TIG welded to further insure closure integrity. The elements ar 1.254 inches OD and 0.439 inches ID. For hydraulic purposes a 0.343 inch diamater flow restrictor has been fitted into the central flow channel of both elements.

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

    Energy Technology Data Exchange (ETDEWEB)

    George W. Griffith

    2011-10-01

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

  12. RF Plasma Torch System for Metal Matrix Composite Production in Nuclear Fuel Cladding

    Science.gov (United States)

    Holik, Eddie, III

    2007-10-01

    For the first time in 30 years, plans are afoot to build new fission power plants in the US. It is timely to develop technology that could improve the safety and efficiency of new reactors. A program of development for advanced fuel cycles and Generation IV reactors is underway. The path to greater efficiency is to increase the core operating temperature. That places particular challenges to the cladding tubes that contain the fission fuel. A promising material for this purpose is a metal matrix composite (MMC) in which ceramic fibers are bonded within a high-strength steel matrix, much like fiberglass. Current MMC technology lacks the ability to effectively bond traditional high-temperature alloys to ceramic strands. The purpose of this project is to design an rf plasma torch system to use titanium as a buffer between the ceramic fibers and the refractory outer material. The design and methods of using an rf plasma torch to produce a non-equilibrium phase reaction to bond together the MMC will be discussed. The effects of having a long lived fuel cladding in the design of future reactors will also be discussed.

  13. Estimation of ring tensile properties of steam oxidized Zircaloy-4 fuel cladding under simulated LOCA condition

    Science.gov (United States)

    Shriwastaw, R. S.; Sawarn, Tapan K.; Banerjee, Suparna; Rath, B. N.; Dubey, J. S.; Kumar, Sunil; Singh, J. L.; Bhasin, Vivek

    2017-09-01

    The present study involves the estimation of ring tensile properties of Indian Pressurised Heavy Water Reactor (IPHWR) fuel cladding made of Zircaloy-4, subjected to experiments under a simulated loss-of-coolant-accident (LOCA) condition. Isothermal steam oxidation experiments were conducted on clad tube specimens at temperatures ranging from 900 to 1200 °C at an interval of 50 °C for different soaking periods with subsequent quenching in water at ambient temperature. The specimens, which survived quenching, were then subjected to ambient temperature ring tension test (RTT). The microstructure was correlated with the mechanical properties. The yield strength (YS) and ultimate tensile strength (UTS) increased initially with rise in oxidation temperature and time duration but then decreased with further increase in oxidation. Ductility is adversely affected with rising oxidation temperature and longer holding time. A higher fraction of load bearing phase and lower oxygen content in it ensures higher residual ductility. Cladding shows almost zero ductility behavior in RIT when load bearing phase fraction is less than 0.72 and its average oxygen concentration is greater than 0.58 wt%.

  14. New method to calculate the mechanical properties of unirradiated fuel cladding from ring tensile tests

    Energy Technology Data Exchange (ETDEWEB)

    Martin-Rengel, M.A. [Departamento de Ciencia de Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren s/n, E-28040 Madrid (Spain); Consejo de Seguridad Nuclear (CSN), Justo Dorado 11, E-28040 Madrid (Spain); Gomez, F.J.; Ruiz-Hervias, J.; Caballero, L.; Valiente, A. [Departamento de Ciencia de Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren s/n, E-28040 Madrid (Spain)

    2009-06-15

    Nuclear fuel cladding is the first barrier used to confine the fuel and the fission products produced during irradiation. Zirconium alloys are used for this purpose due to their remarkable neutron transparency, together with their good mechanical properties at operational temperatures. Consequently, it is very important to be able to characterize the mechanical response of the irradiated cladding. The mechanical behaviour of the material can be modelled as elastoplastic with different stress-strain curves depending on the direction: radial, hoop or longitudinal direction. The ring tensile test has been proposed to determine the mechanical properties of the cladding along the hoop direction. The initial test consisted of applying a force inside the tube, by means of two half cylinders. Later Arsene and Bai [1,2] modified the experimental device to avoid tube bending at the beginning of the test. The same authors proposed a numerical method to obtain the stress-strain curve in the hoop direction from the experimental load versus displacement results and a given friction coefficient between the loading pieces and the sample [3]. This method has been used by different authors [4] with slight modifications. It is based on the existence of two universal curves under small strain hypothesis: the first correlating the hoop strain and the displacement of the loading piece and the second one correlating the hoop stress and the applied load. In this work, a new method to determine the mechanical properties of the cladding from the ring tensile test results is proposed. Non-linear geometry is considered and an iterative procedure is proposed so universal curves are not needed. A stress-strain curve is determined by combining numerical calculations with experimental results in a convergent loop. The two universal curves proposed by Arsene and Bai [3] are substituted by two relationships, one between the equivalent plastic strain in the centre of the specimen ligament and the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

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

  16. Corrosion protected, multi-layer fuel cell interface

    Science.gov (United States)

    Feigenbaum, Haim; Pudick, Sheldon; Wang, Chiu L.

    1986-01-01

    An improved interface configuration for use between adjacent elements of a fuel cell stack. The interface is impervious to gas and liquid and provides resistance to corrosion by the electrolyte of the fuel cell. The multi-layer configuration for the interface comprises a non-cupreous metal-coated metallic element to which is film-bonded a conductive layer by hot pressing a resin therebetween. The multi-layer arrangement provides bridging electrical contact.

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

    Science.gov (United States)

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

    2012-09-01

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

  18. Fireside Corrosion in Oxy-fuel Combustion of Coal

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, Gordon R [National Energy Technology Laboratory; Tylczak, Joseph [National Energy Technology Laboratory; Meier, Gerald H [University of Pittsburgh; Lutz, Bradley [University of Pittsburgh; Jung, Keeyoung [Institute of Industrial Science and Technology, Korea; Mu, Nan; Yanar, Nazik M [University of Pittsburgh; Pettit, Frederick S [University of Pittsburgh; Zhu, Jingxi [Carnegie Mellon University; Wise, Adam [Carnegie Mellon University; Laughlin, David E. [Carnegie Mellon University; Sridhar, Seetharaman [Carnegie Mellon University

    2013-11-25

    Oxy-fuel combustion is burning a fuel in oxygen rather than air for ease of capture of CO2 from for reuse or sequestration. Corrosion issues associated with the environment change (replacement of much of the N2 with CO2 and higher sulfur levels) from air- to oxy-firing were examined. Alloys studied included model Fe–Cr alloys and commercial ferritic steels, austenitic steels, and nickel base superalloys. The corrosion behavior is described in terms of corrosion rates, scale morphologies, and scale/ash interactions for the different environmental conditions. Evidence was found for a hreshold for severe attack between 10-4 and 10-3 atm of SO3 at 700ºC.

  19. Corrosion studies in fuel element reprocessing environments containing nitric acid

    Energy Technology Data Exchange (ETDEWEB)

    Beavers, J A; White, R R; Berry, W E; Griess, J C

    1982-04-01

    Nitric acid is universally used in aqueous fuel element reprocessing plants; however, in the processing scheme being developed by the Consolidated Fuel Reprocessing Program, some of the equipment will be exposed to nitric acid under conditions not previously encountered in fuel element reprocessing plants. A previous report presented corrosion data obtained in hyperazeotropic nitric acid and in concentrated magnesium nitrate solutions used in its preparation. The results presented in this report are concerned with the following: (1) corrosion of titanium in nitric acid; (2) corrosion of nickel-base alloys in a nitric acid-hydrofluoric acid solution; (3) the formation of Cr(VI), which enhances corrosion, in nitric acid solutions; and (4) corrosion of mechanical pipe connectors in nitric acid. The results show that the corrosion rate of titanium increased with the refreshment rate of boiling nitric acid, but the effect diminished rapidly as the temperature decreased. The addition of iodic acid inhibited attack. Also, up to 200 ppM of fluoride in 70% HNO/sub 3/ had no major effect on the corrosion of either titanium or tantalum. In boiling 8 M HNO/sub 3/-0.05 M HF, Inconel 671 was more resistant than Inconel 690, but both alloys experienced end-grain attack. In the case of Inconel 671, heat treatment was very important; annealed and quenched material was much more resistant than furnace-cooled material.The rate of oxidation of Cr(III) to Cr(VI) increased significantly as the nitric acid concentration increased, and certain forms of ruthenium in the solution seemed to accelerate the rate of formation. Mechanical connectors of T-304L stainless steel experienced end-grain attack on the exposed pipe ends, and seal rings of both stainless steel and a titanium alloy (6% Al-4% V) underwent heavy attack in boiling 8 M HNO/sub 3/.

  20. Corrosion in Fuel/Natural Seawater Environments

    Science.gov (United States)

    2011-11-18

    converting the triglyceride oils to methyl (or ethyl) esters with a process known as transesterification.4 The transesterification process reacts alcohol...States the term "biodiesel" is standardized as fatty acid methyl ester (FAME) and is considered a first-generation biofuel. Biodiesel content is... kerosene -based fuel for use in aircraft turbine engines. Since the fuel is the primary aircraft fuel used aboard aircraft carriers, a substantially higher

  1. Surface treatment method for cladding tube of LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suda, Yoshitaka; Matsumoto, Kunio; Ito, Kenji.

    1994-06-07

    Upon surface finishing by polishing, shot peening or blasting is applied on the outer surface of a cladding tube to eliminate orientation of residual stresses on the surface layer in order to eliminate residual stresses formed on the outer surface in the circumferential direction. This can suppress occurrence of cracks in oxide membranes formed on the outer surface to suppress development of corrosion on the outer surface irrespective of the ingredient composition of fuel cladding tube made of zircaloy. (T.M.).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-01

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

  3. DIMENSIONALLY STABLE, CORROSION RESISTANT NUCLEAR FUEL

    Science.gov (United States)

    Kittel, J.H.

    1963-10-31

    A method of making a uranium alloy of improved corrosion resistance and dimensional stability is described. The alloy contains from 0-9 weight per cent of an additive of zirconium and niobium in the proportions by weight of 5 to 1 1/ 2. The alloy is cold rolled, heated to two different temperatures, air-cooled, heated to a third temperature, and quenched in water. (AEC)

  4. The problems of mass transfer and formation of deposits of corrosion products on fuel assemblies of a VVER-1200 reactor

    Science.gov (United States)

    Rodionov, Yu. A.; Kritskii, V. G.; Berezina, I. G.; Gavrilov, A. V.

    2014-03-01

    On the basis of examination of materials published both in Russia and abroad, as well as their own investigations, the authors explain the reasons for the occurrence of such effects as AOA (Axial Offset Anomalies) and an increase in the coolant pressure difference in the core of nuclear reactors of the VVER type. To detect the occurrence of the AOA effect, the authors suggest using the specific activity of 58Co in the coolant. In the VVER-1200 design the thermohydraulic regime for fuel assemblies in the first year of their service life involves slight boiling of the coolant in the upper part of the core, which may induce the occurrence of the AOA effect, intensification of corrosion of fuel claddings, and abnormal increase in deposition of corrosion products. Radiolysis of the water coolant in the boiling section (boiling in pores of deposits) may intensify not only general corrosion but also a localized (nodular) one. As a result of intensification of the corrosion processes and growth of deposits, deterioration of the radiation situation in the rooms of the primary circuit of a VVER-1200 reactor as compared to that at nuclear power plants equipped with reactors of the VVER-1000 type is possible. Recommendations for preventing the AOA effect at nuclear power plants with VVER-1200 reactors on the matter of the direction of further investigations are made.

  5. On the relative role of processes whose sequence results in crack growth in the cladding of LMFBR fuel pins

    Science.gov (United States)

    Mikhlin, E. Ya.

    1991-08-01

    Processes are discussed the joint effect of which results in crack development in austenitic steel-clad oxide fuel pins. Such processes include generation of Te which is considered as the main embrittling agent, its transport and accumulation at the cladding inner surface, where together with Cs it forms a liquid surface-acting medium, and finally, development of intergranular cracks in the cladding caused by the contact with this medium. As the process of crack growth in itself proceeds faster than accumulation of liquid surfactants at the cladding, the cracks will be able to reach the critical length only after the necessary amount of Te has been accumulated. Its accumulation is determined and therefore, controlled by the process of Te transport in the fuel grains. It is shown that the main contribution to the accumulation of Te at the cladding surface is provided by the hottest internal zones of the fuel pellet. On the basis of the analysis given, means are discussed, for inhibiting or blocking the crack growth.

  6. Cold Spray Coating Technique with FeCrAl Alloy Powder for Developing Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Dong Jun; Kim, Hyun Gil; Park, Jeong Yong; Jung, Yang Il; Park, Jung Hwan; Koo, Yang Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Various approaches to enhance safety have been suggested, replacing current Zr-based alloys for fuel cladding with advanced materials exhibiting lower oxidation rates can be a basic solution. Many advanced materials such as FeCrAl alloys; Mn+1AXn, (MAX) phases, where n = 1 to 3, M is an early transition metal, A is an A-group (mostly IIIA and IVA, or groups 13 and 14) element and X is either carbon or nitrogen; Mo; and SiC are being considered as possible candidates. Among the proposed fuel cladding substitutes, Fe-based alloys are one of the most promising candidates owing to their excellent formability, high strength, and oxidation resistance at high temperature. In this work, the ATF technology concept of Fe-based alloy coating on the existing Zr-alloy cladding was considered and results on the optimization study for fabrication of coated tube samples were described. Result obtained from high temperature oxidation test under steam environment at 1200 .deg. C indicates that FeCrAl alloy coated Zr metal matrix may maintain its integrity during LOCA. This means that accident tolerance of FeCrAl alloy coated Zr cladding sample had been greatly improved compared to that of existing Zr-based alloy fuel cladding.

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

    Directory of Open Access Journals (Sweden)

    Shengli Chen

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-09-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-09-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-14

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

  12. The reliability of untempered end plug welds on HT9-clad IFR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, D C; Porter, D L

    1987-02-01

    Welding generally leaves residual stresses in transformed weld zones, which can initiate cracks from flaws already present in the weld zones. When HT9 cools from welding temperatures, a martensite phase forms in the weld fusion zone and heat-affected zone. Because this martensite phase is hard and brittle, it is particularly susceptible to cracking aggravated by residual stresses. This causes concern over the use of untempered welds on HT9-clad fuel elements. To determine if residual stresses present in end-plug weld zones would affect fuel pin performance, HT9 capsules with prototypic TIG- and CD-welded end plugs (in the tempered and as-welded conditions) were pressurized to failure at room temperature, 550{sup 0}C, and 600{sup 0}C. None of the capsules failed in a weld zone. To determine the effects of reactor operating temperatures on untempered welds, prototypic TIG welds were tempered at reactor bulk sodium temperature and an expected sodium outlet temperature for various lengths of time. Subsequent tensile and burst tests of these specimens proved that any embrittling effects that may have been induced in these welds were of no consequence. Hardness tests on longitudinal sections of welds indicated the amount of tempering a weld will receive inreactor after relatively short lengths of time. The pressure burst tests proved that untemperted welds on HT9-clad fuel elements are as reliable as tempered welds; any residual stresses in untempered weld zones were of no consequence. The tempering test showed that welds used in the as-welded condition will sufficiently temper in 7 days at 550{sup 0}C, but will not, sufficiently temper in 7 days at bulk sodium temperature. A comparison of the structure of laser welds to those of CD and TIG welds indicated that untempered laser welds will perform and temper in a manner similar to the TIG welds tested in this effort.

  13. Multilayer (TiN, TiAlN) ceramic coatings for nuclear fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Alat, Ece, E-mail: exa179@psu.edu [Department of Materials Science and Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Motta, Arthur T. [Department of Materials Science and Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Comstock, Robert J.; Partezana, Jonna M. [Westinghouse Electric Co., Beulah Rd, Pittsburgh, PA 1332 (United States); Wolfe, Douglas E. [Department of Materials Science and Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Applied Research Laboratory, The Pennsylvania State University, 119 Materials Research Building, University Park, PA 16802 (United States)

    2016-09-15

    In an attempt to develop an accident-tolerant fuel (ATF) that can delay the deleterious consequences of loss-of-coolant-accidents (LOCA), multilayer coatings were deposited onto ZIRLO{sup ®} coupon substrates by cathodic arc physical vapor deposition (CA-PVD). Coatings were composed of alternating TiN (top) and Ti{sub 1-x}Al{sub x}N (2-layer, 4-layer, 8-layer and 16-layer) layers. The minimum TiN top coating thickness and coating architecture were optimized for good corrosion and oxidation resistance. Corrosion tests were performed in static pure water at 360 °C and 18.7 MPa for up to 90 days. The optimized coatings showed no spallation/delamination and had a maximum of 6 mg/dm{sup 2} weight gain, which is 6 times smaller than that of a control sample of uncoated ZIRLO{sup ®} which showed a weight gain of 40.2 mg/dm{sup 2}. The optimized architecture features a ∼1 μm TiN top layer to prevent boehmite phase formation during corrosion and a TiN/TiAlN 8-layer architecture which provides the best corrosion performance. - Highlights: • The first study on multilayer TiAlN and TiN ceramic coatings on ZIRLO{sup ®} coupons. • Corrosion tests were performed at 360°C and 18.7 MPa for up to 90 days. • Coatings adhered well to the substrate, and showed no spallation/delamination. • Weight gains were six times lower than those of uncoated ZIRLO{sup ®} samples. • Longer and higher temperature corrosion tests will be discussed in a further paper.

  14. Carbon fuel cells with carbon corrosion suppression

    Science.gov (United States)

    Cooper, John F [Oakland, CA

    2012-04-10

    An electrochemical cell apparatus that can operate as either a fuel cell or a battery includes a cathode compartment, an anode compartment operatively connected to the cathode compartment, and a carbon fuel cell section connected to the anode compartment and the cathode compartment. An effusion plate is operatively positioned adjacent the anode compartment or the cathode compartment. The effusion plate allows passage of carbon dioxide. Carbon dioxide exhaust channels are operatively positioned in the electrochemical cell to direct the carbon dioxide from the electrochemical cell.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbaum, H.S. (comp.)

    1980-09-01

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

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

    Science.gov (United States)

    Rico, A.; Martin-Rengel, M. A.; Ruiz-Hervias, J.; Rodriguez, J.; Gomez-Sanchez, F. J.

    2014-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-15

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

  18. Obtention of the constitutive equation of hydride blisters in fuel cladding from nanoindentation tests

    Science.gov (United States)

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

    2017-04-01

    It is well known that the presence of hydrides in nuclear fuel cladding may reduce its mechanical and fracture properties. This situation may be worsened as a consequence of the formation of hydride blisters. These blisters are zones with an extremely high hydrogen concentration and they are usually associated to the oxide spalling which may occur at the outer surface of the cladding. In this work, a method which allows us to reproduce, in a reliable way, hydride blisters in the laboratory has been devised. Depth-sensing indentation tests with a spherical indenter were conducted on a hydride blister produced in the laboratory with the aim of measuring its mechanical behaviour. The plastic stress-strain curve of the hydride blister was calculated for first time by combining depth-sensing indentation tests results with an iterative algorithm using finite element simulations. The algorithm employed reduces, in each iteration, the differences between the numerical and the experimental results by modifying the stress-strain curve. In this way, an almost perfect adjustment of the experimental data was achieved after several iterations. The calculation of the constitutive equation of the blister from nanoindentation tests, may involve a lack of uniqueness. To evaluate it, a method based on the optimization of parameters of analytical equations has been proposed in this paper. An estimation of the error which involves this method is also provided.

  19. Radiation induced corrosion of copper for spent nuclear fuel storage

    Science.gov (United States)

    Björkbacka, Åsa; Hosseinpour, Saman; Johnson, Magnus; Leygraf, Christofer; Jonsson, Mats

    2013-11-01

    The long term safety of repositories for radioactive waste is one of the main concerns for countries utilizing nuclear power. The integrity of engineered and natural barriers in such repositories must be carefully evaluated in order to minimize the release of radionuclides to the biosphere. One of the most developed concepts of long term storage of spent nuclear fuel is the Swedish KBS-3 method. According to this method, the spent fuel will be sealed inside copper canisters surrounded by bentonite clay and placed 500 m down in stable bedrock. Despite the importance of the process of radiation induced corrosion of copper, relatively few studies have been reported. In this work the effect of the total gamma dose on radiation induced corrosion of copper in anoxic pure water has been studied experimentally. Copper samples submerged in water were exposed to a series of total doses using three different dose rates. Unirradiated samples were used as reference samples throughout. The copper surfaces were examined qualitatively using IRAS and XPS and quantitatively using cathodic reduction. The concentration of copper in solution after irradiation was measured using ICP-AES. The influence of aqueous radiation chemistry on the corrosion process was evaluated based on numerical simulations. The experiments show that the dissolution as well as the oxide layer thickness increase upon radiation. Interestingly, the evaluation using numerical simulations indicates that aqueous radiation chemistry is not the only process driving the corrosion of copper in these systems.

  20. Multilayer (TiN, TiAlN) ceramic coatings for nuclear fuel cladding

    Science.gov (United States)

    Alat, Ece; Motta, Arthur T.; Comstock, Robert J.; Partezana, Jonna M.; Wolfe, Douglas E.

    2016-09-01

    In an attempt to develop an accident-tolerant fuel (ATF) that can delay the deleterious consequences of loss-of-coolant-accidents (LOCA), multilayer coatings were deposited onto ZIRLO® coupon substrates by cathodic arc physical vapor deposition (CA-PVD). Coatings were composed of alternating TiN (top) and Ti1-xAlxN (2-layer, 4-layer, 8-layer and 16-layer) layers. The minimum TiN top coating thickness and coating architecture were optimized for good corrosion and oxidation resistance. Corrosion tests were performed in static pure water at 360 °C and 18.7 MPa for up to 90 days. The optimized coatings showed no spallation/delamination and had a maximum of 6 mg/dm2 weight gain, which is 6 times smaller than that of a control sample of uncoated ZIRLO® which showed a weight gain of 40.2 mg/dm2. The optimized architecture features a ∼1 μm TiN top layer to prevent boehmite phase formation during corrosion and a TiN/TiAlN 8-layer architecture which provides the best corrosion performance.

  1. Copper anode corrosion affects power generation in microbial fuel cells

    KAUST Repository

    Zhu, Xiuping

    2013-07-16

    Non-corrosive, carbon-based materials are usually used as anodes in microbial fuel cells (MFCs). In some cases, however, metals have been used that can corrode (e.g. copper) or that are corrosion resistant (e.g. stainless steel, SS). Corrosion could increase current through galvanic (abiotic) current production or by increasing exposed surface area, or decrease current due to generation of toxic products from corrosion. In order to directly examine the effects of using corrodible metal anodes, MFCs with Cu were compared with reactors using SS and carbon cloth anodes. MFCs with Cu anodes initially showed high current generation similar to abiotic controls, but subsequently they produced little power (2 mW m-2). Higher power was produced with microbes using SS (12 mW m-2) or carbon cloth (880 mW m-2) anodes, with no power generated by abiotic controls. These results demonstrate that copper is an unsuitable anode material, due to corrosion and likely copper toxicity to microorganisms. © 2013 Society of Chemical Industry.

  2. SPENT FUEL MANAGEMENT AT THE SAVANNAH RIVER SITE

    Energy Technology Data Exchange (ETDEWEB)

    Vormelker, P; Robert Sindelar, R; Richard Deible, R

    2007-11-03

    Spent nuclear fuels are received from reactor sites around the world and are being stored in the L-Basin at the Savannah River Site (SRS) in Aiken, South Carolina. The predominant fuel types are research reactor fuel with aluminum-alloy cladding and aluminum-based fuel. Other fuel materials include stainless steel and Zircaloy cladding with uranium oxide fuel. Chemistry control and corrosion surveillance programs have been established and upgraded since the early 1990's to minimize corrosion degradation of the aluminum cladding materials, so as to maintain fuel integrity and minimize personnel exposure from radioactivity in the basin water. Recent activities have been initiated to support additional decades of wet storage which include fuel inspection and corrosion testing to evaluate the effects of specific water impurity species on corrosion attack.

  3. Roles of Radiolytic and Externally Generated H2 in the Corrosion of Fractured Spent Nuclear Fuel.

    Science.gov (United States)

    Liu, Nazhen; Wu, Linda; Qin, Zack; Shoesmith, David W

    2016-11-15

    A 2-D model for the corrosion of spent nuclear fuel inside a failed nuclear waste container has been modified to determine the influence of various redox processes occurring within fractures in the fuel. The corrosion process is driven by reaction of the fuel with the dominant α radiolysis product, H2O2. A number of reactions are shown to moderate or suppress the corrosion rate, including H2O2 decomposition and a number of reactions involving dissolved H2 produced either by α radiolysis or by the corrosion of the steel container vessel. Both sources of H2 lead to the suppression of fuel corrosion, with their relative importance being determined by the radiation dose rate, the steel corrosion rate, and the dimensions of the fractures in the fuel. The combination of H2 from these two sources can effectively prevent corrosion when only micromolar quantities of H2 are present.

  4. Alkaline corrosion properties of laser-clad aluminum/titanium coatings

    DEFF Research Database (Denmark)

    Aggerbeck, Martin; Herbreteau, Alexis; Rombouts, Marleen

    2015-01-01

    with supersaturated titanium ( (1 weight per cent), Al3Ti intermetallics and large partially undissolved Ti6Al4V particles. Heat treatment lowered the titanium concentration in the aluminum matrix, changed the shape of the Al3Ti precipitates and increased the degree of dissolution of the Ti6Al4V particles. Corrosion...... testing showed significant localized dissolution of the aluminum matrix. Research limitations/implications – Increased titanium concentration and heat treatment gave improved alkaline corrosion properties. At pH 13.5, the Al3Ti phases were protected, while the aluminum matrix corroded. Practical...... implications – For alkaline corrosion-protection of aluminum in the automobile industry, titanium might be useful at pH values below 13.5 or by using other coating techniques. Originality/value – This is the first study testing the use of titanium as a protective element of aluminum in stringent alkaline...

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

    Energy Technology Data Exchange (ETDEWEB)

    Gray Chang

    2012-03-01

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

  6. Advanced Steels for Accident Tolerant Fuel Cladding in Current Light Water Reactors

    Science.gov (United States)

    Rebak, Raul B.

    After the March 2011 Fukushima events, the U.S. Congress directed the Department of Energy (DOE) to focus efforts on the development of fuel cladding materials with enhanced accident tolerance. In comparison with the stand-ard UO2-Zirconium based system, the new fuels need to tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operation conditions. Advanced steels such as iron-chromium-aluminum (FeCrAl) alloys are being investigated for degradation behavior both under normal operation conditions in high temperature water (e.g. 288°C) and under accident conditions for reaction with steam up to 1400°C. Commercial and experimental alloys were tested for several periods of time in 100% superheated steam from 800°C to 1475°C. Results show that FeCrAl alloys significantly outperform the resistance in steam of the current zirconium alloys.

  7. Formation of intermetallic compound at interface between rare earth elements and ferritic-martensitic steel by fuel cladding chemical interaction

    Institute of Scientific and Technical Information of China (English)

    Jun Hwan Kim; Byoung Oon Lee; Chan Bock Lee; Seung Hyun Jee; Young Soo Yoon

    2012-01-01

    The intermetallic compounds formation at interface between rare earth elements and clad material were investigated to demonstrate the effects of rare earth elements on fuel-cladding chemical interaction (FCCI) behavior.Mischmetal (70Ce-30La) and Nd were prepared as rare earth elements.Diffusion couple testing was performed on the rare earth elements and cladding (9Cr2W steel) near the operation temperature of(sodium-cooled fast reactor) SFR fuel.The performance of a diffusion barrier consisting of Zr and V metallic foil against the rare earth elements was also evaluated.Our results showed that Ce and Nd in the rare earth elements and Fe in the clad material interdiffused and reacted to form intermetallic species according to the parabolic rate law,describing the migration of the rare earth element.The diffusion of Fe limited the reaction progress such that the entire process was governed by the cubic rate law.Rare earth materials could be used as a surrogate for high burnup metallic fuels,and the performance of the barrier material was demonstrated to be effective.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-28

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

  9. Corrosion testing of candidates for the alkaline fuel cell cathode

    Science.gov (United States)

    Singer, Joseph; Fielder, William L.

    1989-01-01

    Current/voltage data was obtained for specially made corrosion electrodes of some oxides and of gold materials for the purpose of developing a screening test of catalysts and supports for use at the cathode of the alkaline fuel cell. The data consists of measurements of current at fixed potentials and cyclic voltammograms. These data will have to be correlated with longtime performance data in order to fully evaluate this approach to corrosion screening. Corrosion test screening of candidates for the oxygen reduction electrode of the alkaline fuel cell was applied to two substances, the pyrochlore Pb2Ru2O6.5 and the spinel NiCo2O4. The substrate gold screen and a sample of the IFC Orbiter Pt-Au performance electrode were included as blanks. The pyrochlore data indicate relative stability, although nothing yet can be said about long term stability. The spinel was plainly unstable. For this type of testing to be validated, comparisons will have to be made with long term performance tests.

  10. Development of ODS FeCrAl alloys for accident-tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoelzer, David T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Unocic, Kinga A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-18

    FeCrAl alloys are prime candidates for accident-tolerant fuel cladding due to their excellent oxidation resistance up to 1400 C and good mechanical properties at intermediate temperature. Former commercial oxide dispersion strengthened (ODS) FeCrAl alloys such as PM2000 exhibit significantly better tensile strength than wrought FeCrAl alloys, which would alloy for the fabrication of a very thin (~250 m) ODS FeCrAl cladding and limit the neutronic penalty from the replacement of Zr-based alloys by Fe-based alloys. Several Fe-12-Cr-5Al ODS alloys where therefore fabricated by ball milling FeCrAl powders with Y2O3 and additional oxides such as TiO2 or ZrO2. The new Fe-12Cr-5Al ODS alloys showed excellent tensile strength up to 800 C but limited ductility. Good oxidation resistance in steam at 1200 and 1400 C was observed except for one ODS FeCrAl alloy containing Ti. Rolling trials were conducted at 300, 600 C and 800 C to simulate the fabrication of thin tube cladding and a plate thickness of ~0.6mm was reached before the formation of multiple edge cracks. Hardness measurements at different stages of the rolling process, before and after annealing for 1h at 1000 C, showed that a thinner plate thickness could likely be achieved by using a multi-step approach combining warm rolling and high temperature annealing. Finally, new Fe-10-12Cr-5.5-6Al-Z gas atomized powders have been purchased to fabricate the second generation of low-Cr ODS FeCrAl alloys. The main goals are to assess the effect of O, C, N and Zr contents on the ODS FeCrAl microstructure and mechanical properties, and to optimize the fabrication process to improve the ductility of the 2nd gen ODS FeCrAl while maintaining good mechanical strength and oxidation resistance.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-31

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

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

    Directory of Open Access Journals (Sweden)

    June-Hyung Kim

    2016-11-01

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

  13. Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors

    Science.gov (United States)

    Rebak, Raul B.

    2015-12-01

    As a consequence of the March 2011 events at the Fukushima site, the U.S. congress asked the Department of Energy (DOE) to concentrate efforts on the development of nuclear fuels with enhanced accident tolerance. The new fuels had to maintain or improve the performance of current UO2-zirconium alloy rods during normal operation conditions and tolerate the loss of active cooling in the core for a considerably longer time period than the current system. DOE is funding cost-shared research to investigate the behavior of advanced steels both under normal operation conditions in high-temperature water [ e.g., 561 K (288 °C)] and under accident conditions for reaction with superheated steam. Current results show that, under accident conditions, the advanced ferritic steels (1) have orders of magnitude lower reactivity with steam, (2) would generate less hydrogen and heat than the current zirconium alloys, (3) are resistant to stress corrosion cracking under normal operation conditions, and (4) have low general corrosion in water at 561 K (288 °C).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-25

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-23

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

  17. Corrosion-resistant, electrically-conductive plate for use in a fuel cell stack

    Science.gov (United States)

    Carter, J David [Bolingbrook, IL; Mawdsley, Jennifer R [Woodridge, IL; Niyogi, Suhas [Woodridge, IL; Wang, Xiaoping [Naperville, IL; Cruse, Terry [Lisle, IL; Santos, Lilia [Lombard, IL

    2010-04-20

    A corrosion resistant, electrically-conductive, durable plate at least partially coated with an anchor coating and a corrosion resistant coating. The corrosion resistant coating made of at least a polymer and a plurality of corrosion resistant particles each having a surface area between about 1-20 m.sup.2/g and a diameter less than about 10 microns. Preferably, the plate is used as a bipolar plate in a proton exchange membrane (PEMFC) fuel cell stack.

  18. The Study Programm Report of the Corrosion Behavior of New Zirconium-based Alloys

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The corrosion of fuel cladding in PWR limits the extension of burnup. To compare the corrosion resistance of Zr-4 and new zirconium-based alloys, the out-of-pile water-side corrosion test has been conducted for these materials.To study the effects of heat flux on the corrosion of cladding, and keep the surface of cladding as an original ’as-received’ statement, the heat elements are introduced into the inside of the cladding tubes.The materials have been exposed for 205 d till now. The oxide film performed on the surface of cladding is black and glossy. The thickness of oxide is measured by the method of eddy current.

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

    Science.gov (United States)

    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.

  20. Investigation of silver and iodine transport through silicon carbide layers prepared for nuclear fuel element cladding

    Science.gov (United States)

    Friedland, E.; van der Berg, N. G.; Malherbe, J. B.; Hancke, J. J.; Barry, J.; Wendler, E.; Wesch, W.

    2011-03-01

    Transport of silver and iodine through polycrystalline SiC layers produced by PBMR (Pty) Ltd. for cladding of TRISO fuel kernels was investigated using Rutherford backscattering analysis and electron microscopy. Fluences of 2 × 10 16 Ag + cm -2 and 1 × 10 16 I + cm -2 were implanted at room temperature, 350 °C and 600 °C with an energy of 360 keV, producing an atomic density of approximately 1.5% at the projected ranges of about 100 nm. The broadening of the implantation profiles and the loss of diffusors through the front surface during vacuum annealing at temperatures up to 1400 °C was determined. The results for room temperature implantations point to completely different transport mechanisms for silver and iodine in highly disordered silicon carbide. However, similar results are obtained for high temperature implantations, although iodine transport is much stronger influenced by lattice defects than is the case for silver. For both diffusors transport in well annealed samples can be described by Fickian grain boundary diffusion with no abnormal loss through the surface as would be expected from the presence of nano-pores and/or micro-cracks. At 1100 °C diffusion coefficients for silver and iodine are below our detection limit of 10 -21 m 2 s -1, while they increase into the 10 -20 m 2 s -1 range at 1300 °C.

  1. Revisiting the method to obtain the mechanical properties of hydrided fuel cladding in the hoop direction

    Energy Technology Data Exchange (ETDEWEB)

    Martin-Rengel, M.A., E-mail: mamartin@mater.upm.es [Departamento de Ciencia de Materiales, UPM, ETSI Caminos, Canales y Puertos, Profesor Aranguren s/n, E-28040 Madrid (Spain); Gomez Sanchez, F.J., E-mail: javier.gomez@amsimulation.com [Advanced Material Simulation, S.L (Spain); Ruiz-Hervias, J.; Caballero, L.; Valiente, A. [Departamento de Ciencia de Materiales, UPM, ETSI Caminos, Canales y Puertos, Profesor Aranguren s/n, E-28040 Madrid (Spain)

    2012-10-15

    The method reported in the literature to calculate the stress-strain curve of nuclear fuel cladding from ring tensile test is revisited in this paper and a new alternative is presented. In the former method, two universal curves are introduced under the assumption of small strain. In this paper it is shown that these curves are not universal, but material-dependent if geometric nonlinearity is taken into account. The new method is valid beyond small strains, takes geometric nonlinearity into consideration and does not need universal curves. The stress-strain curves in the hoop direction are determined by combining numerical calculations with experimental results in a convergent loop. To this end, ring tensile tests were performed in unirradiated hydrogen-charged samples. The agreement among the simulations and the experimental results is excellent for the range of concentrations tested (up to 2000 wppm hydrogen). The calculated stress-strain curves show that the mechanical properties do not depend strongly on the hydrogen concentration, and that no noticeable strain hardening occurs. However, ductility decreases with the hydrogen concentration, especially beyond 500 wppm hydrogen. The fractographic results indicate that as-received samples fail in a ductile fashion, whereas quasicleavage is observed in the hydrogen-charged samples.

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

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, R.H. (comp.)

    1976-06-14

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-31

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

  4. Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

    Science.gov (United States)

    Wang, Hong; Wang, Jy-An John

    2016-10-01

    Behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending was studied. Tests were performed under load or moment control at 5 Hz. The surrogate rods fractured under moment amplitudes greater than 10.16 Nm with fatigue lives between 2.4 × 103 and 2.2 × 106 cycles. Fatigue response of Zry-4 cladding was characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition affect surrogate rod failure. Both debonding of PPI/PCI and pellet fracturing contribute to surrogate rod bending fatigue. The effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective gauge length is effective in sensor spacing correction. The database developed and the understanding gained in this study can serve as input to analysis of SNF (spent nuclear fuel) vibration integrity.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-01

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

  6. In-situ Observation of Boiling Dynamics on Fuel Cladding Surface in Non-pressurized Water Using Acoustic Emission Method

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    In the PWR primary coolant system, a phenomenon of axial offset anomaly (AOA) can be caused due to accumulated boron hide out in porous CRUD deposition on the fuel cladding surface. Up to now, the CRUD deposition has been well known to be driven by subcooled nucleate boiling (SNB) on the cladding surface based on large scale experimental work. Therefore, monitoring and evaluation of the SNB-phenomenon is an important approach to study the CRUD deposition. Many attempts have been made to study the SNB and CRUD deposition using thermal hydraulic or model calculation. However, a comprehensive understanding of the SNB during CRUD deposition is still far from being realized. Acoustic emission (AE) technique, as an in-situ nondestructive evaluation (NDE) method, has been widely used to monitor the boiling activity in containers and pipes. Accordingly, this work aimed to investigate the exact AE characteristics of SNB-phenomenon on the fuel cladding surface at atmospheric pressure, with the purpose of providing an experimental groundwork for the AE investigation on SNB in high-temperature pressurized coolant system. In this study, we conducted an in-situ experimental observation of the bubble dynamic of SNB in non-pressurized water at atmospheric pressure using AE method. The AE of heater noise was confirmed to cluster between 8 and 26 khz. Three AE groups were detected during the boiling process in the Snob zones. AE group 1 and 3 seemed to be the results of bubble growth and collapse, while bubble departure from the cladding surface was reasonably associated with an isolated AE group 2.

  7. Potential high temperature corrosion problems due to co-firing of biomass and fossil fuels

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Vilhelmsen, T.; Jensen, S.A.

    2007-01-01

    , the internal sulphidation is much more significant than that revealed in the demonstration project. Avedøre 2 main boiler is fuelled with wood pellets + heavy fuel oil + gas. Some reaction products due to the presence of vanadium compounds in the heavy oil were detected, i.e. iron vanadates. However, the most...... significant corrosion attack was due to sulphidation attack at the grain boundaries of 18-8 steel after 3 years exposure. The corrosion mechanisms and corrosion rates are compared with biomass firing and coal firing. Potential corrosion problems due to co-firing biomass and fossil fuels are discussed....

  8. Potential high temperature corrosion problems due to co-firing of biomass and fossil fuels

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Vilhelmsen, T.; Jensen, S.A.

    2008-01-01

    in this environment, the internal sulphidation is much more significant than that revealed in the demonstration project. Avedøre 2 main boiler is fuelled with wood pelletsþheavy fuel oilþgas. Some reaction products resulting from the presence of vanadium compounds in the heavy oil were detected, i.e. iron vanadates....... However, the most significant corrosion attack was sulphidation attack at the grain boundaries of 18-8 steel after 3 years exposure. The corrosion mechanisms and corrosion rates are compared with biomass firing and coal firing. Potential corrosion problems due to co-firing biomass and fossil fuels...

  9. Modeling of Zircaloy cladding degradation under repository conditions

    Energy Technology Data Exchange (ETDEWEB)

    Santanam, L.; Raghavan, S.; Chin, B.A. [Auburn Univ., AL (USA). Dept. of Materials Engineering; Shaw, H. [Lawrence Livermore National Lab., CA (USA)

    1989-07-01

    Two potential degradation mechanisms, creep and stress corrosion cracking, of Zircaloy cladding during repository storage of spent nuclear fuel have 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. A stress analysis of fuel rods has been performed. Stresses in the outer zirconium oxide layer and the inner Zircaloy tube have been predicted for typical internal pressurization, oxide layer thickness, volume expansion from formation of the oxide layer and thermal expansion coefficients of the cladding and oxide. Stress relaxation occurring in-reactor has also been taken into account. The calculations indicate that for the anticipated storage conditions investigated, the outer zirconium oxide layer is in a state of compression thus making it unlikely that stress corrosion cracking of the exterior surface will occur. 20 refs., 6 figs., 9 tabs.

  10. Corrosion of copper containers prior to saturation of a nuclear fuel waste disposal vault

    Energy Technology Data Exchange (ETDEWEB)

    King, F.; Kolar, M

    1997-12-01

    The buffer material surrounding the containers in a Canadian nuclear fuel waste disposal vault will partially desiccate as a result of the elevated temperature at the container surface. This will lead to a period of corrosion in a moist air atmosphere. Corrosion will either take the form of slow oxidation if the container surface remains dry or aqueous electrochemical corrosion if the surface is wetted by a thin liquid film. The relevant literature is reviewed, from which it is concluded that corrosion should be uniform in nature, except if the surface is wetted, in which case localized corrosion is a possibility. A quantitative analysis of the extent and rate of uniform corrosion during the unsaturated period is presented. Two bounding cases are considered: first, the case of slow oxidation in moist air following either logarithmic or parabolic oxide-growth kinetics and, second, the case of electrochemically based corrosion occurring in a thin liquid film uninhibited by the growth of corrosion products. (author)

  11. Corrosion studies of carbon steel under impinging jets of simulated slurries of neutralized current acid waste (NCAW) and neutralized cladding removal waste (NCRW)

    Energy Technology Data Exchange (ETDEWEB)

    Smith, H.D.; Elmore, M.R.

    1992-01-01

    Plans for the disposal of radioactive liquid and solid wastes presently stored in double-shell tanks at the Hanford Site call for retrieval and processing of the waste to create forms suitable for permanent disposal. Waste will be retrieved from a tank using a submerged slurry pump in conjunction with one or more rotating slurry jet mixer pumps. Pacific Northwest Laboratory (PNL) has conducted tests using simulated waste slurries to assess the effects of a impinging slurry jet on the corrosion rate of the tank wall and floor, an action that could potentially compromise the tank's structural integrity. Corrosion processes were investigated on a laboratory scale with a simulated neutralized cladding removal waste (NCRW) slurry and in a subsequent test with simulated neutralized current acid waste (NCAW) slurry. The test slurries simulated the actual NCRW and NCAW both chemically and physically. The tests simulated those conditions expected to exist in the respective double-shell tanks during waste retrieval operations. Results of both tests indicate that, because of the action of the mixer pump slurry jets, the waste retrieval operations proposed for NCAW and NCRW will moderately accelerate corrosion of the tank wall and floor. Based on the corrosion of initially unoxidized test specimens, and the removal of corrosion products from those specimens, the maximum time-averaged corrosion rates of carbon steel in both waste simulants for the length of the test was {approximately}4 mil/yr. The protective oxide layer that exists in each storage tank is expected to inhibit corrosion of the carbon steel.

  12. Corrosion studies of carbon steel under impinging jets of simulated slurries of neutralized current acid waste (NCAW) and neutralized cladding removal waste (NCRW)

    Energy Technology Data Exchange (ETDEWEB)

    Smith, H.D.; Elmore, M.R.

    1992-01-01

    Plans for the disposal of radioactive liquid and solid wastes presently stored in double-shell tanks at the Hanford Site call for retrieval and processing of the waste to create forms suitable for permanent disposal. Waste will be retrieved from a tank using a submerged slurry pump in conjunction with one or more rotating slurry jet mixer pumps. Pacific Northwest Laboratory (PNL) has conducted tests using simulated waste slurries to assess the effects of a impinging slurry jet on the corrosion rate of the tank wall and floor, an action that could potentially compromise the tank`s structural integrity. Corrosion processes were investigated on a laboratory scale with a simulated neutralized cladding removal waste (NCRW) slurry and in a subsequent test with simulated neutralized current acid waste (NCAW) slurry. The test slurries simulated the actual NCRW and NCAW both chemically and physically. The tests simulated those conditions expected to exist in the respective double-shell tanks during waste retrieval operations. Results of both tests indicate that, because of the action of the mixer pump slurry jets, the waste retrieval operations proposed for NCAW and NCRW will moderately accelerate corrosion of the tank wall and floor. Based on the corrosion of initially unoxidized test specimens, and the removal of corrosion products from those specimens, the maximum time-averaged corrosion rates of carbon steel in both waste simulants for the length of the test was {approximately}4 mil/yr. The protective oxide layer that exists in each storage tank is expected to inhibit corrosion of the carbon steel.

  13. Boiling performance and material robustness of modified surfaces with multi scale structures for fuel cladding development

    Energy Technology Data Exchange (ETDEWEB)

    Jo, HangJin; Kim, Jin Man [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Yeom, Hwasung [Department of Nuclear Engineering and Engineering physics, UW-Madison, Madison, WI 53706, Unities States (United States); Lee, Gi Cheol [Department of Mechanical Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Kiyofumi, Moriyama; Kim, Moo Hwan [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Sridharan, Kumar; Corradini, Michael [Department of Nuclear Engineering and Engineering physics, UW-Madison, Madison, WI 53706, Unities States (United States)

    2015-09-15

    Highlights: • We improved boiling performance and material robustness using surface modification. • We combined micro/millimeter post structures and nanoparticles with heat treatments. • Compactly-arranged micrometer posts had improved boiling performance. • CHF increased significantly due to capillary pumping by the deposited NP layers. • Sintering procedure increased mechanical strength of the NP coating surface. - Abstract: By regulating the geometrical characteristics of multi-scale structures and by adopting heat treatment for protective layer of nanoparticles (NPs), we improved critical heat flux (CHF), boiling heat transfer (BHT), and mechanical robustness of the modified surface. We fabricated 1-mm and 100-μm post structures and deposited NPs on the structured surface as a nano-scale structured layer and protective layer at the same time, then evaluated the CHF and BHT and material robustness of the modified surfaces. On the structured surfaces without NPs, the surface with compactly-arranged micrometer posts had improved CHF (118%) and BHT (41%). On the surface with structures on which NPs had been deposited, CHF increased significantly (172%) due to capillary pumping by the deposited NP layers. The heat treatment improved robustness of coating layer in comparison to the one of before heat treatment. In particular, low-temperature sintering increased the hardness of the modified surface by 140%. The increased mechanical strength of the NP coating is attributed to reduction in coating porosity during sintering. The combination of micrometer posts structures and sintered NP coating can increase the safety, efficiency and reliability of advanced nuclear fuel cladding.

  14. Irradiation testing of internally pressurized and/or graphite coated Zircaloy-4 clad fuel rods in the NRX Reactor (AWBA Development Program). [LWBR

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, R.C.; Sherman, J.

    1978-11-01

    Irradiation tests on 0.612 inch O.D. by 117-inch long Zircaloy-4 clad fuel rods were performed to assess the effects on fuel rod performance of (1) internal helium pre-pressurization to 500 psi as fabricated, (2) the presence of a graphite barrier coating on the inside cladding surface, and (3) combined pre-pressurization and graphite coating. Periodic dimensional examinations were performed on the test rods, and the results were compared with data obtained from two previously irradiated test rods--both unpressurized and uncoated and one intentionally defected. These comparisons indicate that both pre-pressurization and graphite coating can substantially improve fuel element performance capability.

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

    Science.gov (United States)

    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.

  16. Evaluation of galvanic corrosion of a Zn alloy in alcohol fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ambrozin, Alessandra Regina Pepe; Monteiro, Marcos Roberto; Santos, Andre Oliveira; Kuri, Sebastiao Elias

    2010-11-15

    Galvanic corrosion of zamak was evaluated in alcohol fuel and in some alcoholic solutions that contained ionic impurities. Also, the effect of corrosive process on the quality parameters of ethanol was investigated. The results showed that corrosion of zamak mainly occurred in solutions with high levels of water and impurities. After the assays, the increasing of both pH and conductivity of the alcohols was observed. Therefore, the results showed that the contact between zamak and some materials must be avoided and the quality control of alcohol fuel must be assured as a way of avoiding damages on engines and storage-transportation fuels systems. (author)

  17. Corrosion Behaviour of Carbon Steel in Biodiesel–Diesel–Ethanol (BDE Fuel Blend

    Directory of Open Access Journals (Sweden)

    Thangavelu Saravana Kannan

    2015-01-01

    Full Text Available The biodiesel–diesel–ethanol blend represents an important alternative fuel for diesel engines; however, changes in the fuel composition and the introduction of new alternative fuel often results in corrosion and degradation of the automobile fuel system parts. In this present study, the corrosion behavior of carbon steel in B20D70E10 (biodiesel 20%, diesel 70% and ethanol 10% fuel blend was studied by static immersion at room temperature and 60 °C. The effect of B20D70E10 fuel blend on corrosion rate, morphology of corrosion products, and chemical structure of carbon steel were studied. In addition, the change of fuel properties, namely, total acid number, density, viscosity, calorific value, flash point, and color changes were also investigated. Moreover, fuel compositional changes, such as water content and oxidation product level in the fuel blends were examined. The results showed that the degradation of fuel properties and corrosion rate of carbon steel in B20D70E10 are lower than neat biodiesel (B100, whereas slightly higher than petro-diesel (B0

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

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Laboratory

    2016-09-01

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

  19. Evaluation of the interfacial shear strength and residual stress of TiAlN coating on ZIRLO™ fuel cladding using a modified shear-lag model approach

    Science.gov (United States)

    Liu, Y.; Bhamji, I.; Withers, P. J.; Wolfe, D. E.; Motta, A. T.; Preuss, M.

    2015-11-01

    This paper investigates the residual stresses and interfacial shear strength of a TiAlN coating on Zr-Nb-Sn-Fe alloy (ZIRLO™) substrate designed to improve corrosion resistance of fuel cladding used in water-cooled nuclear reactors, both during normal and exceptional conditions, e.g. a loss of coolant event (LOCA). The distribution and maximum value of the interfacial shear strength has been estimated using a modified shear-lag model. The parameters critical to this analysis were determined experimentally. From these input parameters the interfacial shear strength between the TiAlN coating and ZIRLO™ substrate was inferred to be around 120 MPa. It is worth noting that the apparent strength of the coating is high (∼3.4 GPa). However, this is predominantly due to the large compressive residuals stress (3 GPa in compression), which must be overcome for the coating to fail in tension, which happens at a load just 150 MPa in excess of this.

  20. Effect of the Boron and Nitrogen on precipitation behavior in modified 9Cr steel for SFR fuel cladding after aging

    Energy Technology Data Exchange (ETDEWEB)

    Jeog, Eun-Hee; Kim, Young Do [Hanyang University, Seoul (Korea, Republic of); Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Ferritic-martensitic steels are being considered as an attractive candidate material for a fuel cladding of a SFR due to their low expansion coefficients, high thermal conductivities and excellent irradiation resistances against void swelling. Because of its superior dimensional stability against fast neutron irradiation, Ferritic-martensitic steel of 9Cr and 12Cr steels are preferable to utilize in the fuel cladding of an SFR in KAERI. The soluble boron reduces the coarsening rate of M{sub 23}C{sub 6} carbides along boundaries near prior austenite grain boundaries during creep, enhancing the boundary and sub-boundary hardening for up to long times. The enhancement of boundary and sub-boundary hardening retards the onset of acceleration creep, which decreases the minimum creep rate and improves the creep life. It has been reported that the excess addition of boron and nitrogen promotes the formation of boron nitrides during normalizing heat treatment, which significantly reduces soluble B and N concentrations and offsets the benefit due to boron and nitrogen. In this study, comparison of the microstructure and mechanical properties on SFR fuel cladding steel with different B and N contents after aging were carried out. The addition of B stabilizes the M{sub 23}C{sub 6}, hence the coarsening of M{sub 23}C{sub 6} was not observed in alloy 1 after 7000 hours aging. The size distribution of an alloy 2 was not largely changed with aging time, and this phenomena would be caused by an addition of nitrogen, by stabilize the nitride precipitates such as MX and M{sub 2}X.

  1. Physical and Numerical Difficulties in Computer Modelling of Pellet-Cladding Contact Problems for Burned-Up Fuel

    Directory of Open Access Journals (Sweden)

    M. Dostál

    2005-01-01

    Full Text Available The importance of fuel reliability is growing due to the deregulated electricity market and the demands on operability and availability to the electricity grid of nuclear units. Under these conditions of fuel exploitation, the problems of PCMI (Pellet-Cladding Mechanical Interaction are very important from the point of view of fuel rod integrity and reliability. Severe loading is thermophysically and mechanically expressed as a greater probability of cladding failure especially during power maneuvering. We have to be able to make a realistic prediction of safety margins, which is very difficult by using computer simulation methods. NRI (Nuclear Research Institute has recently been engaged in developing 2D and 3D FEM (Finite Element Method based models dealing with this problem. The latest effort in this field has been to validate 2D r-z models developed in the COSMOS/M system against calculations using the FEMAXI-V code. This paper presents a preliminary comparison between classical FEM based integral code calculations and new models that are still under development. The problem has not been definitely solved. The presented data is of a preliminary nature, and several difficult problems remain to be solved. 

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

    Science.gov (United States)

    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.

  3. Development of laser welded appendages to Zircaloy-4 fuel tubing (sheath/cladding)

    Energy Technology Data Exchange (ETDEWEB)

    Livingstone, S., E-mail: steve.livingstone@cnl.ca [Canadian Nuclear Laboratories Limited, Chalk River, ON, Canada K0J 1J0 (Canada); Xiao, L. [Canadian Nuclear Laboratories Limited, Chalk River, ON, Canada K0J 1J0 (Canada); Corcoran, E.C.; Ferrier, G.A.; Potter, K.N. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, ON, Canada K7K 7B4 (Canada)

    2015-04-01

    Highlights: • Examines feasibility of laser welding appendages to Zr-4 tubing. • Laser welding minimizes the HAZ and removes toxic Be. • Mechanical properties of laser welds appear competitive with induction brazed joints. • Work appears promising and lays the foundation for further investigations. - Abstract: Laser welding is a potential alternative to the induction brazing process commonly used for appendage attachment in CANDU{sup ®} fuel fabrication that uses toxic Be as a filler metal, and creates multiple large heat affected zones in the sheath. For this work, several appendages were laser welded to tubing using different laser heat input settings and then examined with a variety of techniques: visual examination, metallography, shear strength testing, impact testing, and fracture surface analysis. Where possible, the examination results are contrasted against production induction brazed joints. The work to date looks promising for laser welded appendages. Further work on joint optimization, corrosion testing, irradiation testing, and post-irradiation examination will be performed in the future.

  4. Improvement of carbon corrosion resistance through heat-treatment in polymer electrolyte membrane fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Y.J.; Oh, H.S.; Kim, H. [Yonsei Univ., Seoul (Korea, Republic of). Dept. of Chemical and Biomolecular Engineering

    2010-07-01

    Electrochemical corrosion of carbon in the catalyst layer of polymer electrolyte membrane fuel cells (PEMFCs) is a critical factor in limiting their durability. The corrosion rate increases during the iterative abnormal operating conditions known as reverse current phenomenon. The corrosion causes a decrease of the active surface of the platinum (Pt) catalyst. The graphitization of carbon increases corrosion resistance, and the hydrophobicity of the carbon surface can also play an important role in decreasing carbon corrosion. This study investigated the effect of heat-treating carbon nanofibers (CNFs) for use in PEMFC applications. The aim of the study was to determine if heat treatments modified the carbon surface by eliminating the oxygen functional group and increasing hydrophobicity. The electrochemical carbon corrosion of CNFs were compared after heat treatments at various temperatures. Mass spectrometry was used to measure electrochemical carbon corrosion by monitoring the amounts of carbon dioxide (CO{sub 2}) produced during the electrochemical oxidation process. 2 refs.

  5. Chemical effects in the Corrosion of Aluminum and Aluminum Alloys. A Bibliography

    Science.gov (United States)

    1976-10-01

    Reduction of corrosion rate of Al cladding on fuel elements in Savannah River reactors by lithium silicate. 1970-30 B.A. Pandya, S.S. Sampat, J.C. Vora...Resistance Of Al Alloys" Statistical study of the corrosion of Al alloys in sea water and in an industrial atm for 5 yrs. 1974-29 A. Soudan Galvano

  6. Advanced ODS FeCrAl alloys for accident-tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N [ORNL; Unocic, Kinga A [ORNL; Hoelzer, David T [ORNL; Pint, Bruce A [ORNL

    2014-09-01

    ODS FeCrAl alloys are being developed with optimum composition and properties for accident tolerant fuel cladding. Two oxide dispersion strengthened (ODS) Fe-15Cr-5Al+Y2O3 alloys were fabricated by ball milling and extrusion of gas atomized metallic powder mixed with Y2O3 powder. To assess the impact of Mo on the alloy mechanical properties, one alloy contained 1%Mo. The hardness and tensile properties of the two alloys were close and higher than the values reported for fine grain PM2000 alloy. This is likely due to the combination of a very fine grain structure and the presence of nano oxide precipitates. The nano oxide dispersion was however not sufficient to prevent grain boundary sliding at 800 C and the creep properties of the alloys were similar or only slightly superior to fine grain PM2000 alloy. Both alloys formed a protective alumina scale at 1200 C in air and steam and the mass gain curves were similar to curves generated with 12Cr-5Al+Y2O3 (+Hf or Zr) ODS alloys fabricated for a different project. To estimate the maximum temperature limit of use for the two alloys in steam, ramp tests at a rate of 5 C/min were carried out in steam. Like other ODS alloys, the two alloys showed a significant increase of the mas gains at T~ 1380 C compared with ~1480 C for wrought alloys of similar composition. The beneficial effect of Yttrium for wrought FeCrAl does not seem effective for most ODS FeCrAl alloys. Characterization of the hardness of annealed specimens revealed that the microstructure of the two alloys was not stable above 1000 C. Concurrent radiation results suggested that Cr levels <15wt% are desirable and the creep and oxidation results from the 12Cr ODS alloys indicate that a lower Cr, high strength ODS alloy with a higher maximum use temperature could be achieved.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

  8. Corrosion of aluminium, stainless steels and AISI 680 nickel alloy in nitrogen-based fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kap, I.; Starostin, M.; Shter, G.E.; Grader, G.S. [Department of Chemical Engineering, Technion-Israel Institute of Technology, Haifa (Israel)

    2012-07-15

    Nitrogen-based compounds can potentially be used as alternative non-carbon or low-carbon fuels. Nevertheless, the corrosion of construction materials at high temperatures and pressures in the presence of such fuel has not been reported yet. This work is focused on the corrosion of AISI Al 6061, 1005 carbon steel (CS), 304, 316L, 310 austenitic stainless steels (SS) and 680 nickel alloy in highly concentrated water solution of ammonium nitrate and urea (ANU). The corrosion at 50 C and ambient pressure and at 350 C and 20 bar was investigated to simulate storage and working conditions. Sodium chloride was added to the fuel (0-5 wt%) to simulate industrial fertilizers and accelerated corrosion environment. Heavy corrosion of CS was observed in ANU solution at 50 C, while Al 6061, 304 and 316L SS showed high resistance both to uniform and pitting corrosion in ANU containing 1% of sodium chloride. Addition of 5% sodium chloride caused pitting of Al 6061 but had no influence on the corrosion of SS. Tests in ANU at 350 C and 20 bar showed pitting on SS 304 and 316L and 680 nickel alloy. The highest corrosion resistance was found for SS 310 due to formation of stable oxide film on its surface. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

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

    Energy Technology Data Exchange (ETDEWEB)

    1989-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1989-08-01

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

  11. Initial Cladding Condition

    Energy Technology Data Exchange (ETDEWEB)

    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

  12. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, Rodney C.

    2003-09-14

    public that there is a reasonable basis for accepting the long-term extrapolations of spent fuel behavior. In recent years ''natural analogues'' for both the repository environment (e.g., the Oklo natural reactors) and nuclear waste form behavior (e.g., corrosion and alteration of uraninite, UO{sub 2+x}) have been cited as a fundamental means of achieving confirmation of long-term extrapolations. In particular, considerable effort has already been made to establish that uraninite, UO{sub 2+x}, with its impurities, is a good structural and chemical analogue for the analysis of the long-term behavior of the UO{sub 2} in spent nuclear fuel. This proposal is based on the study of uraninite and the naturally occurring alteration products of UO{sub 2+x} under oxidizing and reducing conditions. The UO{sub 2} in spent nuclear fuel is not stable under oxidizing conditions. In oxic solutions, uranium has a strong tendency to exist as U{sup 6+} in the uranyl molecule, UO{sub 2}{sup 2+}. Uranyl ions react with a wide variety of inorganic and organic anions to form complexes. Throughout most of the natural range of pH, U{sup 6+} forms strong complexes with oxygen-bearing ions like CO{sub 3}{sup 2-}, HCO{sup 3-}, SO{sub 4}{sup 2-}, PO{sub 4}{sup 3-}, and AsO{sub 4}{sup 3-}, which are present in most oxidized stream and subsurface waters. In arid environments, the U{sup 6+} ion can precipitate as a wide variety of uranyl oxide hydrates, uranyl silicates and uranyl phosphates. This is well demonstrated in experimental work, e.g., in long term drip tests on UO{sub 2} and is confirmed by natural occurrences of UO{sub 2} in which a wide variety of uranyl phases form as alteration products. The most striking feature of these studies is the very close parallel in the paragenetic sequences (i.e. phase formation sequence) between the very long term (10 year tests) and the young (therefore, low-Pb uraninites) of the Nopal I deposit in Mexico.

  13. Optimization of the deposition process of corrosion resistant Stellite 6 coatings produced by laser cladding; Optimizacion del proceso de aporte de recubrimientos anticorrosion de Stellite 6 producidos mediante plaqueado laser

    Energy Technology Data Exchange (ETDEWEB)

    Vicario, I.; Soriano, C.; Sanz, C.; Bayon, R.; Leunda, J.

    2009-07-01

    Laser cladding is one of the most efficient surface treatment technologies in the industry. It uses a laser heat source to deposit a thin layer of a desired material on a moving substrate, whose properties have to be improved, achieving a metallurgical bonding between them with low heat affected zone and low dilution, compared to other conventional technologies such as PTA, TIG welding or thermal Spraying. In this sense, it is remarkable that there are 3 main application fields for laser cladding technology: restoration of refurbishment of damaged parts, surface coating against corrosion or wear, and rapid proto typing. the present work described a study of the optimization of the laser cladding of Co based coatings (Diamalloy 4060NS) on medium carbon steel C45 (AISI 1945). After laser treatment, the surface of the substrate materials is improved in terms of resistance against corrosion; this confirmed in the analysis performed afterwards. it is also shown that the corrosion barrier properties have direct correlation with the laser cladding variables. (Author) 10 refs.

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

    Science.gov (United States)

    Park, Dong Jun; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

    2016-12-01

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

  15. Corrosion

    Science.gov (United States)

    Slabaugh, W. H.

    1974-01-01

    Presents some materials for use in demonstration and experimentation of corrosion processes, including corrosion stimulation and inhibition. Indicates that basic concepts of electrochemistry, crystal structure, and kinetics can be extended to practical chemistry through corrosion explanation. (CC)

  16. Corrosion

    Science.gov (United States)

    Slabaugh, W. H.

    1974-01-01

    Presents some materials for use in demonstration and experimentation of corrosion processes, including corrosion stimulation and inhibition. Indicates that basic concepts of electrochemistry, crystal structure, and kinetics can be extended to practical chemistry through corrosion explanation. (CC)

  17. Metallurgical and Corrosion Properties of Explosively Welded Ti6Al4V/Low Carbon Steel Clad

    Institute of Scientific and Technical Information of China (English)

    Nizamettin Kahraman; Beh(c)et Gülen(c)

    2005-01-01

    Titanium alloy (Ti6Al4V) and Iow carbon steel (LCS) were joined by explosive welding method using different ratios of explosive. Some metallurgical properties of joined samples were investigated. Joined samples were examined by means of optical microscope, scanning electron microscope (SEM) and tensile-shearing tests. Bending, tensile, hardness and corrosion behaviour of the samples were investigated. Separation was not occurred on the joining interface after tensile-shearing and bending tests. It is seen that hardness of both plates were increased with increasing explosive.It is found that increasing explosive ratio leads to an increase in corrosion. It is also found that corrosion rate was high at the beginning of the experiment but the rate of the corrosion decreased subsequently during the experiment.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-06

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

  19. GRH 12-01 Fireside Corrosion in Oxy-fuel Combustion Poster 0108

    Energy Technology Data Exchange (ETDEWEB)

    G. R. Holcomb; J. Tylczak; G. H. Meier; B. Lutz; K. Jung; N. Mu; N. M. Yanar; F. S. Pettit; J. Zhu; A. Wise; D. Laughlin; S. Sridhar

    2012-05-20

    The goals are to: (1) Achieve 90% CO{sub 2} capture at no more than a 35% increase in levelized cost of electricity of post-combustion capture for new and existing conventional coal-fired power plants; (2) Provide high-temperature corrosion information to aid in materials development and selection for oxy-fuel combustion; and (3) Identify corrosion mechanism and behavior differences between air- and oxy-firing.

  20. PEM fuel cell cathode carbon corrosion due to the formation of air/fuel boundary at the anode

    Science.gov (United States)

    Tang, Hao; Qi, Zhigang; Ramani, Manikandan; Elter, John F.

    The impacts of unprotected start up and shut down on fuel cell performance degradation was investigated using both single cell and dual cell configurations. It was found that the air/fuel boundary developed at the anode side after a fuel cell shut down or during its restart caused extremely quick degradation of the cathode. The thickness, the electrochemical active surface area, and the performance of the cathode catalyst layer were significantly reduced. By using a dual cell configuration, cathode potential as high as two times of open circuit voltage was measured, and the corrosion current flowing externally between the two cells was detected and quantified. Carbon catalyst-support corrosion/oxidation at such a high potential was largely responsible for the accelerated fuel cell performance degradation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-01

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

  2. Hot Corrosion of Nickel-Base Alloys in Biomass-Derived Fuel Simulated Atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Leyens, C.; Pint, B.A.; Wright, I.G.

    1999-02-28

    Biomass fuels are considered to be a promising renewable source of energy. However, impurities present in the fuel may cause corrosion problems with the materials used in the hot sections of gas turbines and only limited data are available so far. As part of the Advanced Turbine Systems Program initiated by the U.S. Department of Energy, the present study provides initial data on the hot corrosion resistance of different nickel-base alloys against sodium sulfate-induced corrosion as a baseline, and against salt compositions simulating biomass-derived fuel deposits. Single crystal nickel-superalloy Rene N5, a cast NiCrAlY alloy, a NiCoCrAlY alloy representing industrially used overlay compositions, and a model {beta}NiAl+Hf alloy were tested in 1h thermal cycles at 950 C with different salt coatings deposited onto the surfaces. Whereas the NiCoCrAlY alloy exhibited reasonable resistance against pure sodium sulfate deposits, the NiCrAiY alloy and Rene N5 were attacked severely. Although considered to be an ideal alumina former in air and oxygen at higher temperatures, {beta}NiAl+Hf also suffered from rapid corrosion attack at 950 C when coated with sodium sulfate. The higher level of potassium present in biomass fuels compared with conventional fuels was addressed by testing a NiCoCrAlY alloy coated with salts of different K/Na atomic ratios. Starting at zero Na, the corrosion rate increased considerably when sodium was added to potassium sulfate. In an intermediate region the corrosion rate was initially insensitive to the K/Na ratio but accelerated when very Na-rich compositions were deposited. The key driver for corrosion of the NiCoCrAlY alloy was sodium sulfate rather than potassium sulfate, and no simple additive or synergistic effect of combining sodium and potassium was found.

  3. Microbiologically induced corrosion of aluminum alloys in fuel-oil/aqueous system.

    Science.gov (United States)

    Yang, S S; Lin, J Y; Lin, Y T

    1998-09-01

    To investigate the microbiologically induced corrosion of aluminum alloys in fuel-oil/aqueous system, aluminum alloys A356, AA 5052, AA 5083 and AA 6061 were chosen as the test alloys and Cladosporium and several fuel-oil contaminated microbes isolated in Taiwan were used as test organisms. Aluminum alloy AA 5083 in fuel-oil/aqueous system was the most susceptible material for microbial corrosion, then followed by aluminum alloys AA 5052 and A356, and AA 6061 was more resistant to microbial aggression. Mixed culture had high capability of corrosion, then followed by Penicillium sp. AM-F5, Fusarium sp. AM-F1, Pseudomonas aeruginosa AM-B5, Ps. fluorescens AM-B9, C. resinae ATCC 22712, Penicillium sp. AM-F2, Candida sp. AM-Y1 and Ps. aeruginosa AM-B11. From energy dispersive spectrometer analysis, aluminum and magnesium contents decreased in the corrosion area, while chlorine and sulfur contents increased. The major organic acid produced in fuel-oil/aqueous system was acetic acid, and the total organic acids content had a positive correlation with the degree of microbial corrosion.

  4. Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ji-Han, E-mail: chunjh@kaeri.re.kr; Lim, Sung-Won; Chung, Bub-Dong; Lee, Won-Jae

    2015-08-15

    Highlights: • Thermal conductivity model of the FCM fuel was developed and adopted in the MARS. • Scoping analysis for candidate FCM FAs was performed to select feasible FA. • Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up. • Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. - Abstract: The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates. A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

  5. Corrosion Performance of Fe-Based Alloys in Simulated Oxy-Fuel Environment

    Science.gov (United States)

    Zeng, Zuotao; Natesan, Ken; Cai, Zhonghou; Rink, David L.

    2017-02-01

    The long-term corrosion of Fe-based alloys in simulated oxy-fuel environment at 1023 K (750 °C) was studied. Detailed results are presented on weight change, scale thickness, internal penetration, microstructural characteristics of the corrosion products, and the cracking of scales for the alloys after exposure at 1023 K (750 °C) for up to 3600 hours. An incubation period during which the corrosion rate was low was observed for the alloys. After the incubation period, the corrosion accelerated, and the corrosion process followed linear kinetics. Effects of alloy, CaO-containing ash, and gas composition on the corrosion rate were also studied. In addition, synchrotron nanobeam X-ray analysis was employed to determine the phase and chemical composition of the oxide layers on the alloy surface. Results from these studies are being used to address the long-term corrosion performance of Fe-based alloys in various coal-ash combustion environments and to develop methods to mitigate high-temperature ash corrosion.

  6. Corrosion Performance of Fe-Based Alloys in Simulated Oxy-Fuel Environment

    Science.gov (United States)

    Zeng, Zuotao; Natesan, Ken; Cai, Zhonghou; Rink, David L.

    2016-09-01

    The long-term corrosion of Fe-based alloys in simulated oxy-fuel environment at 1023 K (750 °C) was studied. Detailed results are presented on weight change, scale thickness, internal penetration, microstructural characteristics of the corrosion products, and the cracking of scales for the alloys after exposure at 1023 K (750 °C) for up to 3600 hours. An incubation period during which the corrosion rate was low was observed for the alloys. After the incubation period, the corrosion accelerated, and the corrosion process followed linear kinetics. Effects of alloy, CaO-containing ash, and gas composition on the corrosion rate were also studied. In addition, synchrotron nanobeam X-ray analysis was employed to determine the phase and chemical composition of the oxide layers on the alloy surface. Results from these studies are being used to address the long-term corrosion performance of Fe-based alloys in various coal-ash combustion environments and to develop methods to mitigate high-temperature ash corrosion.

  7. Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

    Science.gov (United States)

    Balooch, Mehdi; Olander, Donald R.; Terrani, Kurt A.; Hosemann, Peter; Casella, Andrew M.; Senor, David J.; Buck, Edgar C.

    2017-04-01

    A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.

  8. High energy X-ray diffraction measurement of residual stresses in a monolithic aluminum clad uranium-10 wt% molybdenum fuel plate assembly

    Science.gov (United States)

    Brown, D. W.; Okuniewski, M. A.; Almer, J. D.; Balogh, L.; Clausen, B.; Okasinski, J. S.; Rabin, B. H.

    2013-10-01

    Residual stresses are expected in monolithic, aluminum clad uranium 10 wt% molybdenum (U-10Mo) nuclear fuel plates because of the large mismatch in thermal expansion between the two bonded materials. The full residual stress tensor of the U-10Mo foil in a fuel plate assembly was mapped with 0.1 mm resolution using high-energy (86 keV) X-ray diffraction. The in-plane stresses in the U-10Mo foil are strongly compressive, roughly -250 MPa in the longitudinal direction and -140 MPa in the transverse direction near the center of the fuel foil. The normal component of the stress is weakly compressive near the center of the foil and tensile near the corner. The disparity in the residual stress between the two in-plane directions far from the edges and the tensile normal stress suggest that plastic deformation in the aluminum cladding during fabrication by hot isostatic pressing also contributes to the residual stress field. A tensile in-plane residual stress is presumed to be present in the aluminum cladding to balance the large in-plane compressive stresses in the U-10Mo fuel foil, but cannot be directly measured with the current technique due to large grain size.

  9. Corrosion mechanisms of candidate structural materials for supercritical water-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Lefu ZHANG; Fawen ZHU; Rui TANG

    2009-01-01

    Nickel-based alloys, austenitic stainless steel, ferritic/martensitic heat-resistant steels, and oxide dispersion strengthened steel are presently considered to be the candidate structural or fuel-cladding materials for supercritical water-cooled reactor (SCWR), one of the promising generation IV reactor for large-scale electric power production. However, corrosion and stress corrosion cracking of these candidate alloys still remain to be a major problem in the selection of nuclear fuel cladding and other structural materials, such as water rod. Survey of literature and experimental results reveal that the general corrosion mechanism of those candidate materials exhibits quite complicated mechanism in high-temperature and high-pressure supercritical water. Formation of a stable protective oxide film is the key to the best corrosion-resistant alloys. This paper focuses on the mechanism of corrosion oxide film breakdown for SCWR candidate materials.

  10. Field test corrosion experiments in Denmark with biomass fuels Part I Straw firing

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Karlsson, A; Larsen, OH

    2002-01-01

    In Denmark, straw and other types of biomass are used for generating energy in power plants. Straw has the advantage that it is a "carbon dioxide neutral fuel" and therefore environmentally acceptable. Straw combustion is associated with corrosion problems which are not encountered in coal...

  11. Field test corrosion experiments in Denmark with biomass fuels Part I Straw firing

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Karlsson, A; Larsen, OH

    2002-01-01

    In Denmark, straw and other types of biomass are used for generating energy in power plants. Straw has the advantage that it is a "carbon dioxide neutral fuel" and therefore environmentally acceptable. Straw combustion is associated with corrosion problems which are not encountered in coal...

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

    Science.gov (United States)

    De Bremaecker, Anne

    2012-09-01

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

  13. Effects of corrosion and precipitates on mechanical properties in the ferritic/martensitic steel cladding under ultra-long cycle fast reactor environment at 650 .deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Yong; Lee, Jeong Hyeon; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of); Shin, Sang Hun [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This changes chemical compositions of inter-surface and effects on behavior of precipitations. NaCrO{sub 2} which is ternary sodium compound occurs intergranular corrosion resulting in thickness reduction. This change can cause a degradation of mechanical strength of structure material of UCFR. Therefore, we should consider longterm compatibility with sodium and study about life prediction. The research about ferritic/martensitic steel on effects of long term exposure in liquid sodium at 650 .deg. C, 20ppm oxygen includes weight loss of test material (Gr. 92) by corrosion and mechanism about nucleation and growth of precipitates like Laves-phase in bulk. There are many changes such as segregation of component to nucleate precipitates, affecting into microstructural evolution of the steel. Therefore, the thermochemical reaction research to predict behavior about precipitates should be performed. In a specific procedure, the micro-structure and the surface phenomenon of ferritic/martensitic steels (Gr. 92) that are exposed to liquid sodium at 650 .deg. C, 20 ppm oxygen and aged in high pure Argon gas environment to express bulk have been investigated by using scanning electron microscope (SEM) and transmission electron microscope (TEM). At 10 ppm oxygen designed oxygen value for UCFR, there is 107μm thickness reduction for 30 years. Thus, if there is no degradation of mechanical strength caused by aging effect, the tolerance of load of initial cladding should be higher than real load at least 23.6 %. Compared to specimens exposed to Ar-gas environment, Specimen which solutions are leaded into sodium has degradation of strength by reduction of solution hardening.

  14. Boiler and steam generator corrosion: Fossil fuel power plants. (Latest citations from the NTIS bibliographic database). Published Search

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-11-01

    The bibliography contains citations concerning corrosion effects, mechanisms, detection, and inhibition in fossil fuel fired boilers. Fluidized bed combustors and coal gasification are included in the applications. The citations examine hot corrosion, thermal mechanical degradation, and intergranular oxidation corrosion studies performed on the water side and hot gas side of heat exchanger tubes and support structures. Coatings and treatment of material to inhibit corrosion are discussed. Corrosion affecting nuclear powered steam generators is examined in a separate bibliography. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  16. The application of electrorefining for recovery and purification of fuel discharged from the Integral Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burris, L.; Steunenberg, R.K.; Miller, W.E.

    1986-01-01

    An electrorefining process employing a molten salt electrolyte and a molten cadmium anode is proposed for the separation of uranium and plutonium from fission products and cladding material in discharged IFR driver fuel. The use of a liquid cadmium anode, which is the unique feature of the process, permits selective dissolution of the fuel from the cladding and prevents electrolytic corrosion of the steel container and contamination of the product by noble metal fission products.

  17. Modelling fireside corrosion of heat exchangers in co-fired pulverised fuel power systems

    Energy Technology Data Exchange (ETDEWEB)

    Simms, N.J. [Cranfield Univ. (United Kingdom). Energy Technology Centre; Fry, A.T. [National Physical Laboratory, Teddington, Middlesex (United Kingdom)

    2010-07-01

    As a result of concerns about the effects of CO{sub 2} emissions on the global environment, there is increasing pressure to reduce such emissions from power generation systems. The use of biomass co-firing with coal in conventional pulverised fuel power stations has provided the most immediate route to introduce a class of fuel that is regarded as both sustainable and carbon neutral. In the future it is anticipated that increased levels of biomass will need to be used in such systems to achieve the desired CO{sub 2} emission targets. However there are concerns over the risk of fireside corrosion damage to the various heat exchangers and boiler walls used in such systems. Future pulverised fuel power systems will need to be designed to cope with the effects of using a wide range of coal-biomass mixes. However, such systems will also need to use much higher heat exchanger operating temperatures to increase their conversion efficiencies and counter the effects of the CO{sub 2} capture technologies that will need to be used in them. Higher operating temperatures will also increase the risk of fireside corrosion damage to the critical heat exchangers. This paper reports work that has been carried out to develop quantitative corrosion models for heat exchangers in pulverised fuel power systems. These developments have been particularly targeted at producing models that enable the evaluation of the effects of using different coal-biomass mixtures and of increasing heat exchanger operating conditions. Models have been produced that have been targeted at operating conditions and materials used in (a) superheaters/reheaters and (b) waterwalls. Data used in the development of these models has been produced from full scale and pilot scale plants in the UK using a wide range of coal and biomass mixtures, as well as from carefully targeted series of laboratory corrosion tests. Mechanistic and neural network based models have been investigated during this development process to

  18. Corrosion of used nuclear fuel in aqueous perchlorate and carbonate solutions

    Science.gov (United States)

    Shoesmith, D. W.; Sunder, S.; Bailey, M. G.; Miller, N. H.

    1996-01-01

    The corrosion of used fuel was investigated using electrodes constructed from fuel pins discharged from the Pickering, Bruce and Darlington CANDU reactors, and compared to the corrosion behaviour observed on unirradiated UO 2 and SIMFUEL. Experiments were carried out in solutions of NaClO 4 (pH˜ 9.5) in the presence and absence of (a) substantial concentrations of sodium carbonate, and (b) additional external gamma fields. Used fuel electrodes reached oxidizing corrosion potentials ( ECORR) rapidly compared with unirradiated UO 2 electrodes. However, optical and SEM examinations showed no evidence for rapid oxidative dissolution. This reaction, expected to be fast since high values of ECORR are observed, appears to be blocked by the accumulation of secondary phases in grain boundaries. The oxidation and dissolution behaviour of used fuel is determined predominantly by (i) the dose rate in solution near the fuel surface, (ii) the extent of burnup (which determines the degree of fission product doping), and (iii) the degree of non-stoichiometry.

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

    Science.gov (United States)

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

    2017-04-01

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

  20. Corrosion of irradiated MOX fuel in presence of dissolved H 2

    Science.gov (United States)

    Carbol, P.; Fors, P.; Van Winckel, S.; Spahiu, K.

    2009-07-01

    The corrosion behaviour of irradiated MOX fuel (47 GWd/tHM) has been studied in an autoclave experiment simulating repository conditions. Fuel fragments were corroded at room temperature in a 10 mM NaCl/2 mM NaHCO 3 solution in presence of dissolved H 2 for 2100 days. The results show that dissolved H 2 in concentration 1 mM and higher inhibits oxidation and dissolution of the fragments. Stable U and Pu concentrations were measured at 7 × 10 -10 and 5 × 10 -11 M, respectively. Caesium was only released during the first two years of the experiment. The results indicate that the UO 2 matrix of a spent MOX fuel is the main contributor to the measured dissolution, while the corrosion of the high burn-up Pu-rich islands appears negligible.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A [ORNL

    2012-08-01

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

  2. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

    Science.gov (United States)

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

    2015-12-01

    Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10-20Cr, 3-5Al, and 0-0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitive to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741 °C.

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

    Science.gov (United States)

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

    2016-06-01

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

  4. Deformation behavior of laser welds in high temperature oxidation resistant Fe-Cr-Al alloys for fuel cladding applications

    Science.gov (United States)

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

    2014-11-01

    Ferritic-structured Fe-Cr-Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability and post-weld mechanical behavior of three model alloys in a range of Fe-(13-17.5)Cr-(3-4.4)Al (wt.%) with a minor addition of yttrium using modern laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds using sub-sized, flat dog-bone tensile specimens and digital image correlation (DIC) has been carried out to determine the performance of welds as a function of alloy composition. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. For all proposed alloys, laser welding resulted in a defect free weld devoid of cracking or inclusions.

  5. The Impact of Microbially Influenced Corrosion on Spent Nuclear Fuel and Storage Life

    Energy Technology Data Exchange (ETDEWEB)

    J. H. Wolfram; R. E. Mizia; R. Jex; L. Nelson; K. M. Garcia

    1996-10-01

    A study was performed to evaluate if microbial activity could be considered a threat to spent nuclear fuel integrity. The existing data regarding the impact of microbial influenced corrosion (MIC) on spent nuclear fuel storage does not allow a clear assessment to be made. In order to identify what further data are needed, a literature survey on MIC was accomplished with emphasis on materials used in nuclear fuel fabrication, e.g., A1, 304 SS, and zirconium. In addition, a survey was done at Savannah River, Oak Ridge, Hanford, and the INEL on the condition of their wet storage facilities. The topics discussed were the SNF path forward, the types of fuel, ramifications of damaged fuel, involvement of microbial processes, dry storage scenarios, ability to identify microbial activity, definitions of water quality, and the use of biocides. Information was also obtained at international meetings in the area of biological mediated problems in spent fuel and high level wastes. Topics dis cussed included receiving foreign reactor research fuels into existing pools, synergism between different microbes and other forms of corrosion, and cross contamination.

  6. Protection of porous carbon fuel particles from boudouard corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, John F.

    2015-05-26

    A system for producing energy that includes infusing porous carbon particles produced by pyrolysis of carbon-containing materials with an off-eutectic salt composition thus producing pore-free carbon particles, and reacting the carbon particles with oxygen in a fuel cell according to the reaction C+O.sub.2=CO.sub.2 to produce electrical energy.

  7. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore; Ionescu, Drags; Pauna, Eduard [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.

    2012-03-15

    When nuclear power reactors are operated in a load following (LF) mode, the nuclear fuel may be subjected to step changes in power on weekly, daily, or even hourly basis, depending on the grid's needs. Two load following tests performed in TRIGA Research Reactor of Institute for Nuclear Research (INR) Pitesti were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets in the corrosive environment. The 3D finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath at ridge region. This paper summarizes the results of the analytical assessment for SCF and their relation to CANDU fuel performance in LF tests conditions. (orig.)

  8. The influence of hydrogen peroxide and hydrogen on the corrosion of simulated spent nuclear fuel.

    Science.gov (United States)

    Razdan, Mayuri; Shoesmith, David W

    2015-01-01

    The synergistic influence between H(2)O(2) and H(2) on the corrosion of SIMFUEL (simulated spent nuclear fuel) has been studied in solutions with and without added HCO(3)(-)/CO(3)(2-). The response of the surface to increasing concentrations of added H(2)O(2) was monitored by measuring the corrosion potential in either Ar or Ar/H(2)-purged solutions. Using X-ray photoelectron spectroscopy it was shown that the extent of surface oxidation (U(V) + U(VI) content) was directly related to the corrosion potential. Variations in corrosion potential with time, redox conditions, HCO(3)(-)/CO(3)(2-) concentration, and convective conditions showed that surface oxidation induced by H(2)O(2) could be reversed by reaction with H(2), the latter reaction occurring dominantly on the noble metal particles in the SIMFUEL. For sufficiently large H(2)O(2) concentrations, the influence of H(2) was overwhelmed and irreversible oxidation of the surface to U(VI) occurred. Subsequently, corrosion was controlled by the chemical dissolution rate of this U(VI) layer.

  9. A Semi-Empirical Two Step Carbon Corrosion Reaction Model in PEM Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Young, Alan; Colbow, Vesna; Harvey, David; Rogers, Erin; Wessel, Silvia

    2013-01-01

    The cathode CL of a polymer electrolyte membrane fuel cell (PEMFC) was exposed to high potentials, 1.0 to 1.4 V versus a reversible hydrogen electrode (RHE), that are typically encountered during start up/shut down operation. While both platinum dissolution and carbon corrosion occurred, the carbon corrosion effects were isolated and modeled. The presented model separates the carbon corrosion process into two reaction steps; (1) oxidation of the carbon surface to carbon-oxygen groups, and (2) further corrosion of the oxidized surface to carbon dioxide/monoxide. To oxidize and corrode the cathode catalyst carbon support, the CL was subjected to an accelerated stress test cycled the potential from 0.6 VRHE to an upper potential limit (UPL) ranging from 0.9 to 1.4 VRHE at varying dwell times. The reaction rate constants and specific capacitances of carbon and platinum were fitted by evaluating the double layer capacitance (Cdl) trends. Carbon surface oxidation increased the Cdl due to increased specific capacitance for carbon surfaces with carbon-oxygen groups, while the second corrosion reaction decreased the Cdl due to loss of the overall carbon surface area. The first oxidation step differed between carbon types, while both reaction rate constants were found to have a dependency on UPL, temperature, and gas relative humidity.

  10. Ash deposition and high temperature corrosion at combustion of aggressive fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hede Larsen, O. [I/S Fynsvaerket, Faelleskemikerne, Odense (Denmark); Henriksen, N. [Elsamprojekt A/S, Faelleskemikerne, Fredericia (Denmark)

    1996-12-01

    In order to reduce CO{sub 2} emission, ELSAM is investigating the possibilities of using biomass - mainly straw - for combustion in high efficiency power plants. As straw has very high contents of chlorine and potassium, a fuel with high corrosion and ash deposition propensities has been introduced. ELSAM has investigated 3 ultra supercritical boiler concepts for combustion of straw alone or together with coal: (1) PF boilers with a relatively low share of straw, (2) CFB boilers with low to high share of straw and (3) vibrating grate boilers with 100% straw. These investigations has mainly been full-scale tests with straw fed into existing boilers. Corrosion tests have been performed in these boilers using temperature regulated probes and in-plant test tubes in existing superheaters. The corrosion has been determined by detailed measurements of wall thickness reduction and light optical microscopic measurements of the material degradation due to high temperature corrosion. Corrosion mechanisms have been evaluated using SEM/EDX together with thermodynamical considerations based on measurements of the chemical environment in the flue gas. Ash deposition is problematic in CFB boilers and in straw fired boilers, especially in years with high potassium and chlorine content of the straw. This ash deposition also is related to condensation of KCl and can probably only be handled by improved cleaning devices. (EG)

  11. High temperature corrosion by combustion gases produced by burning liquid fuels containing sulphur, sodium and vanadium.

    OpenAIRE

    Khan, Fazlur Rahman

    1980-01-01

    High temperature corrosion, at 730° C, by combustion gases produced by burning liquid fuels in a laboratory combustor has been investigated. A selected range of steels and alloys (mild steel, stainless steel type 347, Nimonic N90, N105, and IN657) have been tested in the combustion gases using fuels containing varying amounts of impurities in the range of 0 - 6% sulphur, 0 - 60 ppm sodium, and 0 - 300 ppm vanadium. On the basis of the comprehensive results a computer programme was written t...

  12. Results of Severe Fuel Damage Experiment QUENCH-14 with Advanced Rod Cladding M5®. (KIT Scientific Reports ; 7549)

    OpenAIRE

    STUCKERT J.; Große, M.; Stegmaier, U.; Steinbrück, M.

    2010-01-01

    The QUENCH experiments are to investigate the hydrogen release resulting from the water injection into an uncovered core of a Light Water Reactor as well as the high-temperature behavior of core materials. The QUENCH-14 experiment investigated the effect of M5® cladding material on bundle oxidation and core reflood, in comparison with the tests QUENCH-06 that used standard Zircaloy-4 and QUENCH-12 that used VVER E110-claddings.

  13. Modelling cladding response to changing conditions

    Energy Technology Data Exchange (ETDEWEB)

    Tulkki, Ville; Ikonen, Timo [VTT Technical Research Centre of Finland ltd (Finland)

    2016-11-15

    The cladding of the nuclear fuel is subjected to varying conditions during fuel reactor life. Load drops and reversals can be modelled by taking cladding viscoelastic behaviour into account. Viscoelastic contribution to the deformation of metals is usually considered small enough to be ignored, and in many applications it merely contributes to the primary part of the creep curve. With nuclear fuel cladding the high temperature and irradiation as well as the need to analyse the variable load all emphasise the need to also inspect the viscoelasticity of the cladding.

  14. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rodney C. Ewing

    2004-10-07

    Spent nuclear fuel, essentially U{sub 2}, accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO{sub 2} in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-01

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

  17. Evaluation of the effect of B and N on the microstructure of 9Cr-2W steel during an aging treatment for SFR fuel cladding tubes

    Science.gov (United States)

    Jeong, Eun Hee; Park, Sang-Gyu; Kim, Sung Ho; Kim, Young Do

    2015-12-01

    In this study, the microstructure of sodium-cooled fast reactor (SFR) fuel cladding steel with different B and N contents after aging is compared. The addition of nitrogen produces a large quantity of MX precipitates with sizes of 0.1 μm or smaller during the initial thermal treatment process and this contributes to help such precipitates maintain stability without being excessively affected by aging. B is primarily distributed in the grain boundary precipitates and grain interior precipitates in the initial stage. The B distribution is believed to move to the Cr precipitates after 7000 h and to contribute to suppressing the growth of M23C6.

  18. In-situ crack repair by laser cladding

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2010-09-01

    Full Text Available sealing is achieved, an overlay layer of typically 1 mm thickness is cladded for improved pitting corrosion resistance. Crack sealing is considered to be a temporary repair technique. In-situ repair requires that the equipment should be mobile... 10 2 3 9 10 deg 10 deg 10 deg Table 1: Typical process parameters for crack sealing 2.2 Overlay cladding Overlay cladding of the sealed cracks is required to improve pitting corrosion...

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

    Science.gov (United States)

    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.

  20. Modelling of iodine-induced stress corrosion cracking in CANDU fuel

    Science.gov (United States)

    Lewis, B. J.; Thompson, W. T.; Kleczek, M. R.; Shaheen, K.; Juhas, M.; Iglesias, F. C.

    2011-01-01

    Iodine-induced stress corrosion cracking (I-SCC) is a recognized factor for fuel-element failure in the operation of nuclear reactors requiring the implementation of mitigation measures. I-SCC is believed to depend on certain factors such as iodine concentration, oxide layer type and thickness on the fuel sheath, irradiation history, metallurgical parameters related to sheath like texture and microstructure, and the mechanical properties of zirconium alloys. This work details the development of a thermodynamics and mechanistic treatment accounting for the iodine chemistry and kinetics in the fuel-to-sheath gap and its influence on I-SCC phenomena. The governing transport equations for the model are solved with a finite-element technique using the COMSOL Multiphysics® commercial software platform. Based on this analysis, this study also proposes potential remedies for I-SCC.

  1. The Role of X-Ray Diffraction for Analyzing Zr-Sn-Nb-Fe Alloys as Power Reactor Fuel Cladding

    Directory of Open Access Journals (Sweden)

    Sugondo

    2010-08-01

    Full Text Available Synthesis of Zr-1%Nb-1%Sn-1%Fe alloy is undertaken in order to develop fuel cladding alloy at high burn-up. Powder specimens of Zr-Sn-Nb-Fe alloy were prepared and then formed into pellets with a dimension of 10 mm in height 10 mm in diameter using a pressure of 1.2 ton/cm2. The 5 gram green pellets were then melted in an arc furnace crucible under argon atmosphere. The pressure in the furnace was set at 2 psi and the current was 50 A. Afterwards, the ingots were heated at a temperature of 1100°C for 2 hours and subsequently quenched in water. The ingots then underwent annealing at temperatures of 400°C, 500°C, 600°C, 700°C, and 750°C for 2 hours. The specimens were analyzed using X-ray diffraction in order to construct diffractograms. Results of the diffraction patterns were fitted with data from JCPDF (Joint Committee Powder Diffraction File to determine the type of crystals in the elements or substances. The greater the crystallite dimension, the smaller the dislocation density. Agreeable results for hardening or strengthening were obtained at annealing temperatures of 500°C and 700, whereas for softening or residual stress at 600°C and 750°C. The nucleation of the secondary phase precipitate (SPP was favourable at annealing temperatures of 400°C, 500°C, and 700°C. For Zr-1%Nb-1%Sn-1%Fe alloy with annealing temperatures between 400°C to 800°C, precipitates of Fe2Nb, ZrSn2,FeSn, SnZr, NbSn2, Zr0.68Nb0.25Fe0.08, Fe2Nb0.4Zr0.6, Fe37Nb9Zr54, and ω-Zr were observed. Satisfactory precipitate stabilization was achieved at annealing temperature of 800°C, growth of precipitates at temperature between 500°C to 600°C, and minimization of precipitate size at 700°C.

  2. Results of NDE Technique Evaluation of Clad Hydrides

    Energy Technology Data Exchange (ETDEWEB)

    Kunerth, Dennis C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This report fulfills the M4 milestone, M4FT-14IN0805023, Results of NDE Technique Evaluation of Clad Hydrides, under Work Package Number FT-14IN080502. During service, zirconium alloy fuel cladding will degrade via corrosion/oxidation. Hydrogen, a byproduct of the oxidation process, will be absorbed into the cladding and eventually form hydrides due to low hydrogen solubility limits. The hydride phase is detrimental to the mechanical properties of the cladding and therefore it is important to be able to detect and characterize the presence of this constituent within the cladding. Presently, hydrides are evaluated using destructive examination. If nondestructive evaluation techniques can be used to detect and characterize the hydrides, the potential exists to significantly increase test sample coverage while reducing evaluation time and cost. To demonstrate the viability this approach, an initial evaluation of eddy current and ultrasonic techniques were performed to demonstrate the basic ability to these techniques to detect hydrides or their effects on the microstructure. Conventional continuous wave eddy current techniques were applied to zirconium based cladding test samples thermally processed with hydrogen gas to promote the absorption of hydrogen and subsequent formation of hydrides. The results of the evaluation demonstrate that eddy current inspection approaches have the potential to detect both the physical damage induced by hydrides, e.g. blisters and cracking, as well as the combined effects of absorbed hydrogen and hydride precipitates on the electrical properties of the zirconium alloy. Similarly, measurements of ultrasonic wave velocities indicate changes in the elastic properties resulting from the combined effects of absorbed hydrogen and hydride precipitates as well as changes in geometry in regions of severe degradation. However, for both approaches, the signal responses intended to make the desired measurement incorporate a number of contributing

  3. Thermal Shock Properties of Cladding with SiC{sub f}/SiC Composite Protective Films

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Donghee; Park, Kwangheon [Kyunghee University, Yongin (Korea, Republic of); Kim, Weonju; Park, Jiyeon; Kim, Daejong; Lee, Hyeon Geun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In general, Zr-4 alloy is used for such nuclear fuel cladding. Zr-4 possesses a very small thermal neutron absorption cross-section and has superior corrosion resistance in the normal operating conditions of a nuclear reactor. However, in the case of a critical accident such as a LOCA (loss-of-coolant accident) in the Fukushima disaster, the risk of hydrogen explosion becomes serious. That is, in the case of coolant leakage, a dramatic reaction between the nuclear fuel cladding and steam can cause a heating reaction accompanied by rapid high-temperature oxidation, while creating a huge amount of hydrogen. Hence, the search for an alternative material for nuclear fuel cladding is being actively undertaken. Ceramic-based nuclear fuel cladding is receiving much attention as a means of improving safety. SiC has excellent properties of resistance to high temperature and high exposure and superior mechanical properties, as well as a very small thermal neutron absorption cross-section (0.09 barns), which causes almost no decrease in mechanical strength or volume change following exposure. This experiment examined the thermal shock properties and microstructure of cladding that has SiCf/SiC composite protective film, using polycarbosilane preceramic polymer.

  4. A Review of Carbide Fuel Corrosion for Nuclear Thermal Propulsion Applications

    Science.gov (United States)

    Pelaccio, Dennis G.; El-Genk, Mohamed S.; Butt, Darryl P.

    1994-07-01

    At the operation conditions of interest in nuclear thermal propulsion reactors, carbide materials have been known to exhibit a number of life limiting phenomena. These include the formation of liquid, loss by vaporization, creep and corresponding gas flow restrictions, and local corrosion and fuel structure degradation due to excessive mechanical and/or thermal loading. In addition, the radiation environment in the reactor core can produce a substantial change in its local physical properties, which can produce high thermal stresses and corresponding stress fractures (cracking). Time-temperature history and cyclic operation of the nuclear reactor can also accelerate some of these processes. The University of New Mexico's Institute for Space Nuclear Power Studies, under NASA sponsorship has recently initiated a study to model the complicated hydrogen corrosion process. In support of this effort, an extensive review of the open literature was performed, and a technical expert workshop was conducted. This paper summarizes the results of this review.

  5. Effect of Heat treatment and Aging Conditions on the Microstructure and Mechanical Properties of HT9 Steel for Fuel Cladding Tube

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyeong Min; Kim, Jong Lyeol [Hanyang university, Ansan (Korea, Republic of); Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel cladding tube is the most important safety barrier in a fission nuclear reactor. Thermal creep and void swelling occur by fission gas at high temperature during service time. Ferritic-martensitic steels are being considered attractive candidate materials for a fuel cladding of the SFR owing to their low expansion coefficients, high thermal conductivities and excellent irradiation resistance to void swelling. However, HT9 steel has a problem of a relatively low high temperature strength and low creep strength. To solve this problem, a study was conducted to increase the high temperature strength by changing the intermediate heat treatment step in the fabrication process of ferritic martensitic steel and controlling the microstructure and precipitate within the material. 700-780 .deg. C contributed to the increase in precipitate size, and the decrease in yield stress and hardness. An empirical equation for the mechanical properties of HT9 was suggested as a function of the microstructure and Hollomon-Jaffe tempering parameter. The results show that the size of the carbide and lath increased after aging, whereas the size of the prior austenite grain was not changed. Both the strength and hardness were decreased with aging, and this tendency saturated after 3000 hours of aging.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-15

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

  7. Permanent monitoring of the corrosive impact of fuel mixtures; Permanentes Monitoring der korrosiven Wirkung von Brennstoff-Mix

    Energy Technology Data Exchange (ETDEWEB)

    Deuerling, Christian; Waldmann, Barbara [Corrmoran GmbH, Augsburg (Germany)

    2013-03-01

    The knowledge of an acute corrosion load of an incinerator facilitates the recognition of enhanced burdens at the time of damage formation and the possibility to react fundamentally on these burdens. Especially with the utilization of complex fuel mixtures, the empirical measurement of the corrosion load is a much more reliable method for the evaluation of the wearing of the incinerator. The corrosion measurement becomes an early warning signal increasing the planning security of a plant by recognizing the strongly stressing operation prior to the occurrence of damages. Thus, follow-up costs can be saved. The clarification of the causes of corrosion facilitates a fact-based procedure between constructor, operator and supplier of fuels in order to guarantee a profitable operation at increasingly difficult boundary conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-26

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

  9. Spent nuclear fuel. A review of properties of possible relevance to corrosion processes

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R. [Caledon Consult AB, Nykoeping (Sweden)

    1995-04-01

    The report reviews the properties of spent fuel which are considered to be of most importance in determining the corrosion behaviour in groundwaters. Pellet cracking and fragment size distribution are discussed, together with the available results of specific surface area measurements on spent fuel. With respect to the importance of fuel microstructure, emphasis is placed on recent work on the so called structural rim effect, which consists of the formation of a zone of high porosity, and the polygonization of fuel grains to form many sub-grains, at the pellet rim, and appears to be initiated when the average pellet burnup exceeds a threshold of about 40 MWd/kgU. Due to neutron spectrum effects, the pellet rim is also associated with the buildup of plutonium and other actinides, which results in an enhanced local burnup and specific activity of both beta-gamma and alpha radiation, thus representing a greater potential for radiolysis effects in ingressed groundwater. The report presents and discusses the results of quantitative determination of the radial profiles of burnup and alpha activity on spent fuel with average burnups from 21.2 to 49 MWd/kgU. In addition to the radial variation of fission product and actinide inventories due to the effects mentioned above, migration, redistribution and release of some fission products can occur during reactor irradiation and the report concludes with a short review of these processes.

  10. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  11. State-of-the-art report on the development of liquid metal reactor fuel cladding materials in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Ho; Kuk, Il Hiun; Ryu, Woo Seog; Jang, Jin Sung; Rhee, Chang Kyu; Kim, Dae Whan; Park, Soon Dong; Kim, Woo Gon; Chung, Man Kyo; Han, Chang Hee

    1998-01-01

    PNC 1520 and PNC-FM5 have been developed as a cladding materials for LMR in Japan. PNC 1520 has superior swelling resistance and high temperature properties to PNC 31.6. And PNC-FMS steel has shown a high rupture stress as well as good neutron irradiation performance. In addition oxide dispersed ferritic steel (PNC-ODS) and 12Cr-8Mo steel have been developed. This report will give an insight for choosing and developing the materials to be applied to the KAERI prototype liquid metal reactor which is going to be operable in 2010 by analysis of the characteristics of cladding materials developed in Japan. (author). 39 refs., 2 tabs., 23 figs

  12. Measurement of fuel corrosion products using planar laser-induced fluorescence

    Science.gov (United States)

    Wantuck, Paul J.; Sappey, Andrew D.; Butt, Darryl P.

    1993-01-01

    Characterizing the corrosion behavior of nuclear fuel material in a high-temperature hydrogen environment is critical for ascertaining the operational performance of proposed nuclear thermal propulsion (NTP) concepts. In this paper, we describe an experimental study undertaken to develop and test non-intrusive, laser-based diagnostics for ultimately measuring the distribution of key gas-phase corrosion products expected to evolve during the exposure of NTP fuel to hydrogen. A laser ablation technique is used to produce high temperature, vapor plumes from uranium-free zirconium carbide (ZrC) and niobium carbide (NbC) forms for probing by various optical diagnostics including planar laser-induced fluorescence (PLIF). We discuss the laser ablation technique, results of plume emission measurements, and we describe both the actual and proposed planar LIF schemes for imaging constituents of the ablated ZrC and NbC plumes. Envisioned testing of the laser technique in rf-heated, high temperature gas streams is also discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hourdequin, N.

    1995-05-01

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

  14. Wavelength dependent neutron transmission and radiography investigations of the high temperature behaviour of materials applied in nuclear fuel and control rod claddings

    Science.gov (United States)

    Grosse, M.; Steinbrueck, M.; Kaestner, A.

    2011-09-01

    Neutron radiography was used for the investigation of the nuclear fuel and control rod cladding behaviour during steam oxidation under severe nuclear accident conditions. In order to verify the hypothesis that the unexpectedly high neutron cross-section found after oxidation of Zircaloy-4 in wet air containing 10% steam is caused by a strong hydrogen uptake, the wavelength dependence of the total macroscopic neutron cross-section of the specimens was measured. The characteristic dependence for hydrogen was not found, which is a proof that hydrogen is not absorbed significantly. The data agree mostly with the behaviour expected for β-Zr. Examinations of control rod simulators annealed until the failure in single-rod tests were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. This procedure clearly showed that the local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube is the reason for the failure.

  15. Wavelength dependent neutron transmission and radiography investigations of the high temperature behaviour of materials applied in nuclear fuel and control rod claddings

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M., E-mail: Mirco.Grosse@KIT.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Steinbrueck, M. [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Kaestner, A. [Department of Spallation Source, Paul Scherrer Institute (PSI), CH-5232 Villigen (Switzerland)

    2011-09-21

    Neutron radiography was used for the investigation of the nuclear fuel and control rod cladding behaviour during steam oxidation under severe nuclear accident conditions. In order to verify the hypothesis that the unexpectedly high neutron cross-section found after oxidation of Zircaloy-4 in wet air containing 10% steam is caused by a strong hydrogen uptake, the wavelength dependence of the total macroscopic neutron cross-section of the specimens was measured. The characteristic dependence for hydrogen was not found, which is a proof that hydrogen is not absorbed significantly. The data agree mostly with the behaviour expected for {beta}-Zr. Examinations of control rod simulators annealed until the failure in single-rod tests were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. This procedure clearly showed that the local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube is the reason for the failure.

  16. Evaluation of the effect of B and N on the microstructure of 9Cr–2W steel during an aging treatment for SFR fuel cladding tubes

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Eun Hee [Hanyang University, Department Materials Science and Engineering, 222, Wangsimni-ro, Seongdong-gu, Seoul, 133-791 (Korea, Republic of); Park, Sang-Gyu; Kim, Sung Ho [KAERI, Advanced Fuel Development Division, 989-111 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Kim, Young Do, E-mail: ydkim1@hanyang.ac.kr [Hanyang University, Department Materials Science and Engineering, 222, Wangsimni-ro, Seongdong-gu, Seoul, 133-791 (Korea, Republic of)

    2015-12-15

    In this study, the microstructure of sodium-cooled fast reactor (SFR) fuel cladding steel with different B and N contents after aging is compared. The addition of nitrogen produces a large quantity of MX precipitates with sizes of 0.1 μm or smaller during the initial thermal treatment process and this contributes to help such precipitates maintain stability without being excessively affected by aging. B is primarily distributed in the grain boundary precipitates and grain interior precipitates in the initial stage. The B distribution is believed to move to the Cr precipitates after 7000 h and to contribute to suppressing the growth of M{sub 23}C{sub 6}.

  17. Ultrasonic Inspection for Zirconium Alloy Nuclear Fuel Cladding Tubes%核燃料锆合金包壳管的超声波探伤

    Institute of Scientific and Technical Information of China (English)

    夏健文; 韩承

    2016-01-01

    介绍压水堆核燃料锆合金包壳管(Φ10.0 mm×0.70 mm)的超声波自动探伤方法和工艺,讨论不同长度、宽度、深度、角度的纵向和横向人工缺陷的超声响应结果.通过对检测出缺陷的典型包壳管进行金相解剖,确定缺陷性质和实际尺寸,验证超声探伤结果.针对实际探伤中的问题,考虑质量和成本控制,提出对不同缺陷的验收准则.实践应用表明,现行探伤方法和工艺能检出管材不同位置处10μm级的微小缺陷.但受缺陷的类型、取向的影响,探伤仪检测得到的回波幅度并不能完全真实地反应缺陷的实际大小和性质,需要在实际探伤时针对管材的制造工艺水平采取适当的加严措施,对不同的缺陷加以控制,才能更好地保证核燃料包壳管的质量.%The cladding tube is the main component of the nuclear fuel assembly,and as the first protective barrier,its quality is very important for the safe operation of nuclear power plants.After the completion of cladding tubes,a non-destructive testing is required,in which the ultrasonic inspection is a primary method.This paper introduces the ultrasonic flaw testing method and techniques of the zirconium alloy nuclear fuel cladding tubes for pressurized water reactor (PWR),which used in automatic ultrasonic inspection equipment,and discusses the detector response to the longitudinal and transverse artificial defects of different length,width,depth and angle.Its actual shape and size are measured by metallographic anatomical analysis for some typical defects to confirm the flaw detection results.Consider its quality and cost control,the acceptance rules are proposed for different defects.The application shows that the existing detection method and process can inspect the fine defects about 10μm at different locations of the cladding tube.Due to the influence of the defect type and orientation,the echo amplitude obtained by the detector is not completely true to the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L. [Pacific Northwest Laboratory, Richland, WA (United States)

    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.

  20. Steam generator corrosion 2007; Dampferzeugerkorrosion 2007

    Energy Technology Data Exchange (ETDEWEB)

    Born, M. (ed.)

    2007-07-01

    Between 8th and 9th November, 2007, SAXONIA Standortentwicklungs- und -verwertungsgesellschaft GmbH (Freiberg, Federal Republic of Germany) performed the 3rd Freiberger discussion conference ''Fireside boiler corrosion''. The topics of the lectures are: (a) Steam generator corrosion - an infinite history (Franz W. Alvert); (b) CFD computations for thermal waste treatment plants - a contribution for the damage recognition and remedy (Klaus Goerner, Thomas Klasen); (c) Experiences with the use of corrosion probes (Siegfried R. Horn, Ferdinand Haider, Barbara Waldmann, Ragnar Warnecke); (d) Use of additives for the limitation of the high temperature chlorine corrosion as an option apart from other measures to the corrosion protection (Wolfgang Spiegel); (e) Current research results and aims of research with respect to chlorine corrosion (Ragnar Warnecke); (f) Systematics of the corrosion phenomena - notes for the enterprise and corrosion protection (Thomas Herzog, Wolfgang Spiegel, Werner Schmidl); (g) Corrosion protection by cladding in steam generators of waste incinerators (Joerg Metschke); (h) Corrosion protection and wear protection by means of thermal spraying in steam generators (Dietmar Bendix); (i) Review of thick film nickelized components as an effective protection against high-temperature corrosion (Johann-Wilhelm Ansey); (j) Fireproof materials for waste incinerators - characteristics and profile of requirement (Johannes Imle); (k) Service life-relevant aspects of fireproof linings in the thermal recycling of waste (Till Osthoevener and Wolfgang Kollenberg); (l) Alternatives to the fireproof material in the heating space (Heino Sinn); (m) Cladding: Inconal 625 contra 686 - Fundamentals / applications in boiler construction and plant construction (Wolfgang Hoffmeister); (n) Thin films as efficient corrosion barriers - thermal spray coating in waste incinerators and biomass firing (Ruediger W. Schuelein, Steffen Hoehne, Friedrich

  1. Microstructure of laser cladded martensitic stainless steel

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2006-08-01

    Full Text Available for laser cladding Table 2 - Process parameters for coaxial and off axis powder cladding Material Laser power (W) Speed (m/min) Powder feed rate (kg/h) Carrying gas (Flow rate l/min) Stepove r (mm) Dilution (%) Fe211-1 (420) Off axis... mill industries. Conventional arc welding processes usually result in microstructures consisting of martensite and ferrite. Delta ferrite lowers the hot cracking susceptibility but also reduce the strength, thermal fatigue and corrosion properties...

  2. A Carbon Corrosion Model to Evaluate the Effect of Steady State and Transient Operation of a Polymer Electrolyte Membrane Fuel Cell

    CERN Document Server

    Pandy, Arun; Gummalla, Mallika; Atrazhev, Vadim V; Kuzminyh, Nikolay Yu; Sultanov, Vadim I; Burlatsky, Sergei F

    2014-01-01

    A carbon corrosion model is developed based on the formation of surface oxides on carbon and platinum of the polymer electrolyte membrane fuel cell electrode. The model predicts the rate of carbon corrosion under potential hold and potential cycling conditions. The model includes the interaction of carbon surface oxides with transient species like OH radicals to explain observed carbon corrosion trends under normal PEM fuel cell operating conditions. The model prediction agrees qualitatively with the experimental data supporting the hypothesis that the interplay of surface oxide formation on carbon and platinum is the primary driver of carbon corrosion.

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

    Energy Technology Data Exchange (ETDEWEB)

    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. Laboratory Investigations of the High Temperature Corrosion of Various Materials in Simulated oxy-fuel and Conventional Coal Firing

    Energy Technology Data Exchange (ETDEWEB)

    Folkeson, N.; Pettersson, J.; Svensson, J.E. [Chalmers Univ. of Technology (Sweden); Hjornhede, A. [Vattenfall Power Consultant AB (Sweden); Montgomery, M. [Vattenfall Heat Nordic/DTU Mekanik (Denmark); Bjurman, M. [Vattenfall Research and Development AB (Sweden)

    2009-07-01

    Laboratory exposures in horizontal tube furnaces were conducted to test various materials for corrosion resistance in simulated oxy-fuel firing and conventional coal firing environments. Two different exposures were done at 630 C for 672 hours. The reaction atmosphere, consisting of CO{sub 2}, H{sub 2}O, O{sub 2}, N{sub 2} and SO{sub 2}, was mixed to resemble that of oxy-fuel firing in the first exposure and that of conventional coal firing in the second exposure (N{sub 2} was added during the second exposure only). Four different materials were tested in the first exposure; Sanicro 63, Alloy 800HT, 304L and 304HCu. In the second exposure four different materials were tested; 304L, Alloy 800HT, Kanthal APMT and NiCrAl. Apart from cleaned sample coupons, some samples pre-exposed in a test rig under oxy-fuel conditions with lignite as fuel and some pre-exposed with bituminous coal as fuel were investigated in the first exposure. In the second exposure some samples were pre-exposed in a rig under conventional firing conditions with lignite as fuel. The corrosion attack on the investigated samples was analysed by gravimetry, x-ray diffraction (XRD) and scanning electron microscopy (SEM) with energy dispersive x-ray (EDX). The SEM/EDX analysis was made on both the sample envelope and metallographic cross sections of the samples. The results show that there is small difference in the corrosion attack between the two environments. There was also little difference in oxide morphology and composition between cleaned samples and pre-exposed samples of the same material. The austenitic chromia former 304HCu suffered the most extensive corrosion attack in the oxy-fuel environment. In the conventional air firing environment 304L showed the highest mass gain. Chromia formers with higher chromium concentrations performed better, especially the super austenitic Alloy 800HT, with its high chromium concentration, formed a thin and protective corundum type oxide. The nickel based

  5. Water corrosion of spent nuclear fuel: radiolysis driven dissolution at the UO2/water interface.

    Science.gov (United States)

    Springell, Ross; Rennie, Sophie; Costelle, Leila; Darnbrough, James; Stitt, Camilla; Cocklin, Elizabeth; Lucas, Chris; Burrows, Robert; Sims, Howard; Wermeille, Didier; Rawle, Jonathan; Nicklin, Chris; Nuttall, William; Scott, Thomas; Lander, Gerard

    2015-01-01

    X-ray diffraction has been used to probe the radiolytic corrosion of uranium dioxide. Single crystal thin films of UO(2) were exposed to an intense X-ray beam at a synchrotron source in the presence of water, in order to simultaneously provide radiation fields required to split the water into highly oxidising radiolytic products, and to probe the crystal structure and composition of the UO(2) layer, and the morphology of the UO(2)/water interface. By modeling the electron density, surface roughness and layer thickness, we have been able to reproduce the observed reflectivity and diffraction profiles and detect changes in oxide composition and rate of dissolution at the Ångström level, over a timescale of several minutes. A finite element calculation of the highly oxidising hydrogen peroxide product suggests that a more complex surface interaction than simple reaction with H(2)O(2) is responsible for an enhancement in the corrosion rate directly at the interface of water and UO(2), and this may impact on models of long-term storage of spent nuclear fuel.

  6. Spent nuclear fuel corrosion: The application of ICP-MS to direct actinide analysis

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R. [Caledon-Consult AB, Nykoeping (Sweden); Eklund, U.B. [Studsvik Nuclear AB, Nykoeping (Sweden)

    1995-01-01

    The ICP-MS technique has been applied to the analysis of the actinide contents of corrodant solutions from experiments performed to study the corrosion of spent nuclear fuel in simulated groundwaters. Analysis was performed directly on the solutions, without employing separation or isotope dilution techniques. The results from two analytical campaigns using natural indium and thorium internal standards are compared. Under both oxic and anoxic conditions, the U contents can be determined with good accuracy and precision. The same applies to Np and Pu under oxic conditions, where the solution concentrations range down to about 0.1 ppb. Under anoxic conditions, where solution concentrations are lower by one or two orders of magnitude, reasonable results for these two actinides can be obtained, but with much lower precision. Direct analysis of Am and Cm, however, gave unsatisfactory results, since the technique is limited by poor measurement statistics and background uncertainty.

  7. Iodine induced stress corrosion of zirconium and zircaloy-4 mechanisms. Translation to pellet cladding interaction conditions in PWR type reactors; Mecanismes de corrosion sous contrainte par l`iode dans le zirconium et le zircaloy-4. Transposition aux conditions d`interaction pastille-gaine dans les reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Fregonese, M

    1997-10-08

    This thesis is linked to the study of the Pellet-Cladding Interaction (PCI) phenomenon in Pressurized Water Reactors, which can lead to cladding rupture by iodine Stress Corrosion Cracking (SCC) of Zircaloy-4. Results are obtained through slow tensile tests performed in iodine methanol and in iodine vapour, on reference material, neutron irradiated material, and iodine zirconium implanted material. They allow to propose an explanation of the rapidity of the ruptures observed during PCI loadings, and to make the link between laboratory SCC tests and power ramp tests. Indeed, neutron irradiation facilitates the initiation and the transgranular propagation steps of the SCC cracks, due to strain localization and hardening associated to the presence of irradiation defects. On the other hand, recoiled iodine does not seem to affect SCC susceptibility of the material. On a chemical point of view, thermally released iodine is then responsible for PCI/SCC ruptures. A detailed calculation of iodine amount created by fission and released in the gap during irradiation makes it possible to show that local iodine concentration facing the pellet-to-pellet and the radial pellet cracks regions is sufficient for SCC cracks to develop in the metal. Finally, a competition between re-passivation and cracking is underscored. This results are in good agreement with the occurrence of an iodine adsorption mechanism. Adsorption could be assisted by a corrosion-deformation interaction phenomenon, and/or by the formation of solid and gaseous zirconium iodides. (author) 132 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Alencar, Donizete Anderson de

    2004-07-01

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

  9. Corrosion investigations on zircaloy-4 and titanium dissolver materials for MOX fuel dissolution in concentrated nitric acid containing fluoride ions

    Energy Technology Data Exchange (ETDEWEB)

    Jayaraj, J.; Krishnaveni, P.; Krishna, D. Nanda Gopala; Mallika, C.; Mudali, U. Kamachi, E-mail: kamachi@igcar.gov.in

    2016-05-15

    Aqueous reprocessing of plutonium-rich mixed oxide fuels require fluoride as a dissolution catalyst in boiling nitric acid for an effective dissolution of the spent fuel. High corrosion rates were obtained for the candidate dissolver materials zircaloy-4 (Zr-4) and commercial pure titanium (CP-Ti grade 2) in boiling 11.5 M HNO{sub 3} + 0.05 M NaF. Complexing the fluoride ions either with Al(NO{sub 3}){sub 3} or ZrO(NO{sub 3}){sub 2} aided in decreasing the corrosion rates of Zr-4 and CP-Ti. From the obtained corrosion rates it is concluded that CP-Ti is a better dissolver material than Zr-4 for extended service life in boiling 11.5 M HNO{sub 3} + 0.05 M NaF, when complexed with 0.15 M ZrO(NO{sub 3}){sub 2}. XPS analysis confirmed the presence of TiO{sub 2} and absence of fluoride on the surface of CP-Ti samples, indicating that effective complexation had occurred in solution leading to passivation of the metal and imparting high corrosion resistance. - Highlights: • Zr-4 and CP-Ti exhibited high corrosion rate in boiling fluorinated nitric acid. • Corrosion rate decreased in fluorinated nitric acid containing ZrO(NO{sub 3}){sub 2} and Al(NO{sub 3}){sub 3}. • High inhibiting efficiency is exhibited by 0.15 M ZrO(NO{sub 3}){sub 2} when compared to Al(NO{sub 3}){sub 3}. • Corrosion rates of CP-Ti were negligible in complexed fluorinated nitric acid. • XPS analysis on CP-Ti confirmed the presence of TiO{sub 2} and absence of fluoride.

  10. Neutron Imaging Investigations of the Secondary Hydriding of Nuclear Fuel Cladding Alloys during Loss of Coolant Accidents

    Science.gov (United States)

    Grosse, M.; Roessger, C.; Stuckert, J.; Steinbrueck, M.; Kaestner, A.; Kardjilov, N.; Schillinger, B.

    The hydrogen concentration and distribution at both sides of the burst opening of cladding tubes used in three QUENCH-LOCA simulation bundle experiments were investigated by means of neutron radiography and tomography. The quantitative correlation between the total macroscopic neutron cross-section and the atomic number density ratio between hydrogen and zirconium was determined by testing calibration specimens with known hydrogen concentrations. Hydrogen enrichments located at the end of the ballooning zone of the tested tubes were detected in the inner rods of the test bundles. Nearly all of the peripheral claddings exposed to lower temperatures do not show such enrichments. This implies that under the conditions investigated a threshold temperature exists below which no hydrogen enrichments can be formed. In order to understand the hydrogen distribution a model was developed describing the processes occurring during loss of coolant accidents after rod burst. The general shape of the hydrogen distributions with a peak each side of the ballooning region is well predicted by this model whereas the absolute concentrations are underestimated compared to the results of the neutron tomography investigations. The model was also used to discuss the influence of the alloy composition on the secondary hydrogenation. Whereas the relations for the maximal hydrogen concentrations agree well for one and the same alloy, the agreement for tests with different alloys is less satisfying, showing that material parameters such as oxidation kinetics, phase transition temperature for the zirconium oxide, and yield strength and ductility at high temperature have to be taken into account to reproduce the results of neutron imaging investigations correctly.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

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

  12. Chemical thermodynamics of the system Cs--U--Zr--H--I--O in the light water reactor fuel-cladding gap

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, T.M.; Lindemer, T.B.

    1978-10-01

    Equilibrium thermodynamic calculations were performed on the Cs-U-Zr-H-I-O system that is assumed to exist in the fuel-cladding gap of light water reactor fuel under in-reactor, steam, and 50% steam--50% air conditions. The in-reactor oxygen potential is assumed to be controlled by either UO/sub 2+x/ + Cs/sub 2/UO/sub 4/ or Zr + ZrO/sub 2/. The important condensed phases in-reactor are UO/sub 2+x/, Cs/sub 2/UO/sub 4/, and CsI, and the major gaseous species are Cs, Cs/sub 2/, CsI, and Cs/sub 2/I/sub 2/. The presence of steam does not alter these species, although CsOH also becomes a major gaseous species. In a 50% steam--50% air mixture, the equilibrium condensed phases are U/sub 3/O/sub 8/ or UO/sub 3/ and Cs/sub 2/U/sub 15/O/sub 46/. Under a nonequilibrium situation where zirconium metal can react with iodine, ZrO/sub 3/ or liquid ZrI/sub 2/ is present, and the gaseous species ZrI/sub 3/ and ZrI/sub 4/ have large partial pressures.

  13. Cladding Alloys for Fluoride Salt Compatibility

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-06-01

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

  14. Simulating the corrosion of zirconium alloys in the water coolant of VVER reactors

    Science.gov (United States)

    Kritskii, V. G.; Berezina, I. G.; Motkova, E. A.

    2013-07-01

    A model for predicting the corrosion of cladding zirconium alloys depending on their composition and operating conditions is proposed. Laws of thermodynamics and chemical kinetics of the reactions through which the multicomponent zirconium alloy is oxidized in the reactor coolant constitute the physicochemical heart of the model. The developed version of the model is verified against the results obtained from tests of fuel rod claddings made of commercial-grade and experimental zirconium alloys carried out by different researchers under autoclave and reactor conditions. It is shown that the proposed model adequately describes the corrosion of alloys in coolants used at nuclear power stations. It is determined that, owing to boiling of coolant and its acidification in a VVER-1200 reactor, Zr-1% Nb alloys with additions of iron and oxygen must be more resistant to corrosion than the commercial-grade alloy E110.

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

    Energy Technology Data Exchange (ETDEWEB)

    DeVolpi, A. (comp.)

    1976-01-01

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

  16. FY04 Inspection Results for Wet Uruguay Fuel in L-Basin

    Energy Technology Data Exchange (ETDEWEB)

    VORMELKER, PHILIP

    2005-09-01

    The 2004 visual inspection of four Uruguay nuclear fuel assemblies stored in L-Basin was completed. This was the third inspection of this wet stored fuel since its arrival in the summer of 1998. Visual inspection photographs of the fuel from the previous and the recent inspections were compared and no evidence of significant corrosion was found on the individual fuel plate photographs. Fuel plates that showed areas of pitting in the cladding during the original receipt inspection were also identified during the 2004 inspection. However, a few pits were found on the non-fuel aluminum clamping plates that were not visible during the original and 2001 inspections.

  17. Corrosion investigations on zircaloy-4 and titanium dissolver materials for MOX fuel dissolution in concentrated nitric acid containing fluoride ions

    Science.gov (United States)

    Jayaraj, J.; Krishnaveni, P.; Krishna, D. Nanda Gopala; Mallika, C.; Mudali, U. Kamachi

    2016-05-01

    Aqueous reprocessing of plutonium-rich mixed oxide fuels require fluoride as a dissolution catalyst in boiling nitric acid for an effective dissolution of the spent fuel. High corrosion rates were obtained for the candidate dissolver materials zircaloy-4 (Zr-4) and commercial pure titanium (CP-Ti grade 2) in boiling 11.5 M HNO3 + 0.05 M NaF. Complexing the fluoride ions either with Al(NO3)3 or ZrO(NO3)2 aided in decreasing the corrosion rates of Zr-4 and CP-Ti. From the obtained corrosion rates it is concluded that CP-Ti is a better dissolver material than Zr-4 for extended service life in boiling 11.5 M HNO3 + 0.05 M NaF, when complexed with 0.15 M ZrO(NO3)2. XPS analysis confirmed the presence of TiO2 and absence of fluoride on the surface of CP-Ti samples, indicating that effective complexation had occurred in solution leading to passivation of the metal and imparting high corrosion resistance.

  18. Multiphysics modeling of two-phase film boiling within porous corrosion deposits

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Miaomiao, E-mail: mmjin@mit.edu; Short, Michael, E-mail: hereiam@mit.edu

    2016-07-01

    Porous corrosion deposits on nuclear fuel cladding, known as CRUD, can cause multiple operational problems in light water reactors (LWRs). CRUD can cause accelerated corrosion of the fuel cladding, increase radiation fields and hence greater exposure risk to plant workers once activated, and induce a downward axial power shift causing an imbalance in core power distribution. In order to facilitate a better understanding of CRUD's effects, such as localized high cladding surface temperatures related to accelerated corrosion rates, we describe an improved, fully-coupled, multiphysics model to simulate heat transfer, chemical reactions and transport, and two-phase fluid flow within these deposits. Our new model features a reformed assumption of 2D, two-phase film boiling within the CRUD, correcting earlier models' assumptions of single-phase coolant flow with wick boiling under high heat fluxes. This model helps to better explain observed experimental values of the effective CRUD thermal conductivity. Finally, we propose a more complete set of boiling regimes, or a more detailed mechanism, to explain recent CRUD deposition experiments by suggesting the new concept of double dryout specifically in thick porous media with boiling chimneys. - Highlights: • A two-phase model of CRUD's effects on fuel cladding is developed and improved. • This model eliminates the formerly erroneous assumption of wick boiling. • Higher fuel cladding temperatures are predicted when accounting for two-phase flow. • Double-peaks in thermal conductivity vs. heat flux in experiments are explained. • A “double dryout” mechanism in CRUD is proposed based on the model and experiments.

  19. CORROSION RESISTANCE OF ORGANOMETALLIC COATING APLICATED IN FUEL TANKS USING ELECTROCHEMICAL IMPEDANCE SPECTROSCOPY IN BIOFUEL – PART I

    Directory of Open Access Journals (Sweden)

    Milene Adriane Luciano

    2014-10-01

    Full Text Available Nowadays, the industry has opted for more sustainable production processes, and the planet has also opted for new energy sources. From this perspective, automotive tanks with organometallic coatings as well as a partial substitution of fossil fuels by biofuels have been developed. These organometallic coated tanks have a zinc layer, deposited by a galvanizing process, formed between the steel and the organometallic coating. This work aims to characterize the organometallic coating used in metal automotive tanks and evaluate their corrosion resistance in contact with hydrated ethyl alcohol fuel (AEHC. For this purpose, the resistance of all layers formed between Zinc and EEP steel and also the tin coated steel, which has been used for over thirty years, were evaluated. The technique chosen was the Electrochemical Impedance Spectroscopy. The results indicated an increase on the corrosion resistance when organometallic coatings are used in AEHC medium. In addition to that, these coatings allow an estimated 25% reduction in tanks production costs.

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

    Directory of Open Access Journals (Sweden)

    Parikin

    2003-08-01

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

  1. Corrosion of high temperature resisting alloys exposed to heavy fuel ash; Corrosion de aleaciones resistentes a altas temperaturas expuestas a ceniza de combustoleo pesado

    Energy Technology Data Exchange (ETDEWEB)

    Wong Moreno, Adriana del Carmen

    1998-03-01

    The objective of the performed research was to study the degradation process by high temperature corrosion of alloys exposed to heavy fuel oil ashes through a comparative experimental evaluation of its performance that allowed to establish the mechanisms involved in the phenomenon. The experimentation carried out involved the determination of the resistance to the corrosion of 14 alloys of different type (low and medium alloy steels, ferritic and austenitic stainless steels, nickel base alloys and a FeCrAl alloy of type ODS) exposed to high temperatures (580 Celsius degrees - 900 Celsius degrees) in 15 ash deposits with different corrosive potential, which were collected in the high temperature zone of boilers of thermoelectric power stations. The later studies to the corrosion tests consisted of the analysis by sweeping electron microscopy supported by microanalysis of the corroded probes, with the purpose of determining the effect of Na, V and S on the corrosivity of the ash deposits and the effect of the main alloying elements on the corrosion resistance of the alloys. Such effects are widely documented to support the proposed mechanisms of degradation that are occurring. The global analysis of the generated results has allowed to propose a model to explain the global mechanism of corrosion of alloys exposed to the high temperatures of ash deposits. The proposed model, complements the processed one by Wilson, widely accepted for fused vanadates, as far as on one hand, it considers the effect of the sodium sulfate presence (in addition to the vanadium compounds) in the deposits, and on the other hand, it extends it to temperatures higher than the point of fusion of constituent vanadium compounds of the deposits. Both aspects involve considering the roll that the process of diffusion of species has on the degradation and the capacity of protection of the alloy. The research performed allowed to confirm what the Wilson model had established for deposits with high

  2. A model to describe the anisotropic viscoplastic mechanical behavior of fresh and irradiated Zircaloy-4 fuel claddings under RIA loading conditions

    Science.gov (United States)

    Le Saux, M.; Besson, J.; Carassou, S.; Poussard, C.; Averty, X.

    2008-08-01

    This paper presents a unified phenomenological model to describe the anisotropic viscoplastic mechanical behavior of cold-worked stress relieved (CWSR) Zircaloy-4 fuel claddings submitted to reactivity initiated accident (RIA) loading conditions. The model relies on a multiplicative viscoplastic formulation and reproduces strain hardening, strain rate sensitivity and plastic anisotropy of the material. It includes temperature, fluence and irradiation conditions dependences within RIA typical ranges. Model parameters have been tuned using axial tensile, hoop tensile and closed-end internal pressurization tests results essentially obtained from the PROMETRA program, dedicated to the study of zirconium alloys under RIA loading conditions. Once calibrated, the model provides a reliable description of the mechanical behavior of the fresh and irradiated (fluence up to 10×1025 nm or burnup up to 64 GWd/tU) material within large temperature (from 20 °C up to 1100 °C) and strain rate ranges (from 3×10-4 s up to 5 s), representative of the RIA spectrum. Finally, the model is used for the finite element analysis of the hoop tensile tests performed within the PROMETRA program.

  3. Laser cladding of Al + Ir powders on ZM5 magnesium base alloy

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Laser cladding of preplaced Al + Ir powders on a ZM5 magnesium alloy was performed to enhance the corrosion resistance of the ZM5 magnesium alloy. A metallurgical bond was obtained at the coating/substrate interface. The corrosion potential (Ecorr) of the laser cladded sample was 169 mV positive to that of the untreated ZM5 substrate, while the corrosion current (Icorr) was some one order of magnitude lower. The laser cladded sample, unlike the untreated ZM5 substrate,showed a passive region in the polarization plot. Immersion tests confirmed that the corrosion resistance of the laser cladded ZM5 sample was significantly enhanced in 3.5 wt.% NaCl solution. The Al-rich phases of AlIr, Mg17Al12, and Al formed in the cladding layer and the rapid solid characteristics were contributed to the improved corrosion behavior of the coating.

  4. Correlation between General Corrosion Behavior and Eddy Current Noise of Alloy 690 Steam Generator Tube

    Energy Technology Data Exchange (ETDEWEB)

    Hur, Do Haeng; Choi, Myung Sik; Lee, Deok Hyun; Shim, Hee-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Nickel and its oxides are released from the surface of steam generator tubes into the primary water. Released nickel and cobalt is activated to Co-58 and Co-60 in the reactor core by a neutron flux, respectively. These activated corrosion products are the main source of high radiation fields and occupational radiation exposure. In addition, some of the corrosion products redeposit on the fuel cladding, hinder the heat transfer, increase the corrosion rate of the fuel cladding, and finally induce an axial offset anomaly. This phenomenon can decrease core shutdown margin, and thus lead to a down-rating of a plant. Recently, many researchers have reported that the surface states of Alloy 690 tubes affect the corrosion product formation and its release in simulated primary water environments. Meanwhile, the surface states of steam generator tubes affect the noise level of eddy current testing. Noise signals arising from the tubes degrade the probability of detection and sizing accuracy of the defects. The corrosion behavior was closely correlated to the tube noise measured using a rotating probe, while it was not related to the noise measured using a bobbin probe. It is suggested that the tube noise value measured using a rotating pancake coil probe can be a decisive measure to estimate the corrosion behavior of tubing.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-13

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

  6. Clad Degradation - FEPs Screening Arguments

    Energy Technology Data Exchange (ETDEWEB)

    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]).

  7. Corrosão metálica associada ao uso de combustíveis minerais e biocombustíveis Metallic corrosion related to mineral fuels and biofuels utilization

    Directory of Open Access Journals (Sweden)

    Alessandra Regina Pepe Ambrozin

    2009-01-01

    Full Text Available Fuels and biofuels have a major importance in the transportation sector of any country, contributing to their economic development. The utilization of these fuels implies their closer contact to metallic materials, which comprise vehicle, storage, and transportation systems. Thus, metallic corrosion could be related to fuels and biofuels utilization. Specially, the corrosion associated to gasoline, ethanol, diesel, biodiesel, and their mixtures is discussed in this article. Briefly, the ethanol is the most corrosive and gasoline the least. Few investigations about the effect of biodiesel indicate that the corrosion is associated to their unsaturation degree and the corrosion of diesel is related to its acidity.

  8. Study of corrosion of aluminium alloys of nuclear purity in ordinary water, пart one

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2004-01-01

    Full Text Available Effects of corrosion of aluminum alloys of nuclear purity in ordinary water of the spent fuel storage pool of the RA research reactor at VINČA Institute of Nuclear Sciences has been examined in the frame work of the International Atomic Energy Agency Coordinated Research Project "Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water" since 2002. The study presented in this paper comprises activities on determination and monitoring of chemical parameters and radio activity of water and sludge in the RA spent fuel storage pool and results of the initial study of corrosion effects obtained by visual examinations of surfaces of various coupons made of aluminum alloys of nuclear purity of the test racks exposed to the pool water for a period from six months to six years.

  9. Electrochemical corrosion studies of carbon supports and electrocatalysts and their effects on the durability of low-temperature PEM fuel cells

    Science.gov (United States)

    Dowlapalli, Madhusudhana R.

    Performance of a PEM fuel cell relies heavily on the durability of the platinum and platinum-alloy based electrocatalysts supported on carbon blacks. Carbon corrosion has been widely accepted as an important issue affecting the degradation of the catalytic layer in PEMFCs. Traditional carbon blacks used in today's fuel cell industry are not tailored to suit the corrosive conditions encountered in PEMFCs. Advanced carbon supports should have excellent electrochemical corrosion resistance, good conductivity, high surface area and optimum hydrophilic properties. The principal objective of this work is to investigate the corrosive behavior of carbon blacks and electrocatalysts supported on such carbon blacks in conditions that are typical for fuel cells. Physical and chemical changes during oxidation of these carbon blacks have been reviewed along with methodology for studying their corrosion in a low-temperature fuel cell environment. This study provides an ex-situ corrosion measurement protocol and a gas diffusion electrode half-cell setup to study the electrochemical oxidation resistance behavior of standard carbon blacks, modified carbon blacks, and advanced carbon supports in acid electrolyte at 25°C. Corrosion current-time relationships were evaluated and transient mode of corrosion study was employed to simulate automobile startup/shutdown. The effects of various surface modifications on carbon corrosion behavior have been studied in detail. The aggravated corrosion of carbon supports at potentials higher than the thermodynamic stable regime of water was investigated and a mechanism is proposed to address the same. The role of the metal phase on carbon corrosion at the catalyst-support interphase has also been investigated. The corrosion current dependence on the microstructure and nature of surface groups present on these carbons was examined. Further, measuring carbon corrosion effects on the durability of a single membrane-electrode assembly (MEA) cathode

  10. Multidimensional simulations of hydrides during fuel rod lifecycle

    Science.gov (United States)

    Stafford, D. S.

    2015-11-01

    In light water reactor fuel rods, waterside corrosion of zirconium-alloy cladding introduces hydrogen into the cladding, where it is slightly soluble. When the solubility limit is reached, the hydrogen precipitates into crystals of zirconium hydride which decrease the ductility of the cladding and may lead to cladding failure during dry storage or transportation events. The distribution of the hydride phase and the orientation of the crystals depend on the history of the spatial temperature and stress profiles in the cladding. In this work, we have expanded the existing hydride modeling capability in the BISON fuel performance code with the goal of predicting both global and local effects on the radial, azimuthal and axial distribution of the hydride phase. We compare results from 1D simulations to published experimental data. We demonstrate the new capability by simulating in 2D a fuel rod throughout a lifecycle that includes irradiation, short-term storage in the spent fuel pool, drying, and interim storage in a dry cask. Using the 2D simulations, we present qualitative predictions of the effects of the inter-pellet gap and the drying conditions on the growth of a hydride rim.

  11. Corrosion Behaviour of Mg Alloys in Various Basic Media: Application of Waste Encapsulation of Fuel Decanning from UNGG Nuclear Reactor

    Science.gov (United States)

    Lambertin, David; Frizon, Fabien; Blachere, Adrien; Bart, Florence

    The dismantling of UNGG nuclear reactor generates a large volume of fuel decanning. These materials are based on Mg-Zr alloy. The dismantling strategy could be to encapsulate these wastes into an ordinary Portland cement (OPC) or geopolymer (aluminosilicate material) in a form suitable for storage. Studies have been performed on Mg or Mg-Al alloy in basic media but no data are available on Mg-Zr behaviour. The influence of representative pore solution of both OPC and geopolymer with Mg-Zr alloy has been studied on corrosion behaviour. Electrochemical methods have been used to determine the corrosion densities at room temperature. Results show that the corrosion densities of Mg-Zr alloy in OPC solution is one order of magnitude more important than in a geopolymer solution environment and the effect of an inhibiting agent has been undertaken with Mg-Zr alloy. Evaluation of corrosion hydrogen production during the encapsulation of Mg-Zr alloy in both OPC and geopolymer has also been done.

  12. Countermeasures to corrosion on water walls. Part 2; Aatgaerder mot eldstadskorrosion paa panntuber. Etapp 2

    Energy Technology Data Exchange (ETDEWEB)

    Storesund, Jan; Elger, Ragna; Nordling, Magnus; Viklund, Peter

    2011-01-15

    Background: The problems with water wall corrosion have been accelerating over the last years. There are a number of reasons for this. Originally mild steels were successfully used in power plant water walls. The magnetite layer that forms at the fire side of the tubes when the boiler is taken into operation protected from corrosion attack. The fuels at that time (oil, coal, gas) were not able to break down the magnetite by corrosion. In addition, there were no restrictions for pollutions and for the combustion itself that could contribute to corrosion attack. The usage of fossil fuels has decreased substantially over the last 25 years, not least by environmental reasons. As a replacement a number of different kinds of bio mass fuels are used. These are typically more or less corrosive and the magnetite layers are attacked. The corrosion is often supported by reducing conditions as a result of the restrictions of the NO{sub x}-pollution. Also the waste fuelled boilers have huge corrosion problems. This has been the case for the last 25 years but nowadays the number of such plants is so much higher and the service data have been turned up. Corrosion protection of the water wall tubes started to be successful in the beginning of the seventies by the introduction of the composite tube. Such tubes are fabricated by mild steel or a low alloy core and corrosion resistant austenite steel or nickel base as an about 2 mm thick corrosion protective coating. Weld cladding of the water wall tubes was introduced in the 1980's as a significantly cheaper alternative to the composite tubes. Thermal spraying and refractory protection are other methods. These corrosion protection methods have not always been effective. For example, depending on incorrect materials selection, incorrect performance and incorrect method selection for the current corrosion or erosion attack. Therefore, there is a need for increased knowledge of which protection method and material that will work

  13. Countermeasures to corrosion on water walls. Part 2; Aatgaerder mot eldstadskorrosion paa panntuber. Etapp 2

    Energy Technology Data Exchange (ETDEWEB)

    Storesund, Jan; Elger, Ragna; Nordling, Magnus; Viklund, Peter

    2011-01-15

    Background: The problems with water wall corrosion have been accelerating over the last years. There are a number of reasons for this. Originally mild steels were successfully used in power plant water walls. The magnetite layer that forms at the fire side of the tubes when the boiler is taken into operation protected from corrosion attack. The fuels at that time (oil, coal, gas) were not able to break down the magnetite by corrosion. In addition, there were no restrictions for pollutions and for the combustion itself that could contribute to corrosion attack. The usage of fossil fuels has decreased substantially over the last 25 years, not least by environmental reasons. As a replacement a number of different kinds of bio mass fuels are used. These are typically more or less corrosive and the magnetite layers are attacked. The corrosion is often supported by reducing conditions as a result of the restrictions of the NO{sub x}-pollution. Also the waste fuelled boilers have huge corrosion problems. This has been the case for the last 25 years but nowadays the number of such plants is so much higher and the service data have been turned up. Corrosion protection of the water wall tubes started to be successful in the beginning of the seventies by the introduction of the composite tube. Such tubes are fabricated by mild steel or a low alloy core and corrosion resistant austenite steel or nickel base as an about 2 mm thick corrosion protective coating. Weld cladding of the water wall tubes was introduced in the 1980's as a significantly cheaper alternative to the composite tubes. Thermal spraying and refractory protection are other methods. These corrosion protection methods have not always been effective. For example, depending on incorrect materials selection, incorrect performance and incorrect method selection for the current corrosion or erosion attack. Therefore, there is a need for increased knowledge of which protection method and material that will work

  14. FY2004 CORROSION SURVEILLANCE RESULTS FOR L-BASIN

    Energy Technology Data Exchange (ETDEWEB)

    VORMELKER, P

    2005-09-05

    This report documents the results of the L-Basin Corrosion Surveillance Program for the fiscal year 2004. Test coupons were removed from the basin on February 12, 2004, shipped to Savannah River National Laboratory (SRNL), and visually examined in a contaminated laboratory hood. Selected coupons were metallurgically characterized to establish the extent of general corrosion and pitting. Pitting was observed on galvanically coupled and on intentionally creviced coupons, thus demonstrating that localized concentration cells were formed during the exposure period. In these cases, the susceptibility to pitting was not attributed to aggressive basin water chemistry but to localized conditions (intentional crevices and galvanic coupling) that allowed the development of oxygen and/or metal ion concentration cells that produced locally aggressive waters. General oxidation was also observed on all of the coupons with localized corrosion observed on some of the coupons. These coupons were not pretreated to produce a protective oxide layer prior to exposure in the basin water. Non-protected coupons are more susceptible to corrosion than fuel cladding which has developed a protective oxide layer from high temperature reactor operations. However, the oxide on spent nuclear fuel (SNF) stored in L-Basin is not necessarily in pristine condition. Some of the oxide may have spalled off or been mechanically damaged prior to arrival at SRS. These areas on the fuel cladding would have the same susceptibility to corrosion as the coupons. Current observations from the test coupons demonstrate that, even with rigorously controlled basin water chemistry, localized aggressive conditions can develop in intentional crevice and galvanic samples. These results do illustrate the potential for corrosion induced degradation and thus the importance of a routine surveillance program similar to that conducted on the Uruguay fuel and on the surveillance coupons stored in L-Basin and future in

  15. MAX Phase Modified SiC Composites for Ceramic-Metal Hybrid Cladding Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yang-Il; Kim, Sun-Han; Park, Dong-Jun; Park, Jeong-Hwan; Park, Jeong-Yong; Kim, Hyun-Gil; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A metal-ceramic hybrid cladding consists of an inner zirconium tube, and an outer SiC fiber-matrix SiC ceramic composite with surface coating as shown in Fig. 1 (left-hand side). The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. In addition, the outermost layer prevents the dissolution of SiC during normal operation. On the other hand, a ceramic-metal hybrid cladding consists of an outer zirconium tube, and an inner SiC ceramic composite as shown in Fig. 1 (right-hand side). The outer zirconium protects the fuel rod from a corrosion during reactor operation, as in the present fuel claddings. The inner SiC composite, additionally, is designed to resist the severe oxidation under a postulated accident condition of a high-temperature steam environment. Reaction-bonded SiC was fabricated by modifying the matrix as the MAX phase. The formation of Ti{sub 3}SiC{sub 2} was investigated depending on the compositions of the preform and melt. In most cases, TiSi{sub 2} was the preferential phase because of its lowest melting point in the Ti-Si-C system. The evidence of Ti{sub 3}SiC{sub 2} was the connection with the pressurizing.

  16. High temperature corrosion of metallic interconnects in solid oxide fuel cells

    Directory of Open Access Journals (Sweden)

    Bastidas, D. M.

    2006-12-01

    Full Text Available Research and development has made it possible to use metallic interconnects in solid oxide fuel cells (SOFC instead of ceramic materials. The use of metallic interconnects was formerly hindered by the high operating temperature, which made the interconnect degrade too much and too fast to be an efficient alternative. When the operating temperature was lowered, the use of metallic interconnects proved to be favourable since they are easier and cheaper to produce than ceramic interconnects. However, metallic interconnects continue to be degraded despite the lowered temperature, and their corrosion products contribute to electrical degradation in the fuel cell. Coatings of nickel, chromium, aluminium, zinc, manganese, yttrium or lanthanum between the interconnect and the electrodes reduce this degradation during operation

    El uso de interconectores metálicos en pilas de combustible de óxido sólido (SOFC en sustitución de materiales cerámicos ha sido posible gracias a la investigación y desarrollo de nuevos materiales metálicos. Inicialmente, el uso de interconectores metálicos fue limitado, debido a la elevada temperatura de trabajo, ocasionando de forma rápida la degradación del material, lo que impedía que fuesen una alternativa. A medida que la temperatura de trabajo de las SOFC descendió, el uso de interconectores metálicos demostró ser una buena alternativa, dado que son más fáciles de fabricar y más baratos que los interconectores cerámicos. Sin embargo, los interconectores metálicos continúan degradándose a pesar de descender la temperatura a la que operan las SOFC y, asimismo, los productos de corrosión favorecen las pérdidas eléctricas de la pila de combustible. Recubrimientos de níquel, cromo, aluminio, zinc, manganeso, itrio y lantano entre el interconector y los electrodos reduce dichas pérdidas eléctricas.

  17. Corrosion of Zircaloy-2 by pH 10 LiOH in heated crevices

    Energy Technology Data Exchange (ETDEWEB)

    Dickinson, D.R.

    1963-11-12

    Both the inner and outer tubes of the N-Reactor fuel elements will have self supports spot welded to the lateral heat-transfer surface of the element. A crevice a few mils thick will exist around the weld between the support tab and the cladding. Because of the heat flux through the cladding at this point and the insulating effect of the support tab, the temperature in this crevice will be higher than that on the free surface away from the support. This can result in boiling in the crevice leading to concentration of LiOH (or impurities in the water) to a level where it can cause severe corrosion of the Zircaloy-2 cladding. The tests described in this report were conducted to determine whether such attack might be encountered in N-Reactor.

  18. Effects of environmental factors on corrosion behaviors of metal-fiber porous components in a simulated direct methanol fuel cell environment

    Institute of Scientific and Technical Information of China (English)

    Wei Yuan; Bo Zhou; Yong Tang; Zhao-chun Zhang; Jun Deng

    2014-01-01

    To enable the use of metallic components in direct methanol fuel cells (DMFCs), issues related to corrosion resistance must be considered because of an acid environment induced by the solid electrolyte. In this study, we report the electrochemical behaviors of metal-fiber-based porous sintered components in a simulated corrosive environment of DMFCs. Three materials were evaluated:pure copper, AISI304, and AISI316L. The environmental factors and related mechanisms affecting the corrosion behaviors were analyzed. The results demonstrated that AISI316L exhibits the best performance. A higher SO42-concentration increases the risk of material corrosion, whereas an increase in methanol concentration inhibits corrosion. The morphological features of the corroded samples were also characterized in this study.

  19. Multilayer graphene for long-term corrosion protection of stainless steel bipolar plates for polymer electrolyte membrane fuel cell

    DEFF Research Database (Denmark)

    Stoot, Adam Carsten; Camilli, Luca; Spiegelhauer, Susie Ann

    2015-01-01

    /SS, both samples exhibiting a similar trend, thus questioning the short-term positive effect of graphene coatings. However, partial immersion in boiling seawater for three weeks reveals a clear superiority of the graphene coating with respect to steel just protected by Ni. After the test, the graphene film......Abstract Motivated by similar investigations recently published (Pu et al., 2015), we report a comparative corrosion study of three sets of samples relevant as bipolar plates for polymer electrolyte fuel cells: stainless steel, stainless steel with a nickel seed layer (Ni/SS) and stainless steel...

  20. Extension and assessment of the cladding ballooning model in the FRAP-T6 code

    Energy Technology Data Exchange (ETDEWEB)

    El-Adham, K

    1987-05-01

    The FRAP-T6 code was extended to calculate: (1) fuel surface azimuthal temperature distribution; (2) work done on cladding by internal pressure; and (3) azimuthal heat conduction in the cladding. The extensions were assessed by comparing calculated and measured cladding ballooning characteristics for four in-pile fuel rod tests. The assessment showed that the calculation of the fuel surface azimuthal temperature distribution improved the calculations of cladding ballooning. Both calculations and experimental results indicate that coplanar blockage due to cladding ballooning is unlikely during a large break LOCA.

  1. ;Study of secondary hydriding at high temperature in zirconium based nuclear fuel cladding tubes by coupling information from neutron radiography/tomography, electron probe micro analysis, micro elastic recoil detection analysis and laser induced breakdown spectroscopy microprobe

    Science.gov (United States)

    Brachet, Jean-Christophe; Hamon, Didier; Le Saux, Matthieu; Vandenberghe, Valérie; Toffolon-Masclet, Caroline; Rouesne, Elodie; Urvoy, Stéphane; Béchade, Jean-Luc; Raepsaet, Caroline; Lacour, Jean-Luc; Bayon, Guy; Ott, Frédéric

    2017-05-01

    This paper gives an overview of a multi-scale experimental study of the secondary hydriding phenomena that can occur in nuclear fuel cladding materials exposed to steam at high temperature (HT) after having burst (loss-of-coolant accident conditions). By coupling information from several facilities, including neutron radiography/tomography, electron probe micro analysis, micro elastic recoil detection analysis and micro laser induced breakdown spectroscopy, it was possible to map quantitatively, at different scales, the distribution of oxygen and hydrogen within M5™ clad segments having experienced ballooning and burst at HT followed by steam oxidation at 1100 and 1200 °C and final direct water quenching down to room temperature. The results were very reproducible and it was confirmed that internal oxidation and secondary hydriding at HT of a cladding after burst can lead to strong axial and azimuthal gradients of hydrogen and oxygen concentrations, reaching 3000-4000 wt ppm and 1.0-1.2 wt% respectively within the β phase layer for the investigated conditions. Consistent with thermodynamic and kinetics considerations, oxygen diffusion into the prior-β layer was enhanced in the regions highly enriched in hydrogen, where the α(O) phase layer is thinner and the prior-β layer thicker. Finally the induced post-quenching hardening of the prior-β layer was mainly related to the local oxygen enrichment. Hardening directly induced by hydrogen was much less significant.

  2. Multiphysics modeling of two-phase film boiling within porous corrosion deposits

    Science.gov (United States)

    Jin, Miaomiao; Short, Michael

    2016-07-01

    Porous corrosion deposits on nuclear fuel cladding, known as CRUD, can cause multiple operational problems in light water reactors (LWRs). CRUD can cause accelerated corrosion of the fuel cladding, increase radiation fields and hence greater exposure risk to plant workers once activated, and induce a downward axial power shift causing an imbalance in core power distribution. In order to facilitate a better understanding of CRUD's effects, such as localized high cladding surface temperatures related to accelerated corrosion rates, we describe an improved, fully-coupled, multiphysics model to simulate heat transfer, chemical reactions and transport, and two-phase fluid flow within these deposits. Our new model features a reformed assumption of 2D, two-phase film boiling within the CRUD, correcting earlier models' assumptions of single-phase coolant flow with wick boiling under high heat fluxes. This model helps to better explain observed experimental values of the effective CRUD thermal conductivity. Finally, we propose a more complete set of boiling regimes, or a more detailed mechanism, to explain recent CRUD deposition experiments by suggesting the new concept of double dryout specifically in thick porous media with boiling chimneys.

  3. Fuel assembly reconstitution

    Energy Technology Data Exchange (ETDEWEB)

    Morgado, Mario M.; Oliveira, Monica G.N.; Ferreira Junior, Decio B.M.; Santos, Barbara O. dos; Santos, Jorge E. dos, E-mail: mongeor@eletronuclear.gov.b [ELETROBRAS Termonuclear S.A. - ELETRONUCLEAR, Angra dos Reis, RJ (Brazil)

    2009-07-01

    Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

  4. Solid-phase zirconium and fluoride species in alkaline zircaloy cladding waste at Hanford.

    Science.gov (United States)

    Reynolds, Jacob G; Huber, Heinz J; Cooke, Gary A; Pestovich, John A

    2014-08-15

    The United States Department of Energy Hanford Site, near Richland, Washington, USA, processed plutonium between 1944 and 1987. Fifty-six million gallons of waste of various origins remain, including waste from removing zircaloy fuel cladding using the so-called Zirflex process. The speciation of zirconium and fluoride in this waste is important because of the corrosivity and reactivity of fluoride as well as the (potentially) high density of Zr-phases. This study evaluates the solid-phase speciation of zirconium and fluoride using X-ray diffraction (XRD) and scanning electron microscopy with energy dispersive spectroscopy (SEM-EDS). Two waste samples were analyzed: one waste sample that is relatively pure zirconium cladding waste from tank 241-AW-105 and another that is a blend of zirconium cladding wastes and other high-level wastes from tank 241-C-104. Villiaumite (NaF) was found to be the dominant fluoride species in the cladding waste and natrophosphate (Na7F[PO4]2 · 19H2O) was the dominant species in the blended waste. Most zirconium was present as a sub-micron amorphous Na-Zr-O phase in the cladding waste and a Na-Al-Zr-O phase in the blended waste. Some zirconium was present in both tanks as either rounded or elongated crystalline needles of Na-bearing ZrO2 that are up to 200 μm in length. These results provide waste process planners the speciation data needed to develop disposal processes for this waste.

  5. Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

    Science.gov (United States)

    Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

    2016-09-01

    Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) or plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

  6. Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

    Science.gov (United States)

    Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

    2016-11-01

    Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

  7. Preliminary Investigation of Candidate Materials for Use in Accident Resistant Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jason M. Harp; Paul A. Lessing; Blair H. Park; Jakeob Maupin

    2013-09-01

    As part of a Collaborative Research and Development Agreement (CRADA) with industry, Idaho National Laboratory (INL) is investigating several options for accident resistant uranium compounds including silicides, and nitrides for use in future light water reactor (LWR) fuels. This work is part of a larger effort to create accident tolerant fuel forms where changes to the fuel pellets, cladding, and cladding treatment are considered. The goal fuel form should have a resistance to water corrosion comparable to UO2, have an equal to or larger thermal conductivity than uranium dioxide, a melting temperature that allows the material to stay solid under power reactor conditions, and a uranium loading that maintains or improves current LWR power densities. During the course of this research, fuel fabricated at INL will be characterized, irradiated at the INL Advanced Test Reactor, and examined after irradiation at INL facilities to help inform industrial partners on candidate technologies.

  8. Polydopamine as a promising candidate for the design of high performance and corrosion-tolerant polymer electrolyte fuel cell electrodes

    Science.gov (United States)

    Long, Hongtao; Del Frari, Doriane; Martin, Arnaud; Didierjean, Joffrey; Ball, Vincent; Michel, Marc; Ahrach, Hicham Ibn El

    2016-03-01

    Carbon materials such as carbon black or nanotubes suffer from degradation when subjected to harsh conditions occurring in a Polymer Electrolyte Membrane Fuel Cells (PEMFCs) electrode. Hence, nowadays it is more and more important to search for alternative support materials. The present work shows the results for the incorporation of alternative materials into PEMFCs electrode architectures. Commercially available Multi-Walled NanoTubes (MWNTs) are used as a support for Pt nanoparticles in combination with Polydopamine (PDA). The role of MWNTs is to confer a high electronic conductivity and help to form a porous network. On the other side the role of polydopamine is both to promote the proton conductivity similarly to ionomers such as Nafion and to protect the MWNTs against corrosion. The fuel cell polarization test shows a maximum power density of 780 mW cm-2 and a Pt utilization of 6051 mW mg(Pt)-1. The Pt utilization reached in this work is almost three times higher than for Pt/MWNTs electrodes containing the same Pt loading. Beside this, it is also shown for the first time that PDA serves as protective layer against carbon corrosion.

  9. Mitsubishi PWR nuclear fuel with advanced design features

    Energy Technology Data Exchange (ETDEWEB)

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2008-10-15

    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Stern, A

    2007-12-15

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

  11. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

    2004-12-01

    This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

  12. Characteristic test technology for PWR fuel and its components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho; Jeong, Yong Hwan; Park, Sang Yoon; Kim, Kyeng Ho; Nam, Cheol; Baek, Jong Hyuk; Lee, Myung Ho; Choi, Byoung Kwon; Song, Kun Woo; Kang, Ki Won; Kim, Keon Sik; Kim, Jong Hun; Kim, Young Min; Yang, Jae Ho; Song, Kee Nam; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Chun, Tae Hyun; In, Wang Kee; Oh, Dong Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-01-01

    Characteristic tests of fuel assembly and its components being developed in the Advanced LWR Fuel Development Project supported by the mid-long term nuclear R and D program are described in this report. Performance verification of fuel and its components by the characteristic tests are essential to their development. Fuel components being developed in the Advanced LWR Fuel Development Project are zirconium alloy cladding, UO{sub 2} and burnable absorber pellets, spacer grid and top and bottom end pieces. Detailed test plans for those fuel components are described in this report, and test procedures of cladding and pellet are also described in the Appendix. Examples of the described tests are in- and out-of- pile corrosion and mechanical tests such as creep and burst tests for the cladding, in-pile capsule and ramp tests for the pellet, mechanical tests such as strength and vibration, and thermal-hydraulic tests such as pressure drop and critical heat flux for the spacer grid and top and bottom end pieces. It is expected that this report could be used as the standard reference for the performance verification tests in the development of LWR fuel and its components. 11 refs., 9 figs., 2 tabs. (Author)

  13. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Corrosion behavior of Fe-Si metallic coatings added with NiCrAlY in an environment of fuel oil ashes at 700 C

    Energy Technology Data Exchange (ETDEWEB)

    Salinas-Bravo, V.M.; Porcayo-Calderon, J.; Romero-Castanon, T. [Instituto de Investigaciones Electricas, Gerencia de Procesos Termicos., Av. Reforma 113, C.P. 62490 Col. Palmira. Temixco. Morelos (Mexico); Dominguez-Patino, G.; Gonzalez-Rodriguez, J.G. [U.A.E.M. Centro de Investigaciones en Ingenieria y Ciencias Aplicadas., Av. Universidad 1001, C.P. 62210, Col. Chamilpa. Cuernavaca, Morelos (Mexico)

    2005-07-01

    Electrochemical potentiodynamic polarization curves and immersion tests for 300 h at 700 C in a furnace have been used to evaluate the corrosion resistance of Fe-Si metallic coatings added with up to 50 wt.% of NiCrAIY. The corrosive environment was fuel oil ashes from a steam generator. The composition of fuel oil ashes includes high content of vanadium, sodium and sulfur. The results obtained show that only the addition of 20 wt.% NiCrAlY to the Fe-Si coating improves its corrosion resistance. The behavior of all tested coatings is explained by the results obtained from the analysis of every coating using electron microscopy and energy dispersive X-ray analysis. (Abstract Copyright [2005], Wiley Periodicals, Inc.)

  15. Modeling of Membrane-Electrode-Assembly Degradation in Proton-Exchange-Membrane Fuel Cells - Local H2 Starvation and Start-Stop Induced Carbon-Support Corrosion

    Science.gov (United States)

    Gu, Wenbin; Yu, Paul T.; Carter, Robert N.; Makharia, Rohit; Gasteiger, Hubert A.

    Carbon-support corrosion causes electrode structure damage and thus electrode degradation. This chapter discusses fundamental models developed to predict cathode carbon-support corrosion induced by local H2 starvation and start-stop in a proton-exchange-membrane (PEM) fuel cell. Kinetic models based on the balance of current among the various electrode reactions are illustrative, yielding much insight on the origin of carbon corrosion and its implications for future materials developments. They are particularly useful in assessing carbon corrosion rates at a quasi-steady-state when an H2-rich region serves as a power source that drives an H2-free region as a load. Coupled kinetic and transport models are essential in predicting when local H2 starvation occurs and how it affects the carbon corrosion rate. They are specifically needed to estimate length scales at which H2 will be depleted and time scales that are valuable for developing mitigation strategies. To predict carbon-support loss distributions over an entire active area, incorporating the electrode pseudo-capacitance appears necessary for situations with shorter residence times such as start-stop events. As carbon-support corrosion is observed under normal transient operations, further model improvement shall be focused on finding the carbon corrosion kinetics associated with voltage cycling and incorporating mechanisms that can quantify voltage decay with carbon-support loss.

  16. The effect of post-treatment of a high-velocity oxy-fuel Ni-Cr-Mo-Si-B coating part 2: Erosion-corrosion behavior

    Science.gov (United States)

    Shrestha, S.; Hodgkiess, T.; Neville, A.

    2001-12-01

    In this paper, a study of the erosion-corrosion characteristics of a Ni-Cr-Mo-Si-B coating applied by the high-velocity oxy-fuel (HVOF) process on to an austenitic stainless steel (UNS S31603) substrate are reported. The coatings were studied in the as-sprayed condition, after vacuum sealing with polymer impregnation and after vacuum furnace fusion. The erosion-corrosion characteristics were assessed in an impinging liquid jet of 3.5% NaCl solution at 18 °C at a velocity of 17 m/s at normal incidence in two conditions: (1) free from added solids and (2) containing 800 ppm silica sand. The methodology employed electrochemical control and monitoring to facilitate the identification of the separate and interrelated erosion and corrosion contributions to the erosion-corrosion process. The rates of erosion-corrosion damage were drastically accelerated in the presence of the suspended solids. The application of cathodic protection significantly reduced the deterioration process. The study showed the effect of sealing with polymer impregnation did not significantly alter the erosion-corrosion behavior of the sprayed coating. However, there was a significant improvement in erosion-corrosion durability afforded by the postfusion process. The mechanisms by which the improved performance of vacuum-fused coatings is achieved are discussed.

  17. Development of models and online diagnostic monitors of the high-temperature corrosion of refractories in oxy/fuel glass furnaces : final project report.

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, Stewart K.; Gupta, Amul (Monofrax Inc., Falconer, NY); Walsh, Peter M.; Rice, Steven F.; Velez, Mariano (University of Missouri, Rolla, MO); Allendorf, Mark D.; Pecoraro, George A. (PPG Industries, Inc., Pittsburgh, PA); Nilson, Robert H.; Wolfe, H. Edward (ANH Refractories, Pittsburgh, PA); Yang, Nancy Y. C.; Bugeat, Benjamin () American Air Liquide, Countryside, IL); Spear, Karl E. (Pennsylvania State University, University Park, PA); Marin, Ovidiu () American Air Liquide, Countryside, IL); Ghani, M. Usman (American Air Liquide, Countryside, IL)

    2005-02-01

    This report summarizes the results of a five-year effort to understand the mechanisms and develop models that predict the corrosion of refractories in oxygen-fuel glass-melting furnaces. Thermodynamic data for the Si-O-(Na or K) and Al-O-(Na or K) systems are reported, allowing equilibrium calculations to be performed to evaluate corrosion of silica- and alumina-based refractories under typical furnace operating conditions. A detailed analysis of processes contributing to corrosion is also presented. Using this analysis, a model of the corrosion process was developed and used to predict corrosion rates in an actual industrial glass furnace. The rate-limiting process is most likely the transport of NaOH(gas) through the mass-transport boundary layer from the furnace atmosphere to the crown surface. Corrosion rates predicted on this basis are in better agreement with observation than those produced by any other mechanism, although the absolute values are highly sensitive to the crown temperature and the NaOH(gas) concentration at equilibrium and at the edge of the boundary layer. Finally, the project explored the development of excimer laser induced fragmentation (ELIF) fluorescence spectroscopy for the detection of gas-phase alkali hydroxides (e.g., NaOH) that are predicted to be the key species causing accelerated corrosion in these furnaces. The development of ELIF and the construction of field-portable instrumentation for glass furnace applications are reported and the method is shown to be effective in industrial settings.

  18. Influence of microstructure on hydrothermal corrosion of chemically vapor processed SiC composite tubes

    Science.gov (United States)

    Kim, Daejong; Lee, Ho Jung; Jang, Changheui; Lee, Hyeon-Geun; Park, Ji Yeon; Kim, Weon-Ju

    2017-08-01

    Multi-layered SiC composites consisting of monolithic SiC and a SiCf/SiC composite are one of the accident tolerant fuel cladding concepts in pressurized light water reactors. To evaluate the integrity of the SiC fuel cladding under normal operating conditions of a pressurized light water reactor, the hydrothermal corrosion behavior of multi-layered SiC composite tubes was investigated in the simulated primary water environment of a pressurized water reactor without neutron fluence. The results showed that SiC phases with good crystallinity such as Tyranno SA3 SiC fiber and monolithic SiC deposited at 1200 °C had good corrosion resistance. However, the SiC phase deposited at 1000 °C had less crystallinity and severely dissolved in water, particularly the amorphous SiC phase formed along grain boundaries. Dissolved hydrogen did not play a significant role in improving the hydrothermal corrosion resistance of the CVI-processed SiC phases containing amorphous SiC, resulting in a significant weight loss and reduction of hoop strength of the multi-layered SiC composite tubes after corrosion.

  19. Evaluation of polymer electrolyte membrane fuel cells by electrochemical impedance spectroscopy under different operation conditions and corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, Masanobu [Taiyo Stainless Spring Co., Ltd., 2-8-6 Shakujiicho, Nerimaku, Tokyo 177-0041 (Japan); Myung, Seung-Taek; Ichikawa, Takuma; Yashiro, Hitoshi [Department of Chemical Engineering, Iwate University, 4-3-5 Ueda, Morioka, Iwate 020-8551 (Japan)

    2010-09-01

    Electrochemical impedance spectroscopy (EIS) was employed for in situ diagnosis for polymer electrolyte membrane fuel cells during operation. First, EIS was measured as a function of operation parameters such as applied current density, gas flow rates and gas humidification temperature. The resistance that correlated with conductivity of the membrane and the contact resistance between bipolar plate and gas diffusion layer (GDL) was set as R{sub m} in the assumed equivalent circuit. The charge transfer resistances were considered for cathode (R{sub ct}(C)). The value of R{sub ct}(C) was sensitive to the parameters that affected cell voltage. Additionally, the diffusion resistance (R{sub d}) was ascribed to the effect of oxygen supply and drainage of generated water. Second, the influence of corrosion of type 430 stainless steel bipolar plates was evaluated by EIS method during operation. Corrosion of the stainless steel bipolar plates resulted in an increase in the value of R{sub d}. (author)

  20. Corrosion Resistance of High Strength Concrete Containing Palm Oil Fuel Ash as Partial Cement Replacement

    OpenAIRE

    F. Mat Yahaya; Muthusamy, K.; Sulaiman, N.

    2014-01-01

    This experimental work investigates the influence of POFA as partial cement replacement towards corrosion resistance of high strength concrete. Plain high strength concrete (P0) with 100% ordinary Portland cement (control specimen) and POFA high strength concrete containing POFA as partial cement replacement material were used. At the first stage, mix with 20% POFA (P20) has been identified as the best performing mix after cubes (150×150×150 mm) containing various content of POFA as partial c...

  1. Study of hydrogen migration produced during the corrosion of PWR reactors fuel cans in zircaloy 4 and zirconia; Etude du transport de l`hydrogene produit lors de la corrosion des gaines d`elements combustibles des reacteurs a eau sous pression dans la zircone et le zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

    Aufore, L

    1997-12-12

    The corrosion of Zircaloy-4-claddings by water from the primary circuit of nuclear power plant goes hand in hand with the release of hydrogen which penetrates the oxide and then the metal. This work focuses on the mechanisms of hydrogen transport in oxide and in metal. Hydrogen transport in oxide is studied on the basis of corrosion tests performed in the autoclave at 360 deg C. These tests are performed on Zircaloy-4 claddings under different chemical conditions (pure water, and pure water with lithium hydroxide). The distribution of hydrogen in oxide film is measured by SIMS. Hydrogen profiles in the oxide are dependent on the oxide microstructure and vary with oxidation time. These observations are confirmed by experiments in which some samples, previously oxidized in the autoclave, are immersed in heavy water. In the oxide layer, two zones are observed: one external porous zone and one internal dense zone. Deuterium diffusion coefficients in dense oxide are determined using SIMS profiles and Fischer diffusion model. Hydrogen transport in metal is also studied by means of gas-phase permeation experiments. These are set up at different temperature (400-500 deg. C) and under different hydrogen pressures and make it possible to determine the hydrogen diffusion coefficients in a Zircaloy-4 cladding under experimental conditions. All these results lead us to discuss of hydrogen transport evolution in cladding during oxidation. A model taking into account hydrogen transport in oxide and in metal, and the hydrides precipitations is proposed. (author) 110 refs.

  2. Corrosion Resistance of High Strength Concrete Containing Palm Oil Fuel Ash as Partial Cement Replacement

    Directory of Open Access Journals (Sweden)

    F. Mat Yahaya

    2014-06-01

    Full Text Available This experimental work investigates the influence of POFA as partial cement replacement towards corrosion resistance of high strength concrete. Plain high strength concrete (P0 with 100% ordinary Portland cement (control specimen and POFA high strength concrete containing POFA as partial cement replacement material were used. At the first stage, mix with 20% POFA (P20 has been identified as the best performing mix after cubes (150×150×150 mm containing various content of POFA as partial cement replacement were prepared, continuously water cured and subjected to compressive strength test at 28 days. At the second stage of study, control specimen (P0 and high strength concrete mix containing 20% POFA (P20 were prepared in form of cylinders with reinforcement bar buried in the middle for corrosion resistance test. Specimens were subjected to half cell potential technique following the procedures outlined in ASTM C876 (1994. Incorporation of POFA as partial cement replacement has contributed to densification of microstructure making the concrete denser thus exhibit higher resistance towards corrosion as compared to plain concrete.

  3. Status of advanced carbide fuels: Past, present, and future

    Science.gov (United States)

    Anghaie, Samim; Knight, Travis

    2002-01-01

    Solid solution, mixed uranium/refractory metal carbide fuels such as (U, Zr, Nb)C, so called ternary carbide or tri-carbide fuels have great potential for applications in next generation advanced nuclear power reactors. Because of their high melting points, high thermal conductivity, improved resistance to hot hydrogen corrosion, and good fission product retention, these advanced nuclear fuels have great potential for high performance reactors with increased safety margins. Despite these many benefits, some concerns regarding carbide fuels include compatibility issues with coolant and/or cladding materials and their endurance under the extreme conditions associated with nuclear thermal propulsion. The status of these fuels is reviewed to characterize their performance for space nuclear power applications. Results of current investigations are presented and as well as future directions of study for these advanced nuclear fuels. .

  4. Hydrogen Sulphide Corrosion of Carbon and Stainless Steel Alloys Immersed in Mixtures of Renewable Fuel Sources and Tested Under Co-processing Conditions

    Directory of Open Access Journals (Sweden)

    Gergely András

    2016-10-01

    Full Text Available In accordance with modern regulations and directives, the use of renewable biomass materials as precursors for the production of fuels for transportation purposes is to be strictly followed. Even though, there are problems related to processing, storage and handling in wide range of subsequent uses, since there must be a limit to the ratio of biofuels mixed with mineral raw materials. As a key factor with regards to these biomass sources pose a great risk of causing multiple forms of corrosion both to metallic and non-metallic structural materials. To assess the degree of corrosion risk to a variety of engineering alloys like low-carbon and stainless steels widely used as structural metals, this work is dedicated to investigating corrosion rates of economically reasonable engineering steel alloys in mixtures of raw gas oil and renewable biomass fuel sources under typical co-processing conditions. To model a desulphurising refining process, corrosion tests were carried out with raw mineral gasoline and its mixture with used cooking oil and animal waste lard in relative quantities of 10% (g/g. Co-processing was simulated by batch-reactor laboratory experiments. Experiments were performed at temperatures between 200 and 300ºC and a pressure in the gas phase of 90 bar containing 2% (m3/m3 hydrogen sulphide. The time span of individual tests were varied between 1 and 21 days so that we can conclude about changes in the reaction rates against time exposure of and extrapolate for longer periods of exposure. Initial and integral corrosion rates were defined by a weight loss method on standard size of coupons of all sorts of steel alloys. Corrosion rates of carbon steels indicated a linear increase with temperature and little variation with composition of the biomass fuel sources. Apparent activation energies over the first 24-hour period remained moderate, varying between 35.5 and 50.3 kJ mol−1. Scales developed on carbon steels at higher

  5. Angra 1 high burnup fuel behaviour under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: dsgomes@ipen.b, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The 16x16 NGF (Next Generation Fuel) fuel assembly, comprising of highly corrosive-resistant ZIRLO clad fuel rods, been replacing the current 16x16 Standard (16STD) fuel assembly in the Angra 1, a pressurized water reactor, with a net output of 626 MWe. The 16x16 NGF fuel assemblies are designed for a peak rod average burnup of up to 75 GWd/MTU, thus improving fuel utilization and reducing spent fuel storage issues. A design basis accident, the Reactivity Initiated Accident (RIA), became a concern for a further increase in burnup as the simulated RIA tests revealed a lower enthalpy threshold for fuel failure. Two fuel performance codes, FRAPCON and FRAPTRAN, were used to predict high burnup behavior of Angra 1, during an RIA. The maximum average linear fuel rating used was 17.62 KW/m. The FRAPCON 3.4 code was applied to simulate the steady-state performance of the 16 NGF fuel rods up to a burnup of 55 GWd/MTU. With FRAPTRAN-1.4 the fuel behavior was simulated for an RIA power pulse of 4.5 ms (FHWH), and enthalpy peak of 130 Cal/g. With FRAPCON-3.4, the corrosion and hydrogen pickup characteristics of the advanced ZIRLO clad fuel rods were added to the code by modifying the actual corrosion model for Zircaloy-4 through the multiplication of empirical factors, which were appropriate to each alloy, and by means of reducing the current hydrogen pickup fraction. (author)

  6. An investigation of the typical corrosion parameters used to test polymer electrolyte fuel cell bipolar plate coatings, with titanium nitride coated stainless steel as a case study

    Science.gov (United States)

    Orsi, A.; Kongstein, O. E.; Hamilton, P. J.; Oedegaard, A.; Svenum, I. H.; Cooke, K.

    2015-07-01

    Stainless steel bipolar plates (BPP) for polymer electrolyte membrane fuel cells (PEMFCs) have good manufacturability, durability and low costs, but inadequate corrosion resistance and elevated interfacial contact resistance (ICR) in the fuel cell environment. Thin film coatings of titanium nitride (TiN) of 1 μm in thickness, were deposited by means of physical vapour deposition (PVD) process on to stainless steel (SS) 316L substrates and were evaluated, in a series of tests, for their level of corrosion protection and ICR. In the ex-situ corrosion tests, variables such as applied potential, experimental duration and pH of the sulphate electrolyte at 80 °C were altered. The ICR values were found to increase after exposure to greater applied potentials and electrolytes of a higher pH. In terms of experimental duration, the ICR increased most rapidly at the beginning of each experiment. It was also found that the oxidation of TiN was accelerated after exposure to electrolytes of a higher pH. When coated BPPs were incorporated into an accelerated fuel cell test, the degradation of the fuel cell cathode resembled the plates that were tested at the highest anodic potential (1.4 VSHE).

  7. Elimination of Start/Stop defects in laser cladding

    NARCIS (Netherlands)

    Ocelik, V.; Eekma, M.; Hemmati, I.; De Hosson, J. Th. M.

    2012-01-01

    Laser cladding represents an advanced hard facing technology for the deposition of hard, corrosion and wear resistant layers of controlled thickness onto a selected area of metallic substrate. When a circular geometry is required, the beginning and the end of the laser track coincide in the same are

  8. 建立反应堆燃料元件破损运行判据的思考%A Scheme for Establishing of Criterions for Reactor Safe Operation in Condition of Fuel Clad Failure

    Institute of Scientific and Technical Information of China (English)

    林晓玲

    2013-01-01

    Operation criterions are used to decide if the reactor can continue to work when the fuel clad failure. The method for establish the limits is presented. The tolerated maximum of failure fuel rods for the reactor safety should be calculated by risk analysis. The parameters are determined which can not only reflect the quantity but also be measured directly. The relationship is set up between the amounts with the parameters. The data calculated corresponding to maximum of failure fuel element which the reactor safety can stand are technical limits used to decide if the reactor can work continually.%运行判据是用于判断反应堆燃料元件发生破损时能否继续运行的指标条件,本文提出建立反应堆燃料元件破损运行判据的思路和方法,通过风险分析,确定监督运行最大容许破损数量;研究提出既能反映燃料元件破损数量又可直接监测的指标参量,并建立破损数量与可监测指标参量之间的对应关系;将最大容许破损数量对应的可监测指标参量值作为运行技术判据.

  9. Carbon corrosion of proton exchange membrane fuel cell catalyst layers studied by scanning transmission X-ray microscopy

    Science.gov (United States)

    Hitchcock, Adam P.; Berejnov, Viatcheslav; Lee, Vincent; West, Marcia; Colbow, Vesna; Dutta, Monica; Wessel, Silvia

    2014-11-01

    Scanning Transmission X-ray Microscopy (STXM) at the C 1s, F 1s and S 2p edges has been used to investigate degradation of proton exchange membrane fuel cell (PEM-FC) membrane electrode assemblies (MEA) subjected to accelerated testing protocols. Quantitative chemical maps of the catalyst, carbon support and ionomer in the cathode layer are reported for beginning-of-test (BOT), and end-of-test (EOT) samples for two types of carbon support, low surface area carbon (LSAC) and medium surface area carbon (MSAC), that were exposed to accelerated stress testing with upper potentials (UPL) of 1.0, 1.2, and 1.3 V. The results are compared in order to characterize catalyst layer degradation in terms of the amounts and spatial distributions of these species. Pt agglomeration, Pt migration and corrosion of the carbon support are all visualized, and contribute to differing degrees in these samples. It is found that there is formation of a distinct Pt-in-membrane (PTIM) band for all EOT samples. The cathode thickness shrinks due to loss of the carbon support for all MSAC samples that were exposed to the different upper potentials, but only for the most aggressive testing protocol for the LSAC support. The amount of ionomer per unit volume significantly increases indicating it is being concentrated in the cathode as the carbon corrosion takes place. S 2p spectra and mapping of the cathode catalyst layer indicates there are still sulfonate groups present, even in the most damaged material.

  10. Microstructural and Mechanical Characterization Study of Cr Coated ATF Claddings After Simulated Integral LOCA Test

    Energy Technology Data Exchange (ETDEWEB)

    Park, Dong Jun; Park, Jung Hwan; Jung, Yang Il; Kim, Hyun Gil; Park, Jeong Yong; Koo, Yang Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Accident tolerant fuel (ATF) cladding has been widely studied by several research groups after Fukushima nuclear reactor accident. Oxidation barrier layer coated Zr fuel cladding is one of the most promising candidate concepts owing to its easy process and lower cost for manufacturing and possibility of developing with short term study compared to other ATF concepts. Coated layer on the surface of Zr tube sample was formed by cold spray coating process. Main requirement of these ATF claddings may be high temperature oxidation resistance. Therefore, their oxidation kinetics and mechanisms have been studied at a wide range of temperatures and in various environments. However, just small plate or short tube samples were simply exposed to a high temperature steam environment. In this study, integral loss-of-coolant accident (LOCA) tests simulating real conditions of fuel claddings during accident were conducted using Cr coated ATF cladding sample for a clear understanding of their behavior under accident conditions. Ballooning behavior and microstructural changes of ATF cladding during the LOCA scenarios were studied systematically and mechanical test results are also presented. Cr coated cladding samples have been successfully fabricated by using existing Zr alloy fuel claddings. For comparative study, integral LOCA test was carried out using Cr coated ATF cladding and existing Zr alloy tube sample. Cr coated ATF cladding showed much smaller rupture opening and circumferential elongation compared to Zr alloy sample. Coated Cr layer prevented outer surface oxidation in spite of exposure for 300s at 1200 .deg. C in steam environment.

  11. Microstructural and electrochemical characterization of laser deposited 18-10 austenitic stainless steel clad layers

    Energy Technology Data Exchange (ETDEWEB)

    Fouquet, F. (GEMPPM/CALFETMAT, 69 Villeurbanne (France)); Sallamand, P. (GEMPPM/CALFETMAT, 69 Villeurbanne (France)); Millet, J.P. (GEMPPM/CALFETMAT, 69 Villeurbanne (France) Physicochimie Industrielle, 69 Villeurbanne (France)); Frenk, A. (GEMPPM/CALFETMAT, 69 Villeurbanne (France) Centre de Traitement des Materiaux par Laser (CTML), Ecole Polytechnique Federale de Lausanne (Switzerland)); Wagniere, J.D. (GEMPPM/CALFETMAT, 69 Villeurbanne (France) Centre de Traitement des Materiaux par Laser (CTML), Ecole Polytechnique Federale de Lausanne (Switzerland))

    1993-11-01

    The present work reports on 18-10 stainless steel coatings produced by laser powder cladding technique on a mild steel. Uniform clad layers - about 600 [mu]m thick - have been produced through partially overlapping single cladding tracks. The clad layers thus obtained show excellent adherence, no cracks, few porosities and good chemical homogeneity. The microstructure is dendritic or cellular. Dentrites or cells have an austenitic structure and a small amount of [delta]-ferrite is detected in the interdendritic areas. The corrosion resistance of the clad layers is tested by electrochemical techniques in various neutral or acidified aqueous saline media, deaerated or naturally aerated. In every case, the coatings show an excellent uniform corrosion resistance. (orig.).

  12. Aviation Fuel Pipeline Buried Corrosion and Protection%航空燃料埋地输送管道的腐蚀与防护

    Institute of Scientific and Technical Information of China (English)

    秦一峰

    2012-01-01

    随着航空事业的发展,管输方式成为航空燃料又一广泛运用的运输方式,但腐蚀穿孔成为埋地管道的重要安全隐患。文章从腐蚀的分类分析了航空燃料埋地输送管道腐蚀的原因、影响因素,并重点从外防腐涂层和阴极保护两方面论述了当前对管道腐蚀防护的技术。%With the development of aviation,aviation fuel pipeline has become a widely used means of transport,but the corrosion of buried pipeline safety become important.In the paper the classification of corrosion analysis of aviation fuel buried pipeline corrosion,influencing factors,and from the point of external anticorrosive coating and cathodic protection are discussed in two aspects of the pipeline corrosion protection technology.

  13. The Mechanical Response of Advanced Claddings during Proposed Reactivity Initiated Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Cinbiz, Mahmut N [ORNL; Brown, Nicholas R [ORNL; Terrani, Kurt A [ORNL; Lowden, Rick R [ORNL; ERDMAN III, DONALD L [ORNL

    2017-01-01

    This study investigates the failure mechanisms of advanced nuclear fuel cladding of FeCrAl at high-strain rates, similar to design basis reactivity initiated accidents (RIA). During RIA, the nuclear fuel cladding was subjected to the plane-strain to equibiaxial tension strain states. To achieve those accident conditions, the samples were deformed by the expansion of high strength Inconel alloy tube under pre-specified pressure pulses as occurring RIA. The mechanical response of the advanced claddings was compared to that of hydrided zirconium-based nuclear fuel cladding alloy. The hoop strain evolution during pressure pulses were collected in situ; the permanent diametral strains of both accident tolerant fuel (ATF) claddings and the current nuclear fuel alloys were determined after rupture.

  14. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    Science.gov (United States)

    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.

  15. An Assessment of Alternative Diesel Fuels: Microbiological Contamination and Corrosion Under Storage Conditions

    Science.gov (United States)

    2010-08-01

    example, Clados- parium (Hormoconis) resinae grew in 80 mg water per 1 of kerosene and after 4 weeks incubation, the concentration of water increased...common isolate related to aircraft fuel and MIC is the fungus Hormoeonis resinae (Churchill 1963; Hendey 1964; Videla et al. 1993). de Mele et al. (1979...Videla (1996) demonstrated acid-etched traces of fungal mycelia on aluminum surfaces colonized by //. resinae . de Meybaum and de Schiapparelli

  16. High temperature corrosion of metallic interconnects in solid oxide fuel cells

    OpenAIRE

    Martínez Bastidas, David

    2006-01-01

    Research and development has made it possible to use metallic interconnects in solid oxide fuel cells (SOFC) instead of ceramic materials. The use of metallic interconnects was formerly hindered by the high operating temperature, which made the interconnect degrade too much and too fast to be an efficient alternative. When the operating temperature was lowered, the use of metallic interconnects proved to be favourable since they are easier and cheaper to produce than ceramic interconnects....

  17. Stone cladding engineering

    National Research Council Canada - National Science Library

    Camposinhos, Rui de Sousa

    2014-01-01

    .... Straightforward formulae are provided for computing action on cladding, with special emphasis on the effect of seismic forces, including an extensive general methodology applied to non-structural elements...

  18. An austenitic steel for fuel cladding tubes and core components of LMFBR`s with high ductility after neutron irradiation; Ein austenitischer Stahl fuer Huellrohre und Kernkomponenten natriumgekuehlter Brueter mit hoher Duktilitaet nach Neutronenbestrahlung

    Energy Technology Data Exchange (ETDEWEB)

    Schaefer, L.; Kempe, H.

    1994-06-01

    Two heats of an austenitic stainless steel with different priority concerning the resistance against Helium-embrittlement (B801) and void-swelling (F218) had been developed and tested as a material for fuel rod claddings and core components of liquid metal fast breeder reactors. The two steels show a ductility five times higher than the reference steel 1.4970 in tensile - and creep-rupture-tests after irradiation in reactors with fast and mixed neutron flux respectively. Just so the swelling resistance had been confirmed up to 40 dpa. Checked claddings of the heat F218 in the dimensions 6x0.38 mm, 6.55x0.45 mm and 7.6x0.5 mm are available for pin- and bundle irradiation experiments. (orig.) [Deutsch] Im Rahmen der Entwicklung austenitischer Staehle als Werkstoffe fuer Huellrohre und Kernkomponenten Schneller Natriumgekuehlter Brutreaktoren wurden zwei Chargen mit unterschiedlicher Prioritaet fuer ihre Widerstandsfaehigkeit gegen Heliumversproedung (B801) und Porenschwellen (F218) konzipiert und untersucht. Beide Staehle zeigten nach Bestrahlung in Reaktoren mit schnellem bzw. gemischtem Neutronenfluss sowohl im Warmzugversuch als auch im Zeitstandversuch eine Duktilitaet, die um den Faktor 5 hoeher liegt als die des Referenzstahles 1.4970. Fuer beide Staehle konnte die Schwellresistenz bis 40 dpa Neutronenbestrahlung nachgewiesen werden. Fuer Brennstab- und Buendelbestrahlungsexperimente stehen gepruefte Huellrohre der Charge F218 mit den Abmessungen 6x0.38 mm, 6.55x0.45 mm und 7.6x0.5 mm zur Verfuegung. (orig.)

  19. Corrosion behavior of technetium waste forms exposed to various aqueous environments

    Energy Technology Data Exchange (ETDEWEB)

    Kolman, David Gary [Los Alamos National Laboratory; Jarvinen, Gordon [Los Alamos National Laboratory; Mausolf, Edward [UNIV OF NEVADA; Czerwinski, Ken [UNIV OF NEVADA; Poineau, Frederic [UNIV OF NEVADA

    2009-01-01

    Technetium is a long-lived beta emitter produced in high yields from uranium as a waste product in spent nuclear fuel and has a high degree of environmental mobility as pertechnetate. It has been proposed that Tc be immobilized into various metallic waste forms to prevent Tc mobility while producing a material that can withstand corrosion exposed to various aqueous medias to prevent the leachability of Tc to the environment over long periods of time. This study investigates the corrosion behavior of Tc and Tc alloyed with 316 stainless steel and Zr exposed to a variety of aqueous media. To date, there is little investigative work related to Tc corrosion behavior and less related to potential Tc containing waste forms. Results indicate that immobilizing Tc into stainless steel-zirconium alloys can be a promising technique to store Tc for long periods of time while reducing the need to separately store used nuclear fuel cladding. Initial results indicate that metallic Tc and its alloys actively corrode in all media. We present preliminary corrosion rates of 100% Tc, 10% Tc - 90% SS{sub 85%}Zr{sub 15%}, and 2% Tc - 98% SS{sub 85%}Zr{sub 15%} in varying concentrations of nitric acid and pH 10 NaOH using the resistance polarization method while observing the trend that higher concentrations of Tc alloyed to the sample tested lowers the corrosion rate of the proposed waste package.

  20. Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor (SFR)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    To achieve manufacturing technology of the fuel cladding tube in order to keep pace with the predetermined schedule in developing SFR fuel, KAERI has launched in developing fuel cladding tube in cooperation with a domestic steelmaking company. After fabricating medium-sized 1.1 ton HT9 ingot, followed by the multiple processes of hot and cold working, preliminary samples of HT9 seamless cladding tube having 7.4mm in outer diameter, 0.56mm in thickness, and 3m in length were fabricated. The objective of this study is to summarize the brief development status of the HT9 cladding tubes. Mechanical properties like axial tension, biaxial burst, pressurized creep and sodium compatibility of the cladding tubes were carried out to set up the performance evaluation technology to test the prototype FMS cladding tube which is going to be manufactured in next stage. As a part of developing fuel cladding for the Sodium-cooled Fast Reactor (SFR), preliminary HT9 cladding tube was fabricated in cooperation with a domestic steelmaking company. Microstructure as well as mechanical tests like axial tensile test, biaxial burst test, and pressurized creep test of the fuel cladding were carried out. Performance of the domestic HT9 tube was revealed to be similar in the previously fabricated foreign HT9 tube. Further prototype FMS cladding tube is going to be manufactured in next year based on this experience. Various test items like mechanical test, sodium compatibility test, microstructural analysis, basic property, cladding performance under transient situation, and performance under ion and neutron irradiation are going be performed in the future to set up the relevant technology for the licensing of the SFR cladding tube.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

  2. Probabilistic Failure Analysis for Wound Composite Ceramic Cladding Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hemrick, James Gordon [ORNL; Lara-Curzio, Edgar [ORNL

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

    Directory of Open Access Journals (Sweden)

    V. I. Solonin

    2014-01-01

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

  4. Pre-oxidized and nitrided stainless steel alloy foil for proton exchange membrane fuel cell bipolar plates: Part 1. Corrosion, interfacial contact resistance, and surface structure

    Science.gov (United States)

    Brady, M. P.; Wang, H.; Turner, J. A.; Meyer, H. M.; More, K. L.; Tortorelli, P. F.; McCarthy, B. D.

    Thermal (gas) nitridation of stainless steel alloys can yield low interfacial contact resistance (ICR), electrically conductive and corrosion-resistant nitride containing surface layers (Cr 2N, CrN, TiN, V 2N, VN, etc.) of interest for fuel cells, batteries, and sensors. This paper presents results of scale-up studies to determine the feasibility of extending the nitridation approach to thin 0.1 mm stainless steel alloy foils for proton exchange membrane fuel cell (PEMFC) bipolar plates. Developmental Fe-20Cr-4V alloy and type 2205 stainless steel foils were treated by pre-oxidation and nitridation to form low-ICR, corrosion-resistant surfaces. As-treated Fe-20Cr-4V foil exhibited target (low) ICR values, whereas 2205 foil suffered from run-to-run variation in ICR values, ranging up to 2× the target value. Pre-oxidized and nitrided surface structure examination revealed surface-through-layer-thickness V-nitride particles for the treated Fe-20Cr-4V, but near continuous chromia for treated 2205 stainless steel, which was linked to the variation in ICR values. Promising corrosion resistance was observed under simulated aggressive PEMFC anode- and cathode-side bipolar plate conditions for both materials, although ICR values were observed to increase. The implications of these findings for stamped bipolar plate foils are discussed.

  5. Nuclear fuel can

    Energy Technology Data Exchange (ETDEWEB)

    Kubo, Toshio.

    1990-08-16

    Oxide membrane of less than 1 {mu}m thickness are formed on the outer surface of a cladding tube by means of an anodic oxidation method. Thus, anodized membranes have greater density and toughness compared with those of oxide layers formed in high temperature water and act as protection membranes against corrosion. Further, when the anodized membranes are annealed in vacuum at 250 to 450degC, cubic oxide membranes are also stabilized in addition to ordinary monoclinic oxide membranes, to form specific oxide membranes in which cubic and monoclinic crystals are present together. Accordingly, this can provide a more excellent corrosion resistance against nodular corrosion compared with conventional cladding tubes. Further, since the improvement of the corrosion resistance is derived from the formation of strong oxide membranes, neutron economy and mechanical properties of the cladding tubes are not degraded as in usual cases. (T.M.).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-01

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

  7. Wear behavior and corrosion resistance of NiCrAl/TiC composite coating on aluminum alloy by laser cladding%铝合金表面激光熔覆NiCrAl/TiC复合涂层的磨损行为和耐蚀性能

    Institute of Scientific and Technical Information of China (English)

    李琦; 刘洪喜; 张晓伟; 姚爽; 张旭

    2014-01-01

    为提高铝合金的摩擦磨损和耐蚀性能,在A390铝合金基体上通过激光熔覆制备NiCrAl/TiC复合涂层。采用XRD和EDS分析了涂层的物相组成,结合SEM观察了涂层的微观组织,运用摩擦磨损试验机和电化学工作站测试了涂层的摩擦磨损和耐腐蚀性能。结果表明:复合涂层主要物相为AlNi、Al 3 Ni 2、TiC ,同时含有少量的Cr 13 Ni 5 Si 2、Cu 9 Al 4和α(Al)。涂层自下至上分别为短棒状树枝晶、胞状晶、柱状树枝晶和等轴晶。相同磨损条件下,A390基体发生了严重的磨粒磨损和剥层磨损,而激光熔覆涂层只产生了轻微的磨粒磨损,熔覆层的相对耐磨性为3.16。在3.5%NaCl溶液中的极化曲线和电化学阻抗谱(EIS)显示:熔覆层自腐蚀电位较A390基体的正移,腐蚀电流密度减小;熔覆层呈单容抗特性,而A390基体在高频区表现为容抗特性,在中低频区则为感抗特性。在Bote图中,低频区熔覆层对应的相位角和中低频段熔覆层的阻抗模值均大于A390基体的,表明熔覆层的耐蚀性远高于A390基体的。熔覆层的腐蚀形貌为局部点蚀,A390基体的腐蚀形貌为晶间腐蚀和剥蚀。%In order to improve the frictional wear behavior and corrosion resistance of aluminum alloy, NiCrAl/TiC composite coating was fabricated on A390 aluminum alloy by laser cladding. The phase constitution, microstructure, frictional wear behavior and corrosion resistance of the composite coating were analyzed using X-ray diffraction (XRD), energy dispersive spectrum (EDS), scanning electron microscope (SEM), friction and wear testing machine and electrochemical workstation. The results show that the coating is mainly composed of AlNi, Al 3 Ni 2 and TiC phases, and a small amount of Cr13Ni5Si2, Cu9Al4 and α(Al) phases. The microstructures of the coating from the bottom to top are dendrite crystal, cellular crystal, columnar dendrite crystal and equiaxed

  8. A model to describe the mechanical behavior and the ductile failure of hydrided Zircaloy-4 fuel claddings between 25 °C and 480 °C

    Science.gov (United States)

    Le Saux, M.; Besson, J.; Carassou, S.

    2015-11-01

    A model is proposed to describe the mechanical behavior and the ductile failure at 25, 350 and 480 °C of Zircaloy-4 cladding tubes, as-received and hydrided up to 1200 wt. ppm (circumferential hydrides). The model is based on the Gurson-Tvergaard-Needleman model extended to account for plastic anisotropy and viscoplasticity. The model considers damage nucleation by both hydride cracking and debonding of the interface between the Laves phase precipitates and the matrix. The damage nucleation rate due to hydride cracking is directly deduced from quantitative microstructural observations. The other model parameters are identified from several experimental tests. Finite element simulations of axial tension, hoop tension, expansion due to compression and hoop plane strain tension experiments are performed to assess the model prediction capability. The calibrated model satisfactorily reproduces the effects of hydrogen and temperature on both the viscoplastic and the failure properties of the material. The results suggest that damage is anisotropic and influenced by the stress state for the non-hydrided or moderately hydrided material and becomes more isotropic for high hydrogen contents.

  9. Crack resistance curve determination of zircaloy-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-03-15

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

  10. Superheater corrosion in a boiler fired with refuse-derived fuel

    Energy Technology Data Exchange (ETDEWEB)

    Blough, J.L.; Stanko, G.J. [Foster Wheeler Development Corp., Livingston, NJ (United States); Bakker, W.T. [Electric Power Research Inst., Palo Alto, CA (United States); Steinbeck, T. [United Power Association, Elk River, MN (United States)

    1995-12-31

    The environment in the superheater of a boiler firing refuse-derived fuel (RDF) is very aggressive. The high wastage rate for the standard T-22 material necessitated a materials testing program. Simples of Types 304H, HR3C, T-22 chromized, 825 and 625 were assembled into tubular test sections and welded into the superheater tubing. After 1,180 hours the test sections were evaluated and the wastage rates calculated for each material. The chlorides contained in the RDF are believed to be the primary corrodent. The chlorine may be interacting with the metal samples as HCl, a low-melting-point eutectic or a combination of these. Of the six materials tested, Alloy 625 exhibited the best resistance--substantially better than the next-best Type 304. Alloy 825 and HR3C corroded approximately 1.5 times the rate of Type 304. The chromized layer on T-22 showed no resistance to the environment and was consumed in large areas.

  11. 用于超临界水堆燃料包壳的ODS铁素体钢的研究进展%Progress of Using Oxide Dispersion Strengthened Ferritic Steels as Fuel Cladding Materials in Supercritical Water Reactor

    Institute of Scientific and Technical Information of China (English)

    何培; 周张健; 李明; 许迎利; 葛昌纯

    2009-01-01

    超临界水堆具有热效率高、系统简化、成本低等优点,成为第四代核反应堆中优先发展的堆型.ODS铁索体钢由于其优异的高温力学性能和良好的抗辐照能力成为超临界水堆包壳最有希望的候选材料.旨在回顾ODS铁素体钢制造工艺,包括机械合金化参数的优化,热处理工艺的选择以消除力学性能上的各向异性.根据超临界水堆包壳的服役条件,结合最新的实验数据,对ODS铁素体钢的高温力学性能、在超临界水中的耐腐蚀性以及中子辐照稳定性进行了总结和展望.%Supercritical water reactor (SCWR) is considered to be the most promising reactor among Gen IV reactors due to its higher thermal efficiency, considerable system simplification and improved economics. ODS ferritic steels have been considered as one of promising cladding candidate materials for SCWR because of the excellent properties, such as superior high temperature strength and outstanding swelling resistance. The aim of this paper is to review both the fabrication technology of ODS ferritic steels, including the optimization of mechanical alloying parameters and thermal treatment methods for ameliorating the densification and deforming work induced mechanical anisotropy, and the evaluation of the high temperature mechanical properties, corrosion resistance in SCW and neutron irradiation resistance of ODS ferritic steels according to the working conditions in SCWR.

  12. An electrochemical investigation of the corrosion behavior of aluminum alloys in chloride containing solutions; Investigacao eletroquimica da corrosao de ligas de aluminio em solucoes contendo cloretos

    Energy Technology Data Exchange (ETDEWEB)

    Campos Filho, Jorge Eustaquio de [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Escola de Engenharia. Dept. de Engenharia Quimica]. E-mail: jorgecamposfilho@yahoo.com.br; Neves, Celia de Figueiredo Cordeiro; Campos, Wagner Reis da Costa; Moreira, Marcilio Soares [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)]. E-mail: caf@cdtn.br; wrcc@cdtn.br; msm@cdtn.br

    2005-07-01

    Aluminum alloys have been used as cladding materials for nuclear fuel in research reactors due to its corrosion resistance. Aluminum owes its good corrosion resistance to a protective barrier oxide film formed and strongly bonded to its surface. In pool type TRIGA IPR-R1 reactor, located at Centro de Desenvolvimento da Tecnologia Nuclear in Belo Horizonte, previous immersion coupon tests revealed that aluminum alloys suffer from pitting corrosion, in spite of high quality of water control. Corrosion attack is initiated by breaking the protective oxide film on aluminum alloy surface. Chloride ions can break this oxide film and stimulate metal dissolution. In this study the aluminum alloys 1050, 5052 and 6061 were used to evaluate their corrosion behavior in chloride containing solutions. The electrochemical techniques used were potentiodynamic anodic polarization and cyclic polarization. Results showed that aluminum alloys 5052 and 6061 present similar corrosion resistance in low chloride solutions (0,1 ppm NaCl) and in reactor water but both alloys are less resistant in high chloride solution (1 ppm NaCl). Aluminum alloy 1050 presented similar behavior in the three electrolytes used, regarding to pitting corrosion, indicating that the concentration of the chloride ions was not the only variable to influence its corrosion susceptibility. (author)

  13. Field test corrosion experiments in Denmark with biomass fuels Part II Co-firing of straw and coal

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Larsen, OH

    2002-01-01

    superheaters. A range of austenitic and ferritic steels was exposed in the steam temperature range of 520-580°C. The flue gas temperature ranged from 925-1100°C. The rate of corrosion was assessed by precision measurement of material loss and measurement of oxide thickness. Corrosion rates are lower than...... and potassium sulphate. These components give rise to varying degrees of accelerated corrosion. This paper concerns co-firing of straw with coal to reduce the corrosion rate from straw to an acceptable level. A field investigation at Midtkraft Studstrup suspension-fired power plant in Denmark has been...... undertaken where coal has been co-fired with 10% straw and 20% straw (% energy basis) for up to approx. 3000 hours. Two types of exposure were undertaken to investigate corrosion: a) the exposure of metal rings on water/air cooled probes, and b) the exposure of a range of materials built into the existing...

  14. Improve Design of Fuel Shear for Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    GAO; Wei; OUYANG; Ying-gen; LI; Wei-min

    2012-01-01

    <正>Due to the deeper burnup and higher fuel swelling, fast reactor metal fuel rod using 316 stainless steel cladding, replacing the traditional zirconia cladding. The diameter of fuel rod of fast reactor is much longer than that of PWR, and the cladding of stainless steel has better ductility than zirconia cladding. Using the existing shear still will cause several aspects of problem: 1) Longer diameter of rod leads to

  15. Accident tolerant fuels for LWRs: A perspective

    Science.gov (United States)

    Zinkle, S. J.; Terrani, K. A.; Gehin, J. C.; Ott, L. J.; Snead, L. L.

    2014-05-01

    The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms.

  16. Fuel performance improvement program. Quarterly/annual progress report, October 1978-September 1979

    Energy Technology Data Exchange (ETDEWEB)

    Crouthamel, C.E. (comp.)

    1979-10-01

    The objective of the Fuel Performance Improvement Program is to develop, test, and demonstrate basically two advanced fuel designs with the capability for improved power ramping performance and thus increase the capability of achieving extended burnup levels to better utilize uranium resources. The irradiations are being supported by out-of-reactor experiments to evaluate the effect of graphite coatings to inhibit stress-corrosion-cracking type cladding failures that are related to pellet-cladding interaction. Instrumented test irradiations in the Halden Boiling Water Reactor (HBWR) have achieved peak burnups of 697 GJ/kgM (8.1 MWd/kgM) with reference, annular-coated-pressurized, and sphere-pac rods.

  17. Microstructure of U 3Si 2 fuel plates submitted to a high heat flux

    Science.gov (United States)

    Leenaers, A.; Van den Berghe, S.; Koonen, E.; Jacquet, P.; Jarousse, C.; Guigon, B.; Ballagny, A.; Sannen, L.

    2004-05-01

    In order to gain insight on the performance limits of U 3Si 2 fuel with Al cladding, fuel plates with a fissile material density of 5.1 and 6.1 g U/cm 3 were irradiated in the BR2 reactor of SCK • CEN in Mol. The plates were intended to be subjected to severe conditions leading to a cladding surface temperature of 180-200 °C and fuel temperatures of 220-240 °C. The irradiation program was stopped after the second cycle based on the visual inspection and wet sipping tests of the elements, correspondingly showing degradations on the outer Al surfaces of the U 3Si 2 plates and the release of fission products. The maximum fuel burn-up was 29% and 25% 235U, respectively. In a PIE program the microstructural causes for this degradation are analysed. It is found that the failure of the plates is entirely related to the corrosion of the Al cladding, which has caused temperatures to rise well beyond the calculated values. In all stages, the fuel grains have retained their integrity and, apart from the formation of an interaction phase with the Al matrix, they do not demonstrate deleterious changes in their physical properties.

  18. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    Science.gov (United States)

    González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

    2015-10-01

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

  19. Accumulation of radioactive corrosion products on steel surfaces of VVER type nuclear reactors. I. 110mAg

    CSIR Research Space (South Africa)

    Hirschberg, G

    1999-03-01

    Full Text Available of radioactive corrosion products on steel surfaces of VVER type nuclear reactors. I. 110mAg G abor Hirschberg a,P al Baradlai a,K alm an Varga a,*, Gerrit Myburg b, J anos Schunk c,P eter Tilky c, Paul Stoddart d a Department of Radiochemistry, University...-cooled nuclear reactors is of great importance for a number of practical reasons. For instance, under normal operating conditions (when there is no ?ssion product release due to fuel cladding failure) the majority of radioactive contamination in the pri- mary...

  20. Most advanced HTP fuel assembly design for EPR

    Energy Technology Data Exchange (ETDEWEB)

    Francillon, Eric [AREVA - Framatome ANP, 10 rue Juliette Recamier - 69456 Lyon Cedex 06 (France); Kiehlmann, Horst-Dieter [AREVA - Framatome ANP GmbH, P.O. Box 3220, 91050 Erlangen (Germany)

    2006-07-01

    End 2003, the Finnish electricity utility Teollisuuden Voima Oy (TVO) signed the contract for building an EPR in Olkiluoto (Finland). Mid 2004, the French electricity utility EDF selected an EPR to be built in France. In 2005, Framatome ANP, an AREVA and Siemens company, announced that they will be pursuing a design certification in the U.S. The EPR development is based on the latest PWR product lines of former Framatome (N4) and Siemens Nuklear (Konvoi). As an introductory part, different aspects of the EPR core characteristics connected to fuel assembly design are presented. It includes means of ensuring reactivity control like hybrid AIC/B4C control rod absorbers and gadolinium as burnable absorber integrated in fuel rods, and specific options for in-core instrumentation, such as Aeroball type instrumentation. Then the design requirements for the EPR fuel assembly are presented in term of very high burnup capacity, rod cladding and fuel assembly reliability. Framatome ANP fuel assembly product characteristics meeting these requirements are then described. EPR fuel assembly design characteristics benefit from the experience feedback of the latest fuel assembly products designed within Framatome ANP, leading to resistance to assembly deformation, high fuel rod restraint and prevention of handling hazards. EPR fuel assembly design features the best components composing the cornerstones of the upgraded family of fuel assemblies that FRAMATOME ANP proposes today. This family is based on a set of common characteristics and associated features, which include the HMP grid as bottom end spacer, the MONOBLOC guide tube and the Robust FUELGUARD as lower tie plate, the use of the M5 Alloy, as cladding and structure material. This fully re-crystallized, ternary Zr-Nb-O alloy produces radically improved in-reactor corrosion, very low hydrogen uptake and growth and an excellent creep behavior, which are described there. EPR fuel assembly description also includes fuel rod

  1. Deep surface rolling for fatigue life enhancement of laser clad aircraft aluminium alloy

    Energy Technology Data Exchange (ETDEWEB)

    Zhuang, W., E-mail: wyman.zhuang@dsto.defence.gov.au [Aerospace Division, Defence Science and Technology Organisation, 506 Lorimer Street, Fishermans Bend, Victoria 3207 (Australia); Liu, Q.; Djugum, R.; Sharp, P.K. [Aerospace Division, Defence Science and Technology Organisation, 506 Lorimer Street, Fishermans Bend, Victoria 3207 (Australia); Paradowska, A. [Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW 2232 (Australia)

    2014-11-30

    Highlights: • Deep surface rolling as a post-repair enhancement technology was applied to the laser cladded 7075-T651 aluminium alloy specimens that simulated corrosion damage blend-out repair. • The residual stresses induced by the deep surface rolling process were measured. • The deep surface rolling process can introduce deep and high magnitude compressive residual stresses beyond the laser clad and substrate interface. • Spectrum fatigue test showed the fatigue life was significantly increased by deep surface rolling. - Abstract: Deep surface rolling can introduce deep compressive residual stresses into the surface of aircraft metallic structure to extend its fatigue life. To develop cost-effective aircraft structural repair technologies such as laser cladding, deep surface rolling was considered as an advanced post-repair surface enhancement technology. In this study, aluminium alloy 7075-T651 specimens with a blend-out region were first repaired using laser cladding technology. The surface of the laser cladding region was then treated by deep surface rolling. Fatigue testing was subsequently conducted for the laser clad, deep surface rolled and post-heat treated laser clad specimens. It was found that deep surface rolling can significantly improve the fatigue life in comparison with the laser clad baseline repair. In addition, three dimensional residual stresses were measured using neutron diffraction techniques. The results demonstrate that beneficial compressive residual stresses induced by deep surface rolling can reach considerable depths (more than 1.0 mm) below the laser clad surface.