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Sample records for aluminum clad al-li

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  2. Proposal of 99.99%-aluminum/7N01-Aluminum clad beam tube for high energy booster of Superconducting Super Collider

    International Nuclear Information System (INIS)

    Ishimaru, Hajime

    1994-01-01

    Proposal of 99.99% pure aluminum/7N01 aluminum alloy clad beam tube for high energy booster in Superconducting Super Collider is described. This aluminum clad beam tube has many good performances, but a eddy current effect is large in superconducting magnet quench collapse. The quench test result for aluminum clad beam tube is basically no problem against magnet quench collapse. (author)

  3. 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.

  4. Scientific basis for storage criteria for interim dry storage of aluminum-clad fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Peacock, H.B. Jr.; Lam, P.S.; Iyer, N.C.; Louthan, M.R. Jr.; Murphy, J.R. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1996-08-01

    An engineered system for dry storage of aluminum-clad foreign and domestic research reactor spent fuel owned by the US Department of Energy is being considered to store the fuel up to a nominal period of 40 years prior to ultimate disposition. Scientifically-based criteria for environmental limits to drying and storing the fuels for this system are being developed to avoid excessive degradation in sealed and non-sealed (open to air) dry storage systems. These limits are based on consideration of degradation modes that can cause loss of net section of the cladding, embrittlement of the cladding, distortion of the fuel, or release of fuel and fission products from the fuel/clad system. Potential degradation mechanisms include corrosion mechanisms from exposure to air and/or sources of humidity, hydrogen blistering of the aluminum cladding, distortion of the fuel due to creep, and interdiffusion of the fuel and fission products with the cladding. The aluminum-clad research reactor fuels are predominantly highly-enriched aluminum uranium alloy fuel which is clad with aluminum alloys similar to 1100, 5052, and 6061 aluminum. In the absence of corrodant species, degradation due to creep and diffusion mechanisms limit the maximum fuel storage temperature to 200 C. The results of laboratory scale corrosion tests indicate that this fuel could be stored under air up to 200 C at low relative humidity levels (< 20%) to limit corrosion of the cladding and fuel (exposed to the storage environment through assumed pre-existing pits in the cladding). Excessive degradation of fuels with uranium metal up to 200 C can be avoided if the fuel is sufficiently dried and contained in a sealed system; open storage can be achieved if the temperature is controlled to avoid excessive corrosion even in dry air.

  5. Finite Element Modeling for Wind Response of Aluminum Wall Cladding in Tall Building

    Directory of Open Access Journals (Sweden)

    Okafor Chinedum Vincent

    2017-09-01

    Full Text Available This paper analyzed wind loading responses of Aluminum Wall cladding panels in Tall Building using Ikeja Lagos State Nigeria. The wind loads were calculated according to Standard and Specification From BS6399-2:1997 Using the wind speed data of Lagos state Nigeria and finite element analysis, we predicted the responses of these Aluminum wall Cladding panels to the design wind loads being calculated. The result of the calculation from BS6399-2:1997 showed that the aluminum cladding panels located on the facade upwind was subject to positive pressure, which increases with height. Also, the cladding panels located on the leeward, as well as sidewalls, were subjected to negative pressure, which tended to be high at the top and bottom corners due to flow separation. From the result of the modeling and analysis, the researcher found out that stresses on the aluminum wall cladding panels were generally below the material yield point, showing that the high wind speed were not the reason for the collapse of aluminum cladding panels in the locality being considered. Instead, the reality lies on one or more issue on the materials construction and placement as discussed.This paper analyzed wind loading responses of Aluminum Wall cladding panels in Tall Building using Ikeja Lagos State Nigeria. The wind loads were calculated according to Standard and Specification From BS6399-2:1997 Using the wind speed data of Lagos state Nigeria and finite element analysis, we predicted the responses of these Aluminum wall Cladding panels to the design wind loads being calculated. The result of the calculation from BS6399-2:1997 showed that the aluminum cladding panels located on the facade upwind was subject to positive pressure, which increases with height. Also, the cladding panels located on the leeward, as well as sidewalls, were subjected to negative pressure, which tended to be high at the top and bottom corners due to flow separation. From the result of the

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

    DEFF Research Database (Denmark)

    Aggerbeck, Martin; Herbreteau, Alexis; Rombouts, Marleen

    2015-01-01

    Purpose - The purpose of this paper is to study the use of titanium as a protecting element for aluminum in alkaline conditions. Design/methodology/approach - Aluminum coatings containing up to 20 weight per cent Ti6Al4V were produced using laser cladding and were investigated using light optical...... microscope, scanning electron microscope - energy-dispersive X-ray spectroscopy and X-Ray Diffraction, together with alkaline exposure tests and potentiodynamic measurements at pH 13.5. Findings - Cladding resulted in a heterogeneous solidification microstructure containing an aluminum matrix...... 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...

  7. Foam coating on aluminum alloy with laser cladding

    NARCIS (Netherlands)

    Ocelik, V.; van Heeswijk, V.; de Hosson, J.T.M.; Csach, K.

    dThis article concentrates on the creation of a foam layer on an Al-Si substrate with laser technology. The cladding of At-Si powder in the front of a laser track has been separated from the side injection of mixture of Al-Si/TiH2 powder (foaming agent), which allows for fine tuning of the main

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

    International Nuclear Information System (INIS)

    Fernandes, Stela M.C.; Correa, Olandir V.; Souza, Jose A.; Ramanathan, Lalgudi V.; Antunes, Renato A.

    2013-01-01

    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)

  9. Physico Chemistry of the Chlorination of Aluminum Claddings in the Framework of HALOX Project

    International Nuclear Information System (INIS)

    Alvarez, Fabiola; De Micco, Georgina; Bohe, Ana; Pasquevich, Daniel

    2003-01-01

    The conditioning of spent nuclear fuels from test and research reactors requires a previous physicochemical treatment to stabilize them chemically.A possible way of processing is through what was called in CNEA as Process HALOX (Halogenation and Oxidation).It consists of the selective separation of cladding by halogenation and the subsequent oxidation of the core, previously to insert it into a vitreous matrix.The halogenation aim is to transform the constituents of the 6061aluminum alloy into volatile halides.In this work we present preliminary results of the chlorination of two aluminum alloys: AA 6061 and a type of CuZnAl alloy

  10. The corrosion of aluminum-clad spent nuclear fuel in wet basin storage

    International Nuclear Information System (INIS)

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

    1997-01-01

    This paper discusses the corrosion of the aluminum-clad spent fuel and the improvements that have been made in the SRS basins since 1993 which have essentially mitigated new corrosion on the fuel. It presents the results of a metallographic examination of two Mk-31A target slugs stored in the L-Reactor basin for about 5 years and a summary of results from the corrosion surveillance programs through 1996

  11. Transitioning aluminum clad spent fuels from wet to interim dry storage

    International Nuclear Information System (INIS)

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

    1994-01-01

    The United States Department of Energy (DOE) currently owns several hundred metric tons of aluminum clad, spent nuclear fuel and target assemblies. The vast majority of these irradiated assemblies are currently stored in water basins that were designed and operated for short term fuel cooling prior to fuel reprocessing. Recent DOE decisions to severely limit the reprocessing option have significantly lengthened the time of storage, thus increasing the tendency for corrosion induced degradation of the fuel cladding and the underlying core material. The portent of continued corrosion, coupled with the age of existing wet storage facilities and the cost of continuing basin operations, including necessary upgrades to meet current facility standards, may force the DOE to transition these wet stored, aluminum clad spent fuels to interim dry storage. The facilities for interim dry storage have not been developed, partially because fuel storage requirements and specifications for acceptable fuel forms are lacking. In spite of the lack of both facilities and specifications, current plans are to dry store fuels for approximately 40 to 60 years or until firm decisions are developed for final fuel disposition. The transition of the aluminum clad fuels from wet to interim dry storage will require a sequence of drying and canning operations which will include selected fuel preparations such as vacuum drying and conditioning of the storage atmosphere. Laboratory experiments and review of the available literature have demonstrated that successful interim dry storage may also require the use of fuel and canister cleaning or rinsing techniques that preclude, or at least minimize, the potential for the accumulation of chloride and other potentially deleterious ions in the dry storage environment. This paper summarizes an evaluation of the impact of fuel transitioning techniques on the potential for corrosion induced degradation of fuel forms during interim dry storage

  12. Ultimate disposition of aluminum clad spent nuclear fuel in the United States

    International Nuclear Information System (INIS)

    Messick, C.E.; Clark, W.D.; Clapper, M.; Mustin, T.P.

    2001-01-01

    Treatment and disposition of spent nuclear fuel (SNF) in the United States has changed significantly over the last decade due to change in world climate associated with nuclear material. Chemical processing of aluminum based SNF is ending and alternate disposition paths are being developed that will allow for the ultimate disposition of the enriched uranium in this SNF. Existing inventories of aluminum based SNF are currently being stored primarily in water-filled basins at the Savannah River Site (SRS) while these alternate disposition paths are being developed and implemented. Nuclear nonproliferation continues to be a worldwide concern and it is causing a significant influence on the development of management alternatives for SNF. SRS recently completed an environmental impact statement for the management of aluminum clad SNF that selects alternatives for all of the fuels in inventory. The U.S. Department of Energy and SRS are now implementing a dual strategy of processing small quantities of 'problematic' SNF while developing an alternative technology to dispose of the remaining aluminum clad SNF in the proposed monitored geologic repository. (author)

  13. Very large mode area ytterbium fiber amplifier with aluminum-doped pump cladding made by powder sinter technology

    International Nuclear Information System (INIS)

    He, Wenbin; Leich, Martin; Grimm, Stephan; Kobelke, Jens; Zhu, Yuan; Bartelt, Hartmut; Jäger, Matthias

    2015-01-01

    We demonstrate amplification experiments using a very large mode area Yb-doped double-clad fiber with 100 µm aluminum-cer codoped core and 440 µm pump cladding realized by high aluminum codoping. The material for core and pump cladding was fabricated by reactive powder sinter technology. A high numerical aperture (NA) of the pump cladding with NA = 0.21 and a low one of the core with NA = 0.084 could be realized. Using a 0.55 m short fiber sample as the main amplifier in a three-stage ns pulsed fiber master oscillator power amplifier system we achieved 3 ns, 2 mJ output pulses with 360 kW peak power limited by the available pump power. Stimulated Raman scattering effects and amplified spontaneous emission were successfully suppressed. (letter)

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

    Directory of Open Access Journals (Sweden)

    Rebak Raul B.

    2017-01-01

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

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

    Science.gov (United States)

    Rebak, Raul B.

    2017-12-01

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

  16. Acceptance criteria for interim dry storage of aluminum-clad fuels

    International Nuclear Information System (INIS)

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

    1994-01-01

    Direct repository disposal of foreign and domestic research reactor fuels owned by the United States Department of Energy is an alternative to reprocessing (together with vitrification of the high level waste and storage in an engineered barrier) for ultimate disposition. Neither the storage systems nor the requirements and specifications for acceptable forms for direct repository disposal have been developed; therefore, an interim storage strategy is needed to safely store these fuels. Dry storage (within identified limits) of the fuels received from wet-basin storage would avoid excessive degradation to assure post-storage handleability, a full range of ultimate disposal options, criticality safety, and provide for maintaining confinement by the fuel/clad system. Dry storage requirements and technologies for US commercial fuels, specifically zircaloy-clad fuels under inert cover gas, are well established. Dry storage requirements and technologies for a system with a design life of 40 years for dry storage of aluminum-clad foreign and domestic research reactor fuels are being developed by various groups within programs sponsored by the DOE

  17. The corrosion of aluminum-clad spent nuclear fuel in wet basin storage

    International Nuclear Information System (INIS)

    Howell, J.P.; Burke, S.D.

    1996-01-01

    Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980's and these fuels are caught in the pipeline awaiting stabilization or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced visible pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1996. This paper presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as discussions of fuel storage basins at other production sites of the Department of Energy

  18. The corrosion of aluminum-clad spent nuclear fuel in wet basin storage

    Energy Technology Data Exchange (ETDEWEB)

    Howell, J.P.; Burke, S.D.

    1996-02-20

    Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980`s and these fuels are caught in the pipeline awaiting stabilization or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced visible pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1996. This paper presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as discussions of fuel storage basins at other production sites of the Department of Energy.

  19. Corrosion of aluminum-clad alloys in wet spent fuel storage

    International Nuclear Information System (INIS)

    Howell, J.P.

    1995-09-01

    Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980's and these fuels are caught in the pipeline awaiting processing or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced significant pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1995, but the ultimate solution is to remove the fuel from the basins and to process it to a more stable form using existing and proven technology. This report presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as other fuel storage basins within the Department of Energy production sites

  20. Effect of Sr addition on the characteristics of as-cast and rolled 3003/4004 clad aluminum

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Guangyuan; Mao, Feng; Jie, Jinchuan; Cao, Zhiqiang, E-mail: caozq@dlut.edu.cn; Li, Tingju; Wang, Tongmin, E-mail: tmwang@dlut.edu.cn

    2016-09-05

    This paper examines the effects of Sr addition on the microstructure, composition distribution and Vickers hardness in the interfacial region of the as-cast and rolled 3003/4004 clad aluminum. The results reveal that the optimum adding amount of Sr on the as-cast Al-1.2Mn/Al−10Si-xSr clad is 0.08 wt%. With Sr content increasing from 0 to 0.08 wt%, the average length and number of the primary α-Al phase growing from the diffusion layer significantly decreased and whose morphology appears in columar dendritic crystals, the celluar dendrite crystals, deep celluar crystals, fine celluar crystals and planar crystals, while the dendritic-crystal primary α-Al phase nucleating and growing from inner Al−Si alloy side also show obvious decease in secondary dendrite spacing; meanwhile, eutectic Si phases were gradually modified from coarse plates, coralloid-plates mixed structure to fine branchy coralloid structure in three-dimensional morphology. After rolling, the diffusion layer thickness of the Al-1.2Mn/Al−10Si−0.08Sr clad is decreased by 66.7%, compared to that of unmodified clad alloy. This decreased diffusion layer thickness may be determined by augmented plastic strain and restraining diffusion of Si atoms in diffusion layer. Morever, average Vickers hardness on interface and Al−Si side of the Al-1.2Mn/Al−10Si−0.08Sr clad showed slight increase and more uniform distribution than that of unmodified clad alloy. This uniform distribution and improved hardness primarily attribute to presence of fine branchy coralloid silicon phase and its stronger dispersion strengthening as well as solution strengthening caused by interdiffusion of Si, Mn and Sr elements. - Highlights: • 3003/4004 clad aluminum was firstly modified by various Sr addition levels. • The optimum adding amount of Sr on the Al−1.2Mn/Al−10Si−xSr clad is 0.08 wt%. • Sr can refine primary α-Al and eutectic silicon phase of the clad simultaneously. • The Sr-modified rolled clad has

  1. Utilizing various test methods to study the stress corrosion behavior of Al-Li-Cu alloys

    Science.gov (United States)

    Pizzo, P. P.; Galvin, R. P.; Nelson, H. G.

    1984-01-01

    Recently, much attention has been given to aluminum-lithium alloys because of rather substantial specific-strength and specific-stiffness advantages offered over commercial 2000and 7000-series aluminum alloys. An obstacle to Al-Li alloy development has been inherent limited ductility. In order to obtain a more refined microstructure, powder metallurgy (P/M) has been employed in alloy development programs. As stress corrosion (SC) of high-strength aluminum alloys has been a major problem in the aircraft industry, the possibility of an employment of Al-Li alloys has been considered, taking into account a use of Al-Li-Cu alloys. Attention is given to a research program concerned with the evaluation of the relative SC resistance of two P/M processed Al-Li-Cu alloys. The behavior of the alloys, with and without an addition of magnesium, was studied with the aid of three test methods. The susceptibility to SC was found to depend on the microstructure of the alloys.

  2. Characteristics of copper-clad aluminum rods prepared by horizontal continuous casting

    Science.gov (United States)

    Zhang, Yubo; Fu, Ying; Jie, Jinchuan; Wu, Li; Svynarenko, Kateryna; Guo, Qingtao; Li, Tingju; Wang, Tongmin

    2017-11-01

    An innovative horizontal continuous casting method was developed and successfully used to prepare copper-clad aluminum (CCA) rods with a diameter of 85 mm and a sheath thickness of 16 mm. The solidification structure and element distribution near the interface of the CCA ingots were investigated by means of a scanning electron microscope, an energy dispersive spectrometer, and an electron probe X-ray microanalyzer. The results showed that the proposed process can lead to a good metallurgical bond between Cu and Al. The interface between Cu and Al was a multilayered structure with a thickness of 200 μm, consisting of Cu9Al4, CuAl2, α-Al/CuAl2 eutectic, and α-Al + α-Al/CuAl2 eutectic layers from the Cu side to the Al side. The mean tensile-shear strength of the CCA sample was 45 MPa, which fulfills the requirements for the further extrusion process. The bonding and diffusion mechanisms are also discussed in this paper.

  3. Prediction model for oxide thickness on aluminum alloy cladding during irradiation

    International Nuclear Information System (INIS)

    Kim, Yeon Soo; Hofman, G.L.; Hanan, N.A.; Snelgrove, J.L.

    2003-01-01

    An empirical model predicting the oxide film thickness on aluminum alloy cladding during irradiation has been developed as a function of irradiation time, temperature, heat flux, pH, and coolant flow rate. The existing models in the literature are neither consistent among themselves nor fit the measured data very well. They also lack versatility for various reactor situations such as a pH other than 5, high coolant flow rates, and fuel life longer than ∼1200 hrs. Particularly, they were not intended for use in irradiation situations. The newly developed model is applicable to these in-reactor situations as well as ex-reactor tests, and has a more accurate prediction capability. The new model demonstrated with consistent predictions to the measured data of UMUS and SIMONE fuel tests performed in the HFR, Petten, tests results from the ORR, and IRIS tests from the OSIRIS and to the data from the out-of-pile tests available in the literature as well. (author)

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-11-01

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

  6. Long term immersion test of aluminum alloy AA 6061 used for fuel cladding in MTR type reactors

    International Nuclear Information System (INIS)

    Linardi, Evelina M.; Rodriguez, Sebastian; Haddad, Roberto; Lanzani, Liliana

    2009-01-01

    In this work we present the results of long term immersion tests performed in the aluminum alloy AA 6061, used for fuel cladding in MTR type reactors. The tests were performed at open circuit potential in high purity water (ρ = 18.2 MΩ.cm) and in 10 -3 M NaCl solution. Two kinds of assemblies were studied: simple sheets and artificial crevices, immersed during 6, 12 and 18 months at room temperature. In both media and both assemblies, the aluminum hydroxide phases crystalline bayerite and bohemite were identified. It was found that a kind of localized attack named alkaline attack occurs around the iron-rich intermetallics. These particles were confirmed to control the corrosion of the AA 6061 alloy in an aerated medium. Immersion times for up to 18 months did not increase the oxide growth or the alkaline attack on the AA 6061 alloy. (author)

  7. Mercury-free dissolution of aluminum-clad fuel in nitric acid

    Science.gov (United States)

    Christian, Jerry D.; Anderson, Philip A.

    1994-01-01

    A mercury-free dissolution process for aluminum involves placing the aluminum in a dissolver vessel in contact with nitric acid-fluoboric acid mixture at an elevated temperature. By maintaining a continuous flow of the acid mixture through the dissolver vessel, an effluent containing aluminum nitrate, nitric acid, fluoboric acid and other dissolved components are removed.

  8. Effects of Thermal Exposure on Properties of Al-Li Alloys

    Science.gov (United States)

    Shah, Sandeep; Wells, Douglas; Stanton, William; Lawless, Kirby; Russell, Carolyn; Wagner, John; Domack, Marcia; Babel, Henry; Farahmand, Bahram; Schwab, David; hide

    2002-01-01

    Aluminum-Lithium (Al-Li) alloys offer significant performance benefits for aerospace structural applications due to their higher specific properties compared with conventional Al alloys. For example, the application of Al-Li alloy 2195 to the space shuffle external cryogenic fuel tank resulted in weight savings of over 7,000 lb, enabling successful deployment of International Space Station components. The composition and heat treatment of 2195 were optimized specifically for strength-toughness considerations for an expendable cryogenic tank. Time-dependent properties related to reliability, such as thermal stability, fatigue, and corrosion, will be of significant interest when materials are evaluated for a reusable cryotank structure. Literature surveys have indicated that there is limited thermal exposure data on Al-Li alloys. The effort reported here was designed to establish the effects of thermal exposure on the mechanical properties and microstructure of Al-Li alloys C458, L277, and 2195 in plate gages. Tensile, fracture toughness, and corrosion resistance were evaluated for both parent metal and friction stir welds (FSW) after exposure to temperatures as high as 300 F for up to 1000 hrs. Microstructural changes were evaluated with thermal exposure in order to correlate with the observed data trends. The ambient temperature parent metal data showed an increase in strength and reduction in elongation after exposure at lower temperatures. Strength reached a peak with intermediate temperature exposure followed by a decrease at highest exposure temperature. Friction stir welds of all alloys showed a drop in elongation with increased length of exposure. Understanding the effect of thermal exposure on the properties and microstructure of Al-Li alloys must be considered in defining service limiting temperatures and exposure times for a reusable cryotank structure.

  9. Processing and Characterization of Polycrystalline Yag (Yttrium Aluminum Garnet) Core-Clad Fibers - Postprint

    Science.gov (United States)

    2015-01-01

    agglomerates and contamination from the milling media. Binder and plasticizer were mixed with the Yb-doped YAG powder and the mixture was extruded through a...unlimited. 3.4 Scattering loss of polycrystalline YAG core-clad fiber and Yb excitation Spatially resolved optical scattering collected from the...which scattered light is collected and, while this could be used to improve our spatial resolution, it also greatly reduced the power levels incident

  10. On the occurrence of Portevin-Le Chatelier effect in fusion welded 2091 Al-Li based alloy joints

    Energy Technology Data Exchange (ETDEWEB)

    Vidal, A.C.; Darwish, F.A.; Solorzano, I.G. [Catholic Univ., Rio de Janeiro (Brazil)

    1995-09-01

    Al-Li based alloys are characterized by their lower density and higher stiffness, in comparison with conventional aluminum alloys. This makes the former very attractive for replacing the latter in structural and cryogenic applications, particularly in aeronautic and aerospace industries. The potential use of lithium-bearing also has stimulated studies on the weldability of these alloys as well as on the mechanical properties of the resulting welded joints. Al-Li-Cu-Mg alloy systems studied by Gomiero were found to exhibit serrations in their stress-strain curves, indicating the occurrence of Portevin-Le Chatelier (PLC) effects during plastic flow in these systems. The present study was therefore undertaken to determine the effect of fusion welding on the uniaxial tensile behavior of a 2091 (Al-Li-Cu-Mg-Zr) alloy. Microstructural aspects pertinent to the PLC effect are to be emphasized and the influence of post-weld heat treatment is to be presented and discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  12. Effects of Annealing Process on the Formability of Friction Stir Welded Al-Li Alloy 2195 Plates

    Science.gov (United States)

    Chen, Po-Shou; Bradford, Vann; Russell, Carolyn

    2011-01-01

    Large rocket cryogenic tank domes have typically been fabricated using Al-Cu based alloys like Al-Cu alloy 2219. The use of aluminum-lithium based alloys for rocket fuel tank domes can reduce weight because aluminum-lithium alloys have lower density and higher strength than Al-Cu alloy 2219. However, Al-Li alloys have rarely been used to fabricate rocket fuel tank domes because of the inherent low formability characteristic that make them susceptible to cracking during the forming operations. The ability to form metal by stretch forming or spin forming without excessive thinning or necking depends on the strain hardening exponent "n". The stain hardening exponent is a measure of how rapidly a metal becomes stronger and harder. A high strain hardening exponent is beneficial to a material's ability to uniformly distribute the imposed strain. Marshall Space Flight Center has developed a novel annealing process that can achieve a work hardening exponent on the order of 0.27 to 0.29, which is approximately 50% higher than what is typically obtained for Al-Li alloys using the conventional method. The strain hardening exponent of the Al-Li alloy plates or blanks heat treated using the conventional method is typically on the order of 0.17 to 0.19. The effects of this novel annealing process on the formability of friction stir welded Al-Li alloy blanks are being studied at Marshall Space Flight Center. The formability ratings will be generated using the strain hardening exponent, strain rate sensitivity and forming range. The effects of forming temperature on the formability will also be studied. The objective of this work is to study the deformation behavior of the friction stir welded Al-Li alloy 2195 blank and determine the formability enhancement by the new annealing process.

  13. Fracture Strength of AlLiB14

    Science.gov (United States)

    Wan, L. F.; Beckman, S. P.

    2012-10-01

    The orthorhombic boride crystal family XYB14, where X and Y are metal atoms, plays a critical role in a unique class of superhard compounds, yet there have been no studies aimed at understanding the origin of the mechanical strength of this compound. We present here the results from a comprehensive investigation into the fracture strength of the archetypal AlLiB14 crystal. First principles, ab initio, methods are used to determine the ideal brittle cleavage strength for several high-symmetry orientations. The elastic tensor and the orientation-dependent Young’s modulus are calculated. From these results the lower bound fracture strength of AlLiB14 is predicted to be between 29 and 31 GPa, which is near the measured hardness reported in the literature. These results indicate that the intrinsic strength of AlLiB14 is limited by the interatomic B-B bonds that span between the B layers.

  14. Modeling-Based Processing of Al-Li Alloys for Delamination Resistance Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Al-Li alloys are of interest for use in aerospace structures due to the desirable combination of high strength and low density. However, high strength Al-Li alloys...

  15. Microstructure and Mechanical Properties of an Al-Li-Mg-Sc-Zr Alloy Subjected to ECAP

    Directory of Open Access Journals (Sweden)

    Anna Mogucheva

    2016-10-01

    Full Text Available The effect of post-deformation solution treatment followed by water quenching and artificial aging on microstructure and mechanical properties of an Al-Li-Mg-Sc-Zr alloy subjected to equal-channel angular pressing (ECAP was examined. It was shown that the deformed microstructure produced by ECAP remains essentially unchanged under solution treatment. However, extensive grain refinement owing to ECAP processing significantly affects the precipitation sequence during aging. In the aluminum-lithium alloy with ultrafine-grained (UFG microstructure, the coarse particles of the S1-phase (Al2LiMg precipitate on high-angle boundaries; no formation of nanoscale coherent dispersoids of the δ′-phase (Al3Li occurs within grain interiors. Increasing the number of high-angle boundaries leads to an increasing portion of the S1-phase. As a result, no significant increase in strength occurs despite extensive grain refinement by ECAP.

  16. Calculation of Dingle temperatures and residual resistivity in dilute Al Li

    International Nuclear Information System (INIS)

    Cole, L.A.; Lawrence, W.E.

    1981-03-01

    We have calculated the Dingle temperatures and residual resistivity for dilute Al Li using a realistic band structure and Fermi surface for aluminum, and a self consistent calculation of the bound and scattering state wavefunctions of the lithium impurity. We have examined the effect of the position of the Li impurity, and found that substitutionally it is a strong S-wave scatterer, while interstitially it is a strong P-wave scatterer (each position results in two bound states). In the former case the Dingle temperatures vary considerably over the 3d zone Fermi surface; in the latter case they do not. The resistivities are similar for the two cases. We have also examined the effect of nonlocal exchange and correlation. The Dingle temperatures typically change by 10% or 20%, but their anisotropy is not affected significantly

  17. The physical metallurgy of mechanically-alloyed, dispersion-strengthened Al-Li-Mg and Al-Li-Cu alloys

    Science.gov (United States)

    Gilman, P. S.

    1984-01-01

    Powder processing of Al-Li-Mg and Al-Li-Cu alloys by mechanical alloying (MA) is described, with a discussion of physical and mechanical properties of early experimental alloys of these compositions. The experimental samples were mechanically alloyed in a Szegvari attritor, extruded at 343 and 427 C, and some were solution-treated at 520 and 566 C and naturally, as well as artificially, aged at 170, 190, and 210 C for times of up to 1000 hours. All alloys exhibited maximum hardness after being aged at 170 C; lower hardness corresponds to the solution treatment at 566 C than to that at 520 C. A comparison with ingot metallurgy alloys of the same composition shows the MA material to be stronger and more ductile. It is also noted that properly aged MA alloys can develop a better combination of yield strength and notched toughness at lower alloying levels.

  18. Evolution of the thickness of the aluminum oxide film due to the pH of the cooling water and surface temperature of the fuel elements clad of a nuclear reactor

    International Nuclear Information System (INIS)

    Babiche, Ivan

    2013-01-01

    This paper describes the mechanism of growth of a film of aluminum oxide on an alloy of the same material, which serves as a protective surface being the constituent material of the RP-10 nuclear reactor fuel elements clads. The most influential parameters on the growth of this film are: the pH of the cooling water and the clad surface temperature of the fuel element. For this study, a mathematical model relating the evolution of the aluminum oxide layer thickness over the time, according to the same oxide film using a power law is used. It is concluded that the time of irradiation, the heat flux at the surface of the aluminum material, the speed of the coolant, the thermal conductivity of the oxide, the initial thickness of the oxide layer and the solubility of the protective oxide are parameters affecting in the rate and film formation. (author).

  19. Development of the hardness-gradient blade material by aluminum cladding; Arumikuraddo ni yoru keisha kodo hamono zairyo no kaihatsu

    Energy Technology Data Exchange (ETDEWEB)

    Imai, J.; Hamada, T.; Yamada, S.; Takegawa, Y. [Matsushita Electric Industrial Co. Ltd., Kadoma, Osaka (Japan)] Yamada, H. [Daido Steel Co. Ltd., Nagoya (Japan)

    1998-06-20

    Both high speed drive of a blade and sharpening of the edge of the blade are important for improving the cutting performance of a shaver. An elongation percentage of 20% or greater is requisite for processing the blade and a high surface hardness after heat treatment is necessary for sharpening the edge of the blade. Therefore, an object of this research is to develop a blade material wherein the base material hardness is similar to that of quenched steel of Hv=5GP or greater and the surface hardness is Hv=10GPa or greater. The surface and base material-hardened blade material is obtained by performing heat treatment after the processing of a material with Al applied as cladding to both sides of F3-13CR-2Mo martensite stainless steel. In this material, Al diffuses from the surface to the inside of the material, FeAl-based intermetallic compound is formed in the surface, thus the hardness decreases gradiently toward the inside of the material. When the above-described material is applied to a shaver, the edge of the blade can be sharpened to have a curve radius of or smaller than 1{mu}m, and therefore a blade material with excellent cutting performance is accomplished. 4 refs., 7 figs.

  20. Thermal Exposure Effects on Properties of Al-Li Alloy Plate Products

    Science.gov (United States)

    Shah, Sandeep; Wells, Douglas; Wagner, John; Babel, Henry

    2003-01-01

    The objective of this viewgraph representation is to evaluate the effects of thermal exposure on the mechanical properties of both production mature and developmental Al-Li alloys. The researchers find for these alloys, the data clearly shows that there is no deficit in mechanical properties at lower exposure temperatures in some cases, and a signficant deficit in mechanical properties at higher exposure temperatures in all cases. Topics considered include: Al-Li alloys composition, key characteristics of Al-Li alloys and thermal exposure matrix.

  1. 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.

  2. Mechanical properties of ultra-fine grained structure formed in Al-Li alloys

    International Nuclear Information System (INIS)

    Adamczyk-Cieslak, B.; Lewandowska, M.; Mizera, J.; Kurzydlowski, K.J.

    2004-01-01

    This paper describes the mechanical properties (microhardness, yield stress) of two model Al-Li alloys by the Equal-Channel-Angular-Extrusion (ECAE) process. The applied ECAE process reduced the grain size from an initial value of ∼300 μm to a value of ∼0.7 μm leading to profound increase of plastic flow resistance. Such an increase is related to the grain size refinement and strengthening due to Li atoms in solid solution. Microhardness data confirm the Hall - Petch relation for grain sizes not available so far in Al-Li alloys. (author)

  3. Dislocation Dynamics in Al-Li Alloys. Mean Jump Distance and Activation Length of Moving Dislocations

    NARCIS (Netherlands)

    Hosson, J.Th.M. De; Huis in 't Veld, A.; Tamler, H.; Kanert, O.

    1984-01-01

    Pulsed nuclear magnetic resonance proved to be a complementary new technique for the study of moving dislocations in Al-Li alloys. The NMR technique, in combination with transmission electron microscopy and strain-rate change experiments have been applied to study dislocation motion in Al-2.2 wt% Li

  4. Cryogenic mechanical properties of low density superplastically formable Al-Li alloys

    Science.gov (United States)

    Verzasconi, S. L.; Morris, J. W., Jr.

    1989-01-01

    The aerospace industry is considering the use of low density, superplastically formable (SPF) materials, such as Al-Li alloys in cryogenic tankage. SPF modifications of alloys 8090, 2090, and 2090+In were tested for strength and Kahn tear toughness. The results were compared to those of similar tests of 2219-T87, an alloy currently used in cryogenic tankage, and 2090-T81, a recently studied Al-Li alloy with exceptional cryogenic properties (1-9). With decreasing temperature, all materials showed an increase in strength, while most materials showed an increase in elongation and decrease in Kahn toughness. The indium addition to 2090 increased alloy strength, but did not improve the strength-toughness combination. The fracture mode was predominantly intergranular along small, recrystallized grains, with some transgranular fracture, some ductile rupture, and some delamination on large, unrecrystallized grains.

  5. Stone cladding engineering

    CERN Document Server

    Sousa Camposinhos, Rui de

    2014-01-01

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

  6. Origin of unusual fracture in stirred zone for friction stir welded 2198-T8 Al-Li alloy joints

    International Nuclear Information System (INIS)

    Tao, Y.; Ni, D.R.; Xiao, B.L.; Ma, Z.Y.; Wu, W.; Zhang, R.X.; Zeng, Y.S.

    2017-01-01

    Friction stir welded (FSW) joints of conventional precipitation-hardened aluminum alloys usually fracture in the lowest hardness zone (LHZ) during tension testing. However, all of the FSW joints of a 2198-T8 Al-Li alloy fractured in the stirred zone (SZ) instead of the LHZ with the welding parameters of 800 rpm-200 mm/min and 1600 rpm-200 mm/min under the condition that no welding defects existed in the SZ. The experiment results revealed that lazy S was not the dominant factor resulting in the unusual fracture. The SZ consisted of three subzones, i.e., the shoulder-affected zone, the pin-affected zone, and the transition zone between them. While the former two zones were characterized by fine and equiaxed recrystallized grains, incompletely dynamically recrystallized microstructure containing coarse elongated non-recrystallized grains was observed in the transition zone. The transition zone exhibited the lowest average Taylor factor in the SZ, resulting in a region that was crystallographically weak. Furthermore, obvious lithium segregation at grain boundaries was observed in the transition zone via time-of-flight secondary ion mass spectroscopy analysis, but not in the shoulder-affected zone or the pin-affected zone. The combined actions of both the two factors resulted in the appearance of preferential intergranular fracture in the transition zone and eventually caused the failure in the SZ. The lithium segregation at grain boundaries in the transition zone was closely associated with both the segregation in the base material and the partially dynamically recrystallized microstructure resulting from the inhomogeneous plastic deformation in the SZ.

  7. Evolution of the internal friction in SIC particle reinforced 8090 Al-Li metal matrix composite

    International Nuclear Information System (INIS)

    Gutierrez-Urrutia, I.; Gallego, I.; No, M. L.; San Juan, J. M.

    2001-01-01

    The present study has been undertaken to investigate the mechanisms of thermal stress relief at the range of temperatures below room temperature for the metal matrix composite Al-Li 8090/SiC. For this aim the experimental technique of internal friction has been used which has been showed up very effective. Several thermal cycles from 453 K to 100 K were used in order to measures the internal friction as well as the elastic modules of the material concluding that thermal stresses are relaxed by microplastic deformation around the reinforcements. It has been also related the variation in the elastic modules with the different levels of precipitation. (Author) 18 refs

  8. X-ray online detection for laser welding T-joint of Al-Li alloy

    Science.gov (United States)

    Zhan, Xiaohong; Bu, Xing; Qin, Tao; Yu, Haisong; Chen, Jie; Wei, Yanhong

    2017-05-01

    In order to detect weld defects in laser welding T-joint of Al-Li alloy, a real-time X-ray image system is set up for quality inspection. Experiments on real-time radiography procedure of the weldment are conducted by using this system. Twin fillet welding seam radiographic arrangement is designed according to the structural characteristics of the weldment. The critical parameters including magnification times, focal length, tube current and tube voltage are studied to acquire high quality weld images. Through the theoretical and data analysis, optimum parameters are settled and expected digital images are captured, which is conductive to automatic defect detection.

  9. TEC – Thin Environmental Cladding

    Directory of Open Access Journals (Sweden)

    Alan Tomasi

    2015-05-01

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

  10. The stress-corrosion behavior of Al-Li-Cu alloys: A comparison of test methods

    Science.gov (United States)

    Rizzo, P. P.; Galvin, R. P.; Nelson, H. G.

    1982-01-01

    Two powder metallurgy processed (Al-Li-Cu) alloys with and without Mg addition were studied in aqueous 3.5% NaCl solution during the alternate immersion testing of tuning fork specimens, slow crack growth tests using fracture mechanics specimens, and the slow strain rate testing of straining electrode specimens. Scanning electron microscopy and optical metallography were used to demonstrate the character of the interaction between the Al-Li-Cu alloys and the selected environment. Both alloys are susceptible to SC in an aqueous 3.5% NaCl solution under the right electrochemical and microstructural conditions. Each test method yields important information on the character of the SC behavior. Under all conditions investigated, second phase particles strung out in rows along the extrusion direction in the alloys were rapidly attacked, and played principal role in the SC process. With time, larger pits developed from these rows of smaller pits and under certain electrochemical conditions surface cracks initiated from the larger pits and contributed directly to the fracture process. Evidence to support slow crack growth was observed in both the slow strain rate tests and the sustained immersion tests of precracked fracture mechanics specimens. The possible role of H2 in the stress corrosion cracking process is suggested.

  11. An investigation into the effect of equal channel angular extrusion process on mechanical and microstructural properties of middle layer in copper clad aluminum composite

    International Nuclear Information System (INIS)

    Tolaminejad, B.; Karimi Taheri, A.; Arabi, H.; Shahmiri, M.

    2009-01-01

    Equal channel angular extrusion is a promising technique for production of ultra fine-grain materials of few hundred nanometers size. In this research, the grain refinement of aluminium strip is accelerated by sandwiching it between two copper strips and then subjecting the three strips to Equal channel angular extrusion process simultaneously. The loosely packed copper-aluminium-copper laminated billet was passed through Equal channel angular extrusion die up to 8 passes using the Bc route. Then, tensile properties and some microstructural characteristics of the aluminium layer were evaluated. The scanning and transmission electron microscopes, and X-ray diffraction were used to characterize the microstructure. The results show that the yield stress of middle layer (Al) is increased significantly by about four times after application of Equal channel angular extrusion throughout the four consecutive passes and then it is slightly decreased when more Equal channel angular extrusion passes are applied. An ultra fine grain within the range of 500 to 600 nm was obtained in the Al layer by increasing the thickness of the copper layers. lt was observed that the reduction of grain size in the aluminium layer is nearly 55% more than that of a equal channel angular-extruded single layer aluminium billet, i.e. extruding a single aluminium strip or a billet without any clad for the same amount of deformation. This behaviour was attributed to the higher rates of dislocations interaction and cell formation and texture development during the Equal channel angular extrusion of the laminated composite compared to those of a single billet.

  12. 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)

  13. Structure and Mechanical Properties of Al-Li Alloys as Cast

    Directory of Open Access Journals (Sweden)

    Augustyn-Pieniążek J.

    2013-06-01

    Full Text Available The high mechanical properties of the Al-Li-X alloys contribute to their increasingly broad application in aeronautics, as an alternative for the aluminium alloys, which have been used so far. The aluminium-lithium alloys have a lower specific gravity, a higher nucleation and crack spread resistance, a higher Young’s module and they characterize in a high crack resistance at lower temperatures. The aim of the research planned in this work was to design an aluminium alloy with a content of lithium and other alloy elements. The research included the creation of a laboratorial melt, the microstructure analysis with the use of light microscopy, the application of X-ray methods to identify the phases existing in the alloy, and the microhardness test.

  14. Tool geometry optimization for drilling CFRP/Al-Li stacks with a lightning strike protection

    Science.gov (United States)

    El Bouami, Souhail; Habak, Malek; Velasco, Raphaël; Santos, Baptise Dos; Franz, Gérald; Vantomme, Pascal

    2017-10-01

    One-shot drilling of Carbon Fiber-Reinforced Polymer materials with a Lightning Strike Protection (LSP)/metal stacks is a challenging task due to the inherent difference physical and mechanical properties and processing mechanisms of each component. The objective of the present work is to optimize tool geometry width in drilling of CFRP/Al-Li with a LSP. Firstly, a set of conventional uncoated carbide drills which are commercially available for the drilling of aeronautic composites was used to study the effect of tool geometry on drilled-hole quality. The set encompasses a twist drill bit, a step drill bit and a point spur drill bit. Based on references and cutting conditions recommended by drill manufacturers, the drilling tests performed are based on full-factorial experimental design using three cutting speeds and two feed rates. Results showed that, on the one hand, spur drill gave the best results causing small damage extension in the hole perimeter but we noticed a rapid tool wear at the spur which increases with feed. On the other hand, step drill presented higher LSP delamination located at the hole entrance but reduces the level of thrust force. The choice of tool geometry process should be a compromise in drilling aluminium as well as drilling carbon fiber with LSP. In the second phase of the current work, three different new uncoated carbide geometries were developed: a Spur Step Drill, a Three Steps Drill and a Square Step Drill. Same cutting conditions were used for the three drills. Results showed a rapid tool wear for the Spur Step Drill at the spur. In terms of LSP delamination, burr and drill wear, the drill adapted to drilling CFRP/Al-Li with LSP stacks is the three steps drill.

  15. Stress Concentration and Fracture at Inter-variant Boundaries in an Al-Li Alloy

    Science.gov (United States)

    Crooks, Roy; Tayon, Wes; Domack, Marcia; Wagner, John; Beaudoin, Armand

    2009-01-01

    Delamination fracture has limited the use of lightweight Al-Li alloys. Studies of secondary, delamination cracks in alloy 2090, L-T fracture toughness samples showed grain boundary failure between variants of the brass texture component. Although the adjacent texture variants, designated B(sub s1) and B(sub s2), behave similarly during rolling, their plastic responses to mechanical tests can be quite different. EBSD data from through-thickness scans were used to generate Taylor factor maps. When a combined boundary normal and shear tensor was used in the calculation, the delaminating grains showed the greatest Taylor Factor differences of any grain pairs. Kernel Average Misorientation (KAM) maps also showed damage accumulation on one side of the interface. Both of these are consistent with poor slip accommodation from a crystallographically softer grain to a harder one. Transmission electron microscopy was used to confirm the EBSD observations and to show the role of slip bands in the development of large, interfacial stress concentrations. A viewgraph presentation accompanies the provided abstract.

  16. Effect of cold compression on precipitation and conductivity of an Al-Li-Cu alloy.

    Science.gov (United States)

    Khan, A K; Robinson, J S

    2008-12-01

    Transmission electron microscopy has been used to investigate the effect of increasing the degree of deformation applied by cold compression on the ageing kinetics and electrical conductivity response of an Al-Li-Cu alloy containing Mg and Ag. When cold compressed greater than 3%, the increased dislocation density accelerates the widespread precipitation of the T(1) phase resulting in an enhanced age hardening response. The lengthening rate of T(1) precipitates is also reduced in this cold compressed condition owing to the reduced local solute supersaturation, a result of the widespread precipitation of T(1) plates. Cold compression by less than 3% does not increase the age hardening response, and the precipitation of GP zones/theta'' appears to be suppressed. Precipitation of the T(1) phase is also not significantly enhanced compared with that of the more than 3% cold compressed conditions. The anomalous decrease in electrical conductivity is associated with the nucleation and growth of the T(1) phase. Strain fields around T(1) precipitates combined with the increased volume fraction of T(1) are thought to be the cause of the anomalous conductivity behaviour.

  17. Superplasticity and grain boundary character distribution in overaged Al-Li-Cu-Mg-Zr alloy

    International Nuclear Information System (INIS)

    Avramovic-Cingara, G.; Aust, K.T.; Perovic, D.D.; McQueen, H.J.

    1995-01-01

    Samples of 8091 alloy were subjected to a thermomechanical processing (TMP) treatment that included the following stages: overaging before deformation, multistage deformation at 300 deg C and strain rate change tests for superplasticity. Torsional deformation was utilized both to develop the refined microstructure and to test for superplasticity. The strain rate sensitivity, m, of the material ranged between 0.30 and 0.45 at 450 deg C for strain rates between 8 x 10 -2 and 10 -3 s -1 . The grain boundary character distribution (GBCD) of thermomechanically processed Al-Li-Cu-Mg-Zr (8091) alloy, which develops good superplastic response, has been determined by an electron backscattering diffraction technique (EBSD). All grain boundaries have been classified into one of three categories in terms of Σ values : low angle, coincidence site lattice and random high angle boundaries. Quantitative studies of grain boundary character were done after various processing stages to obtain evidence about structure evolution and indicate an increase in Σ boundary frequency following TMP. Selected area electron diffraction examination (SAD) gave evidence about the refined structure, in which the grain boundary misorientation increased EBSD how the grain boundary character was changed to high Σ values. TEM analyses indicate that the T 2 phase is responsible for substructure stabilization. There is no evidence of cavity formation during superplastic deformation by torsion, which suggests that cavity nucleation is strongly influenced by the nature of stress. (author). 32 refs., 3 tabs., 9 figs

  18. Removing hydrochloric acid exhaust products from high performance solid rocket propellant using aluminum-lithium alloy

    Energy Technology Data Exchange (ETDEWEB)

    Terry, Brandon C., E-mail: terry13@purdue.edu [School of Aeronautics and Astronautics, Purdue University, Zucrow Laboratories, 500 Allison Rd, West Lafayette, IN 47907 (United States); Sippel, Travis R. [Department of Mechanical Engineering, Iowa State University, 2025 Black Engineering, Ames, IA 50011 (United States); Pfeil, Mark A. [School of Aeronautics and Astronautics, Purdue University, Zucrow Laboratories, 500 Allison Rd, West Lafayette, IN 47907 (United States); Gunduz, I.Emre; Son, Steven F. [School of Mechanical Engineering, Purdue University, Zucrow Laboratories, 500 Allison Rd, West Lafayette, IN 47907 (United States)

    2016-11-05

    Highlights: • Al-Li alloy propellant has increased ideal specific impulse over neat aluminum. • Al-Li alloy propellant has a near complete reduction in HCl acid formation. • Reduction in HCl was verified with wet bomb experiments and DSC/TGA-MS/FTIR. - Abstract: Hydrochloric acid (HCl) pollution from perchlorate based propellants is well known for both launch site contamination, as well as the possible ozone layer depletion effects. Past efforts in developing environmentally cleaner solid propellants by scavenging the chlorine ion have focused on replacing a portion of the chorine-containing oxidant (i.e., ammonium perchlorate) with an alkali metal nitrate. The alkali metal (e.g., Li or Na) in the nitrate reacts with the chlorine ion to form an alkali metal chloride (i.e., a salt instead of HCl). While this technique can potentially reduce HCl formation, it also results in reduced ideal specific impulse (I{sub SP}). Here, we show using thermochemical calculations that using aluminum-lithium (Al-Li) alloy can reduce HCl formation by more than 95% (with lithium contents ≥15 mass%) and increase the ideal I{sub SP} by ∼7 s compared to neat aluminum (using 80/20 mass% Al-Li alloy). Two solid propellants were formulated using 80/20 Al-Li alloy or neat aluminum as fuel additives. The halide scavenging effect of Al-Li propellants was verified using wet bomb combustion experiments (75.5 ± 4.8% reduction in pH, ∝ [HCl], when compared to neat aluminum). Additionally, no measurable HCl evolution was detected using differential scanning calorimetry coupled with thermogravimetric analysis, mass spectrometry, and Fourier transform infrared absorption.

  19. Strain and strain-rate hardening characteristics of a superplastic Al-Li-Cu-Zr alloy

    International Nuclear Information System (INIS)

    Ash, B.A.; Hamilton, C.H.

    1988-01-01

    A number of alloys based on the composition of Al-Li-Zr have been shown to be superplastic under at least one of two different microstructural conditions: 1. fully recrystallized to a fine, stable grain size, and 2. warm- or cold-worked and unrecrystallized prior to superplastic deformation. For the latter case, static recrystallization was impaired by the presence of fine Al 3 Zr particles, and dynamic recrystallization was observed to occur during superplastic deformation in which the heavily worked microstructure evolved into a fine grained fully recrystallized microstructure. This process is observed in other Al alloys as well, such as the Al-Cu-Zr alloys (Supral alloys), Al-Zn-Mg-Zr alloys, Al-Mn-Zr alloys, and Al-Mg-Mn alloys where the dynamic recrystallization has been suggested to be a continuous reaction in which recrystallization occurs by a gradual and homogeneous process during deformation rather than by the more common nucleation and growth process. Experimental observations of continuous recrystallization show development of a subgrain structure which coarsens continuously while deformation proceeds, with a concurrent increase in the misorientation angle between adjacent subgrains which ultimately approaches that of a high-angle boundary, characteristic of a fully- recrystallized microstructure. During the first 50 to 300% deformation, the microstructure evolves from the heavily worked to a fully recrystallized microstructure after which the fully recrystallized microstructure apparently exhibits the typical micro-grain superplastic characteristics. Superplasticity under continuous dynamic recrystallization is of interest both from scientific and technological standpoints since the rates at which superplastic deformation can be obtained are often higher than those for the fully recrystallized microstructures

  20. TEC – Thin Environmental Cladding

    Directory of Open Access Journals (Sweden)

    Alan Tomasi

    2014-06-01

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

  1. Laser cladding with powder

    NARCIS (Netherlands)

    Schneider, M.F.; Schneider, Marcel Fredrik

    1998-01-01

    This thesis is directed to laser cladding with powder and a CO2 laser as heat source. The laser beam intensity profile turned out to be an important pa6 Summary rameter in laser cladding. A numerical model was developed that allows the prediction of the surface temperature distribution that is

  2. Evolution of the internal friction in SIC particle reinforced 8090 Al-Li metal matrix composite; Evolucion de la friccion interna del material compuesto de matriz Al-Li 8090 reforzado con particulas de SiC

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez-Urrutia, I.; Gallego, I.; No, M. L.; San Juan, J. M.

    2001-07-01

    The present study has been undertaken to investigate the mechanisms of thermal stress relief at the range of temperatures below room temperature for the metal matrix composite Al-Li 8090/SiC. For this aim the experimental technique of internal friction has been used which has been showed up very effective. Several thermal cycles from 453 K to 100 K were used in order to measures the internal friction as well as the elastic modules of the material concluding that thermal stresses are relaxed by microplastic deformation around the reinforcements. It has been also related the variation in the elastic modules with the different levels of precipitation. (Author) 18 refs.

  3. Removing hydrochloric acid exhaust products from high performance solid rocket propellant using aluminum-lithium alloy.

    Science.gov (United States)

    Terry, Brandon C; Sippel, Travis R; Pfeil, Mark A; Gunduz, I Emre; Son, Steven F

    2016-11-05

    Hydrochloric acid (HCl) pollution from perchlorate based propellants is well known for both launch site contamination, as well as the possible ozone layer depletion effects. Past efforts in developing environmentally cleaner solid propellants by scavenging the chlorine ion have focused on replacing a portion of the chorine-containing oxidant (i.e., ammonium perchlorate) with an alkali metal nitrate. The alkali metal (e.g., Li or Na) in the nitrate reacts with the chlorine ion to form an alkali metal chloride (i.e., a salt instead of HCl). While this technique can potentially reduce HCl formation, it also results in reduced ideal specific impulse (ISP). Here, we show using thermochemical calculations that using aluminum-lithium (Al-Li) alloy can reduce HCl formation by more than 95% (with lithium contents ≥15 mass%) and increase the ideal ISP by ∼7s compared to neat aluminum (using 80/20 mass% Al-Li alloy). Two solid propellants were formulated using 80/20 Al-Li alloy or neat aluminum as fuel additives. The halide scavenging effect of Al-Li propellants was verified using wet bomb combustion experiments (75.5±4.8% reduction in pH, ∝ [HCl], when compared to neat aluminum). Additionally, no measurable HCl evolution was detected using differential scanning calorimetry coupled with thermogravimetric analysis, mass spectrometry, and Fourier transform infrared absorption. Copyright © 2016 Elsevier B.V. All rights reserved.

  4. 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...

  5. Interactions between drops of molten Al-Li alloys and liquid water

    Energy Technology Data Exchange (ETDEWEB)

    Hyder, M.L. [Westinghouse Savannah River Co., Aiken, SC (United States); Nelson, L.S. [Sandia National Labs., Albuquerque, NM (United States); Duda, P.M.; Hyndman, D.A. [Ktech Corp., Albuquerque, NM (United States)

    1993-08-01

    Sandia National Laboratories, at the request of the Savannah River Technology Center (SRTC), studied the interactions between single drops of molten aluminum-lithium alloys and water. Most experiments were performed with ``B`` alloy (3.1 w/o Li, balance A1). Objectives were to develop experimental procedures for preparing and delivering the melt drops and diagnostics for characterizing the interactions, measure hydrogen generated by the reaction between melt and water, examine debris recovered after the interaction, determine changes in the aqueous phase produced by the melt-water chemical reactions, and determine whether steam explosions occur spontaneously under the conditions studied. Although many H{sub 2} bubbles were generated after the drops entered the water, spontaneous steam explosions never occurred when globules of the ``B`` alloy at temperatures between 700 and 1000C fell freely through water at room temperature, or upon or during subsequent contact with submerged aluminum or stainless steel surfaces. Total amounts of H{sub 2} (STP) increased from about 2 to 9 cm{sup 3}/per gram of melt as initial melt temperature increased over this range of temperatures.

  6. Initial Cladding Condition

    International Nuclear Information System (INIS)

    Siegmann, E.

    2000-01-01

    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 andO 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis will be used in

  7. Electra-Clad

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-05-04

    The study relates to the use of building-integrated photovoltaics. The Electra-Clad project sought to use steel-based cladding as a substrate for direct fabrication of a fully integrated solar panel of a design similar to the ICP standard glass-based panel. The five interrelated phases of the project are described. The study successfully demonstrated that the principles of the panel design are achievable and sound. But, despite intensive trials, a commercially realistic solar performance has not been achieved: the main failing was the poor solar conversion efficiency as the active area of the panel was increased in size. The problem lies with the coating used on the steel cladding substrates and it was concluded that a new type of coating will be required. ICP Solar Technologies UK carried out the work under contract to the DTI.

  8. Study on the characteristics and thermal stability of nanostructures in adiabatic shear band of 2195 Al-Li alloy

    Science.gov (United States)

    Yang, Yang; Chen, Yadong; Jiang, Lihong; Li, Meng; Zhang, Qingming; Tang, Tiegang

    2015-11-01

    Adiabatic shear bands (ASB) were obtained by dynamic shearing with a split Hopkinson pressure bar in the hat-shaped specimens of 2195-T6 Al-Li alloy. TEM observations reveal that grains in ASB are mainly equiaxed with the grain size from 50 to 100 nm. The kinetics possibility of instant refinement of grains can well be explained with the rotation dynamic recrystallization mechanism. EBSD is used to investigate microstructure evolution in ASB after annealed at 100-400 °C for 1 h. Results show that grain size increases rapidly at higher annealing temperature, and grains grow from 0.22 μm at 300 °C to 1.77 μm at 400 °C. Microhardness measurement indicated that the microhardness value rises slowly with temperature increases and then drops quickly at 300 °C. The study indicates that the nanostructure in ASB is thermally stable below 300 °C.

  9. Russian aluminum-lithium alloys for advanced reusable spacecraft

    International Nuclear Information System (INIS)

    Charette, Ray O.; Leonard, Bruce G.; Bozich, William F.; Deamer, David A.

    1998-01-01

    Cryotanks that are cost-affordable, robust, fuel-compatible, and lighter weight than current aluminum design are needed to support next-generation launch system performance and operability goals. The Boeing (McDonnell Douglas Aerospace-MDA) and NASA's Delta Clipper-Experimental Program (DC-XA) flight demonstrator test bed vehicle provided the opportunity for technology transfer of Russia's extensive experience base with weight-efficient, highly weldable aluminum-lithium (Al-Li) alloys for cryogenic tank usage. As part of NASA's overall reusable launch vehicle (RLV) program to help provide technology and operations data for use in advanced RLVs, MDA contracted with the Russian Academy of Sciences (RAS/IMASH) for design, test, and delivery of 1460 Al-Li alloy liquid oxygen (LO 2 ) cryotanks: one for development, one for ground tests, and one for DC-XA flight tests. This paper describes the development of Al-Li 1460 alloy for reusable LO 2 tanks, including alloy composition tailoring, mechanical properties database, forming, welding, chemical milling, dissimilar metal joining, corrosion protection, completed tanks proof, and qualification testing. Mechanical properties of the parent and welded materials exceeded expectations, particularly the fracture toughness, which promise excellent reuse potential. The LO 2 cryotank was successfully demonstrated in DC-XA flight tests

  10. Temperature and humidity effects on the corrosion of aluminium-base reactor fuel cladding materials during dry storage

    International Nuclear Information System (INIS)

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

    2004-01-01

    The effect of temperature and relative humidity on the high temperature (up to 200 deg. C) corrosion of aluminum cladding alloys was investigated for dry storage of spent nuclear fuels. A dependency on alloy type and temperature was determined for saturated water vapor conditions. Models were developed to allow prediction of cladding behaviour of 1100, 5052, and 6061 aluminum alloys for up to 50+ years at 100% relative humidity. Calculations show that for a closed system, corrosion stops after all moisture and oxygen is used up during corrosion reactions with aluminum alloys. (author)

  11. Aerogel-clad optical fiber

    Science.gov (United States)

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

    1997-01-01

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

  12. Aluminum nitride grating couplers.

    Science.gov (United States)

    Ghosh, Siddhartha; Doerr, Christopher R; Piazza, Gianluca

    2012-06-10

    Grating couplers in sputtered aluminum nitride, a piezoelectric material with low loss in the C band, are demonstrated. Gratings and a waveguide micromachined on a silicon wafer with 600 nm minimum feature size were defined in a single lithography step without partial etching. Silicon dioxide (SiO(2)) was used for cladding layers. Peak coupling efficiency of -6.6 dB and a 1 dB bandwidth of 60 nm have been measured. This demonstration of wire waveguides and wideband grating couplers in a material that also has piezoelectric and elasto-optic properties will enable new functions for integrated photonics and optomechanics.

  13. Stone cladding engineering

    National Research Council Canada - National Science Library

    Camposinhos, Rui de Sousa

    2014-01-01

    ...) for the most commonly used modern anchorage systems. Calculation examples allow designers to solve several anchorage engineering problems in a detailed and objective manner, underlining the key parameters. The design of the anchorage metal parts, either in stainless steel or aluminum, is also presented.

  14. Corrosion of aluminum alloys in a reactor disassembly basin

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-12-01

    This document discusses storage of aluminum clad fuel and target tubes of the Mark 22 assembly takes place in the concrete-lined, light-water-filled, disassembly basins located within each reactor area at the Savannah River Site (SRS). A corrosion test program has been conducted in the K-Reactor disassembly basin to assess the storage performance of the assemblies and other aluminum clad components in the current basin environment. Aluminum clad alloys cut from the ends of actual fuel and target tubes were originally placed in the disassembly water basin in December 1991. After time intervals varying from 45--182 days, the components were removed from the basin, photographed, and evaluated metallographically for corrosion performance. Results indicated that pitting of the 8001 aluminum fuel clad alloy exceeded the 30-mil (0.076 cm) cladding thickness within the 45-day exposure period. Pitting of the 1100 aluminum target clad alloy exceeded the 30-mil (0.076 cm) clad thickness in 107--182 days exposure. The existing basin water chemistry is within limits established during early site operations. Impurities such as Cl{sup {minus}}, NO{sub 3}{sup {minus}} and SO{sub 4}{sup {minus}} are controlled to the parts per million level and basin water conductivity is currently 170--190 {mu}mho/cm. The test program has demonstrated that the basin water is aggressive to the aluminum components at these levels. Other storage basins at SRS and around the US have successfully stored aluminum components for greater than ten years without pitting corrosion. These basins have impurity levels controlled to the parts per billion level (1000X lower) and conductivity less than 1.0 {mu}mho/cm.

  15. Aluminum Nanoholes for Optical Biosensing

    Directory of Open Access Journals (Sweden)

    Carlos Angulo Barrios

    2015-07-01

    Full Text Available Sub-wavelength diameter holes in thin metal layers can exhibit remarkable optical features that make them highly suitable for (biosensing applications. Either as efficient light scattering centers for surface plasmon excitation or metal-clad optical waveguides, they are able to form strongly localized optical fields that can effectively interact with biomolecules and/or nanoparticles on the nanoscale. As the metal of choice, aluminum exhibits good optical and electrical properties, is easy to manufacture and process and, unlike gold and silver, its low cost makes it very promising for commercial applications. However, aluminum has been scarcely used for biosensing purposes due to corrosion and pitting issues. In this short review, we show our recent achievements on aluminum nanohole platforms for (biosensing. These include a method to circumvent aluminum degradation—which has been successfully applied to the demonstration of aluminum nanohole array (NHA immunosensors based on both, glass and polycarbonate compact discs supports—the use of aluminum nanoholes operating as optical waveguides for synthesizing submicron-sized molecularly imprinted polymers by local photopolymerization, and a technique for fabricating transferable aluminum NHAs onto flexible pressure-sensitive adhesive tapes, which could facilitate the development of a wearable technology based on aluminum NHAs.

  16. Aluminum Nanoholes for Optical Biosensing.

    Science.gov (United States)

    Barrios, Carlos Angulo; Canalejas-Tejero, Víctor; Herranz, Sonia; Urraca, Javier; Moreno-Bondi, María Cruz; Avella-Oliver, Miquel; Maquieira, Ángel; Puchades, Rosa

    2015-07-09

    Sub-wavelength diameter holes in thin metal layers can exhibit remarkable optical features that make them highly suitable for (bio)sensing applications. Either as efficient light scattering centers for surface plasmon excitation or metal-clad optical waveguides, they are able to form strongly localized optical fields that can effectively interact with biomolecules and/or nanoparticles on the nanoscale. As the metal of choice, aluminum exhibits good optical and electrical properties, is easy to manufacture and process and, unlike gold and silver, its low cost makes it very promising for commercial applications. However, aluminum has been scarcely used for biosensing purposes due to corrosion and pitting issues. In this short review, we show our recent achievements on aluminum nanohole platforms for (bio)sensing. These include a method to circumvent aluminum degradation--which has been successfully applied to the demonstration of aluminum nanohole array (NHA) immunosensors based on both, glass and polycarbonate compact discs supports--the use of aluminum nanoholes operating as optical waveguides for synthesizing submicron-sized molecularly imprinted polymers by local photopolymerization, and a technique for fabricating transferable aluminum NHAs onto flexible pressure-sensitive adhesive tapes, which could facilitate the development of a wearable technology based on aluminum NHAs.

  17. Aluminum Nanoholes for Optical Biosensing

    Science.gov (United States)

    Barrios, Carlos Angulo; Canalejas-Tejero, Víctor; Herranz, Sonia; Urraca, Javier; Moreno-Bondi, María Cruz; Avella-Oliver, Miquel; Maquieira, Ángel; Puchades, Rosa

    2015-01-01

    Sub-wavelength diameter holes in thin metal layers can exhibit remarkable optical features that make them highly suitable for (bio)sensing applications. Either as efficient light scattering centers for surface plasmon excitation or metal-clad optical waveguides, they are able to form strongly localized optical fields that can effectively interact with biomolecules and/or nanoparticles on the nanoscale. As the metal of choice, aluminum exhibits good optical and electrical properties, is easy to manufacture and process and, unlike gold and silver, its low cost makes it very promising for commercial applications. However, aluminum has been scarcely used for biosensing purposes due to corrosion and pitting issues. In this short review, we show our recent achievements on aluminum nanohole platforms for (bio)sensing. These include a method to circumvent aluminum degradation—which has been successfully applied to the demonstration of aluminum nanohole array (NHA) immunosensors based on both, glass and polycarbonate compact discs supports—the use of aluminum nanoholes operating as optical waveguides for synthesizing submicron-sized molecularly imprinted polymers by local photopolymerization, and a technique for fabricating transferable aluminum NHAs onto flexible pressure-sensitive adhesive tapes, which could facilitate the development of a wearable technology based on aluminum NHAs. PMID:26184330

  18. Ice-clad volcanoes

    Science.gov (United States)

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

    2015-01-01

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

  19. Metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, Nina

    by Qi et al [Zm Qi et al, Sens. Actuators B 81, 2002] before, however the sensing principle we present results in a broad detection range from gasses to solid materials and is different from the principle suggested by Qi et al with a highlylimited detection range. Metal-clad waveguide sensors......, where single cell detection isshown by use of the metal-clad waveguide sensors.......This work concerns planar optical waveguide sensors for biosensing applications, with the focus on deep-probe sensing for micron-scale biological objects like bacteria and whole cells. In the last two decades planar metal-clad waveguides have been brieflyintroduced in the literature applied...

  20. Explosive Cladding of Titanium and Aluminium Alloys on the Example of Ti6Al4V-AA2519 Joints / Wybuchowe Platerowanie Stopów Tytanu I Aluminium Na Przykładzie Połączenia Ti6Al4V-AA2519

    Directory of Open Access Journals (Sweden)

    Gałka A.

    2015-12-01

    Full Text Available Explosive cladding is currently one of the basic technologies of joining metals and their alloys. It enables manufacturing of the widest range of joints and in many cases there is no alternative solution. An example of such materials are clads that include light metals such as titanium and aluminum. ach new material combination requires an appropriate adaptation of the technology by choosing adequate explosives and tuning other cladding parameters. Technology enabling explosive cladding of Ti6Al4V titanium alloy and aluminum AA2519 was developed. The clads were tested by means of destructive and nondestructive testing, analyzing integrity, strength and quality of the obtained joint.

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

    Science.gov (United States)

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

    1982-01-01

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

  2. Aluminum: Reflective Aluminum Chips

    Energy Technology Data Exchange (ETDEWEB)

    Recca, L.

    1999-01-29

    This fact sheet reveals how the use of reflective aluminum chips on rooftops cuts down significantly on heat absorption, thus decreasing the need for air conditioning. The benefits, including energy savings that could reach the equivalent of 1.3 million barrels of oil annually for approximately 100,000 warehouses, are substantial.

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

    Energy Technology Data Exchange (ETDEWEB)

    Brian K Castle

    2012-05-01

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

  4. Laser-induced reversion of δ′ precipitates in an Al-Li alloy: Study on temperature rise in pulsed laser atom probe

    KAUST Repository

    Khushaim, Muna Saeed Amin

    2016-06-14

    The influence of tuning the laser pulse energy during the analyses on the resulting microstructure in a specimen utilizing an ultra-fast laser assisted atom probe was demonstrated by a case study of a binary Al-Li alloy. The decomposition parameters, such as the size, number density, volume fraction, and composition of δ\\' precipitates, were carefully monitored after each analysis. A simple model was employed to estimate the corresponding specimen temperature for each value of the laser energy. The results indicated that the corresponding temperatures for the laser pulse energy in the range of 10 to 80 pJ are located inside the miscibility gap of the binary Al-Li phase diagram and fall into the metastable equilibrium field. In addition, the corresponding temperature for a laser pulse energy of 100 pJ was in fairly good agreement with reported range of δ\\' solvus temperature, suggesting a result of reversion upon heating due to laser pulsing. © 2016 Wiley Periodicals, Inc.

  5. Fine Structure of Diffuse Scattering Rings in Al-Li-Cu Quasicrystal: A Comparative X-ray and Electron Diffraction Study

    Science.gov (United States)

    Donnadieu, P.; Dénoyer, F.

    1996-11-01

    A comparative X-ray and electron diffraction study has been performed on Al-Li-Cu icosahedral quasicrystal in order to investigate the diffuse scattering rings revealed by a previous work. Electron diffraction confirms the existence of rings but shows that the rings have a fine structure. The diffuse aspect on the X-ray diffraction patterns is then due to an averaging effect. Recent simulations based on the model of canonical cells related to the icosahedral packing give diffractions patterns in agreement with this fine structure effect. Nous comparons les diagrammes de diffraction des rayon-X et des électrons obtenus sur les mêmes échantillons du quasicristal icosaèdrique Al-Li-Cu. Notre but est d'étudier les anneaux de diffusion diffuse mis en évidence par un travail précédent. Les diagrammes de diffraction électronique confirment la présence des anneaux mais ils montrent aussi que ces anneaux possèdent une structure fine. L'aspect diffus des anneaux révélés par la diffraction des rayons X est dû à un effet de moyenne. Des simulations récentes basées sur la décomposition en cellules canoniques de l'empilement icosaédrique produisent des diagrammes de diffraction en accord avec ces effects de structure fine.

  6. Clad Degradation - FEPs Screening Arguments

    International Nuclear Information System (INIS)

    E. Siegmann

    2004-01-01

    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. Aluminum Hydroxide

    Science.gov (United States)

    Aluminum hydroxide is used for the relief of heartburn, sour stomach, and peptic ulcer pain and to ... Aluminum hydroxide comes as a capsule, a tablet, and an oral liquid and suspension. The dose and ...

  8. Development of Aluminum-Lithium 2195 Gores by the Stretch Forming Process

    Science.gov (United States)

    Volz, M. P.; Chen, P. S.; Gorti, S.; Salvail, P.

    2014-01-01

    Aluminum-Lithium alloy 2195 exhibits higher mechanical properties and lower density than aluminum alloy 2219, which is the current baseline material for Space Launch System (SLS) cryogenic tank components. Replacement of Al 2219 with Al-Li 2195 would result in substantial weight savings, as was the case when this replacement was made on the shuttle external tank. A key component of cryogenic tanks are the gores, which are welded together to make the rounded ends of the tanks. The required thicknesses of these gores depend on the specific SLS configuration and may exceed the current experience base in the manufacture of such gores by the stretch forming process. Here we describe the steps taken to enhance the formability of Al-Li 2195 by optimizing the heat treatment and stretch forming processes for gore thicknesses up to 0.75", which envelopes the maximum expected gore thicknesses for SLS tanks. An annealing treatment, developed at Marshall Space Flight Center, increased the forming range and strain hardening exponent of Al-Li 2195 plates. Using this annealing treatment, one 0.525" thick and two 0.75" thick gores were manufactured by the stretch forming process. The annealing treatment enabled the stretch forming of the largest ever cross sectional area (thickness x width) of an Al-Li 2195 plate achieved by the manufacturer. Mechanical testing of the gores showed greater than expected ultimate tensile strength, yield strength, modulus, and elongation values. The gores also exhibited acceptable fracture toughness at room and LN2 temperatures. All of the measured data indicate that the stretch formed gores have sufficient material properties to be used in flight domes.

  9. Development of high performance cladding

    International Nuclear Information System (INIS)

    Kiuchi, Kiyoshi

    2003-01-01

    The developments of superior next-generation light water reactor are requested on the basis of general view points, such as improvement of safety, economics, reduction of radiation waste and effective utilization of plutonium, until 2030 year in which conventional reactor plants should be renovate. Improvements of stainless steel cladding for conventional high burn-up reactor to more than 100 GWd/t, developments of manufacturing technology for reduced moderation-light water reactor (RMWR) of breeding ratio beyond 1.0 and researches of water-materials interaction on super critical pressure-water cooled reactor are carried out in Japan Atomic Energy Research Institute. Stable austenite stainless steel has been selected for fuel element cladding of advanced boiling water reactor (ABWR). The austenite stain less has the superiority for anti-irradiation properties, corrosion resistance and mechanical strength. A hard spectrum of neutron energy up above 0.1 MeV takes place in core of the reduced moderation-light water reactor, as liquid metal-fast breeding reactor (LMFBR). High performance cladding for the RMWR fuel elements is required to get anti-irradiation properties, corrosion resistance and mechanical strength also. Slow strain rate test (SSRT) of SUS 304 and SUS 316 are carried out for studying stress corrosion cracking (SCC). Irradiation tests in LMFBR are intended to obtain irradiation data for damaged quantity of the cladding materials. (M. Suetake)

  10. Evaluation of water chemistry on the pitting susceptibility of aluminum

    International Nuclear Information System (INIS)

    Chandler, G.T.; Sindelar, R.L.; Lam, P.S.

    1997-01-01

    Aluminum-clad spent nuclear fuels are being stored in water in the Receiving Basin for Off-site Fuels (RBOF) and the reactor disassembly (cooling) basins at the Savannah River Site (SRS). Experience shows that fuels stored in water are subject to rapid pitting corrosion if the water quality is poor. Upgrade projects and actions, including those to improve water quality, were recently undertaken to upgrade the disassembly basins for extended storage. A technical strategy was developed for continued basin storage of aluminum-clad fuel assemblies. The strategy includes development and implementation of basin technical standards for water quality to minimize attack due to pitting corrosion over a desired storage period. In the absence of localized corrosion, only slow, general corrosion of the cladding would be expected. A laboratory corrosion program is being performed to provide the bases for technical standards by identifying the region of aggressive water qualities where existing oxide films would tend to break down and pits would initiate and remain active. Initial results from corrosion potential and cyclic polarization testing of aluminum alloys in various water chemistries have shown that low conductivity water (< 50 μS/cm) should not be aggressive to cause self-pitting corrosion. Initial results from tests of 8001 and 5052 aluminum and aluminium-10% uranium alloy indicate that a strong galvanic couple should not exist between the aluminum cladding materials and the aluminum-uranium fuel. Additional laboratory testing will include immersion testing to allow characterization of the growth rate of active pits to benchmark a kinetic model. This model will form the basis for a water quality technical standard and enable prediction of the life of aluminum-clad spent nuclear fuels in basin storage

  11. Evaluation of water chemistry on the pitting susceptibility of aluminum

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, G.T.; Sindelar, R.L.; Lam, P.S. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1997-12-01

    Aluminum-clad spent nuclear fuels are being stored in water in the Receiving Basin for Off-site Fuels (RBOF) and the reactor disassembly (cooling) basins at the Savannah River Site (SRS). Experience shows that fuels stored in water are subject to rapid pitting corrosion if the water quality is poor. Upgrade projects and actions, including those to improve water quality, were recently undertaken to upgrade the disassembly basins for extended storage. A technical strategy was developed for continued basin storage of aluminum-clad fuel assemblies. The strategy includes development and implementation of basin technical standards for water quality to minimize attack due to pitting corrosion over a desired storage period. In the absence of localized corrosion, only slow, general corrosion of the cladding would be expected. A laboratory corrosion program is being performed to provide the bases for technical standards by identifying the region of aggressive water qualities where existing oxide films would tend to break down and pits would initiate and remain active. Initial results from corrosion potential and cyclic polarization testing of aluminum alloys in various water chemistries have shown that low conductivity water (< 50 {micro}S/cm) should not be aggressive to cause self-pitting corrosion. Initial results from tests of 8001 and 5052 aluminum and aluminium-10% uranium alloy indicate that a strong galvanic couple should not exist between the aluminum cladding materials and the aluminum-uranium fuel. Additional laboratory testing will include immersion testing to allow characterization of the growth rate of active pits to benchmark a kinetic model. This model will form the basis for a water quality technical standard and enable prediction of the life of aluminum-clad spent nuclear fuels in basin storage.

  12. Study of laser cladding nuclear valve parts

    International Nuclear Information System (INIS)

    Shi Shihong; Wang Xinlin; Huang Guodong

    1998-12-01

    The mechanism of laser cladding is discussed by using heat transfer model of laser cladding, heat conduction model of laser cladding and convective transfer mass model of laser melt-pool. Subsequently the laser cladding speed limit and the influence of laser cladding parameters on cladding layer structure is analyzed. A 5 kW with CO 2 transverse flow is used in the research for cladding treatment of sealing surface of stop valve parts of nuclear power stations. The laser cladding layer is found to be 3.0 mm thick. The cladding surface is smooth and has no such defects as crack, gas pore, etc. A series of comparisons with plasma spurt welding and arc bead welding has been performed. The results show that there are higher grain grade and hardness, lower dilution and better performances of resistance to abrasion, wear and of anti-erosion in the laser cladding layer. The new technology of laser cladding can obviously improve the quality of nuclear valve parts. Consequently it is possible to lengthen the service life of nuclear valve and to raise the safety and reliability of the production system

  13. Aluminum Target Dissolution in Support of the Pu-238 Program

    Energy Technology Data Exchange (ETDEWEB)

    McFarlane, Joanna [ORNL; Benker, Dennis [ORNL; DePaoli, David W [ORNL; Felker, Leslie Kevin [ORNL; Mattus, Catherine H [ORNL

    2014-09-01

    Selection of an aluminum alloy for target cladding affects post-irradiation target dissolution and separations. Recent tests with aluminum alloy 6061 yielded greater than expected precipitation in the caustic dissolution step, forming up to 10 wt.% solids of aluminum hydroxides and aluminosilicates. We present a study to maximize dissolution of aluminum metal alloy, along with silicon, magnesium, and copper impurities, through control of temperature, the rate of reagent addition, and incubation time. Aluminum phase transformations have been identified as a function of time and temperature, using X-ray diffraction. Solutions have been analyzed using wet chemical methods and X-ray fluorescence. These data have been compared with published calculations of aluminum phase diagrams. Temperature logging during the transients has been investigated as a means to generate kinetic and mass transport data on the dissolution process. Approaches are given to enhance the dissolution of aluminum and aluminosilicate phases in caustic solution.

  14. Study of the density of electrons in momentum space in the Al-Li-Cu icosahedral phase by means of positron annihilation

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Yoshikazu; Nanao, Susumu [Institute of Industrial Science, The University of Tokyo, Roppongi, Minato, Tokyo 106 (Japan); Tanigawa, Shoichiro [Institute of Materials Science, University of Tsukuba, Tsukuba, Ibaraki 305 (Japan)

    1997-12-15

    The three-dimensional momentum density of annihilating electron - positron pairs has been studied for a single Al-Li-Cu icosahedral quasicrystal. A direct Fourier transform method is employed to reconstruct the three-dimensional momentum density from measurements of the two-dimensional angular correlation of positron annihilation radiation (2 D-ACAR). The crystallographic anisotropy in the momentum density is observed to be very small. The asphericity of the Fermi surface is not found explicitly within the experimental resolution in the momentum space. The features of the three-dimensional electron - positron momentum density agree with those obtained by means of Compton profile measurement. It is suggested that a strong lattice - electron interaction at the Fermi level occurs in this icosahedral phase. (author)

  15. Coincidence Doppler broadening and 3DAP study of the pre-precipitation stage of an Al-Li-Cu-Mg-Ag alloy

    International Nuclear Information System (INIS)

    Honma, T.; Yanagita, S.; Hono, K.; Nagai, Y.; Hasegawa, M.

    2004-01-01

    Pre-precipitation solute clustering in Al-Li-Cu-Mg-Ag and Al-Cu-Mg-Ag alloys has been investigated by coincidence Doppler broadening (CDB) spectroscopy of positron annihilation and three-dimensional atom probe (3DAP) analysis. Although Ag-Mg co-clusters form in the Al-Cu-Mg-Ag alloy in the early stage of aging, no evidence for the co-cluster formation was obtained from the Li containing alloy using 3DAP. While CDB spectra indicated that vacancies are associated with Ag after aging for 15 s in the Al-Cu-Mg-Ag alloy, vacancy-Ag association is suppressed in the Li containing alloy. Based on the 3DAP and CDB results, the reasons for the completely different clustering behaviors observed in these two similar alloys are discussed

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-03-02

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

  18. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    International Nuclear Information System (INIS)

    R. Schreiner

    2004-01-01

    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

  19. Fracture Toughness Of Zircaloy Claddings

    International Nuclear Information System (INIS)

    Bertsch, J.; Hoffelner, W

    2003-01-01

    Zirconium-based alloys (Zircaloy) have been used as cladding material in Light Water Reactors for many years. During fabrication, or in in-reactor service, crack-type defects can be formed, posing questions regarding mechanical integrity. As claddings change their mechanical properties (mainly toughness) during service as a result of irradiation-induced degradation, oxidation and hydride formation, it is essential for integrity considerations to provide parameters for the assessment of the influence of flaws on rupture behaviour. Usually, fracture-mechanics parameters are employed such as the fracture toughness, K IC , or, for high plastic strains, the J-integral, JIC. The applicability of these parameters is, however, limited by the dimensions of the samples (e.g. thickness). In claddings with a wall thickness of below 1 mm, determination of toughness necessitates an extension of the J-integral concept. A method based on the traditional J-approach, but applicable to thin-walled structures, is presented in this paper. (author)

  20. LASER SURFACE CLADDING FOR STRUCTURAL REPAIR

    OpenAIRE

    SANTANU PAUL

    2018-01-01

    Laser cladding is a powder deposition technique, which is used to deposit layers of clad material on a substrate to improve its surface properties. It has widespread application in the repair of dies and molds used in the automobile industry. These molds and dies are subjected to cyclic thermo-mechanical loading and therefore undergo localized damage and wear. The final clad quality and integrity is influenced by various physical phenomena, namely, melt pool morphology, microst...

  1. Pulsed Laser Cladding of Ni Based Powder

    Science.gov (United States)

    Pascu, A.; Stanciu, E. M.; Croitoru, C.; Roata, I. C.; Tierean, M. H.

    2017-06-01

    The aim of this paper is to optimize the operational parameters and quality of one step Metco Inconel 718 atomized powder laser cladded tracks, deposited on AISI 316 stainless steel substrate by means of a 1064 nm high power pulsed laser, together with a Precitec cladding head manipulated by a CLOOS 7 axes robot. The optimization of parameters and cladding quality has been assessed through Taguchi interaction matrix and graphical output. The study demonstrates that very good cladded layers with low dilution and increased mechanical proprieties could be fabricated using low laser energy density by involving a pulsed laser.

  2. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-07-01

    The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a 238 PuO 2 -powered pacemaker could be transformed into a terrorism weapon

  3. 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.

  4. Rigorous modeling of cladding modes in photonic crystal fibers

    DEFF Research Database (Denmark)

    Rindorf, Lars Henning; Bang, Ole

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

  5. Advanced powder metallurgy aluminum alloys via rapid solidification technology, phase 2

    Science.gov (United States)

    Ray, Ranjan; Jha, Sunil C.

    1987-01-01

    Marko's rapid solidification technology was applied to processing high strength aluminum alloys. Four classes of alloys, namely, Al-Li based (class 1), 2124 type (class 2), high temperature Al-Fe-Mo (class 3), and PM X7091 type (class 4) alloy, were produced as melt-spun ribbons. The ribbons were pulverized, cold compacted, hot-degassed, and consolidated through single or double stage extrusion. The mechanical properties of all four classes of alloys were measured at room and elevated temperatures and their microstructures were investigated optically and through electron microscopy. The microstructure of class 1 Al-Li-Mg alloy was predominantly unrecrystallized due to Zr addition. Yield strengths to the order of 50 Ksi were obtained, but tensile elongation in most cases remained below 2 percent. The class 2 alloys were modified composition of 2124 aluminum alloy, through addition of 0.6 weight percent Zr and 1 weight percent Ni. Nickel addition gave rise to a fine dispersion of intermetallic particles resisting coarsening during elevated temperature exposure. The class 2 alloy showed good combination of tensile strength and ductility and retained high strength after 1000 hour exposure at 177 C. The class 3 Al-Fe-Mo alloy showed high strength and good ductility both at room and high temperatures. The yield and tensile strength of class 4 alloy exceeded those of the commercial 7075 aluminum alloy.

  6. Development of advanced zirconium fuel cladding

    International Nuclear Information System (INIS)

    Jeong, Young Hwan; Park, S. Y.; Lee, M. H.

    2007-04-01

    This report includes the manufacturing technology developed for HANA TM claddings, a series of their characterization results as well as the results of their in-pile and out-of pile performances tests which were carried out to develop some fuel claddings for a high burn-up (70,000MWd/mtU) which are competitive in the world market. Some of the HANA TM claddings, which had been manufactured based on the results from the 1st and 2nd phases of the project, have been tested in a research reactor in Halden of Norway for an in-pile performance qualification. The results of the in-pile test showed that the performance of the HANA TM claddings for corrosion and creep was better than 50% compared to that of Zircaloy-4 or A cladding. It was also found that the out-of pile performance of the HANA TM claddings for such as LOCA and RIA in some accident conditions corrosion creep, tensile, burst and fatigue was superior or equivalent to that of the Zircaloy-4 or A cladding. The project also produced the other many data which were required to get a license for an in-pile test of HANA TM claddings in a commercial reactor. The data for the qualification or characterization were provided for KNFC to assist their activities to get the license for the in-pile test of HANA TM Lead Test Rods(LTR) in a commercial reactor

  7. 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.

  8. 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.

  9. Aluminum Analysis.

    Science.gov (United States)

    Sumrall, William J.

    1998-01-01

    Presents three problems based on the price of aluminum designed to encourage students to be cooperative and to use an investigative approach to learning. Students collect and synthesize information, analyze results, and draw conclusions. (AIM)

  10. Characterisation of a natural quartz crystal as a reference material for microanalytical determination of Ti, Al, Li, Fe, Mn, Ga and Ge

    Science.gov (United States)

    Audetat, Andreas; Garbe-Schonberg, Dieter; Kronz, Andreas; Pettke, Thomas; Rusk, Brian G.; Donovan, John J.; Lowers, Heather

    2015-01-01

    A natural smoky quartz crystal from Shandong province, China, was characterised by laser ablation ICP-MS, electron probe microanalysis (EPMA) and solution ICP-MS to determine the concentration of twenty-four trace and ultra trace elements. Our main focus was on Ti quantification because of the increased use of this element for titanium-in-quartz (TitaniQ) thermobarometry. Pieces of a uniform growth zone of 9 mm thickness within the quartz crystal were analysed in four different LA-ICP-MS laboratories, three EPMA laboratories and one solution-ICP-MS laboratory. The results reveal reproducible concentrations of Ti (57 ± 4 μg g-1), Al (154 ± 15 μg g-1), Li (30 ± 2 μg g-1), Fe (2.2 ± 0.3 μg g-1), Mn (0.34 ± 0.04 μg g-1), Ge (1.7 ± 0.2 μg g-1) and Ga (0.020 ± 0.002 μg g-1) and detectable, but less reproducible, concentrations of Be, B, Na, Cu, Zr, Sn and Pb. Concentrations of K, Ca, Sr, Mo, Ag, Sb, Ba and Au were below the limits of detection of all three techniques. The uncertainties on the average concentration determinations by multiple techniques and laboratories for Ti, Al, Li, Fe, Mn, Ga and Ge are low; hence, this quartz can serve as a reference material or a secondary reference material for microanalytical applications involving the quantification of trace elements in quartz.

  11. Defect features, texture and mechanical properties of friction stir welded lap joints of 2A97 Al-Li alloy thin sheets

    International Nuclear Information System (INIS)

    Chen, Haiyan; Fu, Li; Liang, Pei; Liu, Fenjun

    2017-01-01

    1.4 mm 2A97 Al-Li alloy thin sheets were welded by friction stir lap welding using the stirring tools with different pin length at different rotational speeds. The influence of pin length and rotational speed on the defect features and mechanical properties of lap joints were investigated in detail. Microstructure observation shows that the hook defect geometry and size mainly varies with the pin length instead of the rotational speed. The size of hook defects on both the advancing side (AS) and the retreating side (RS) increased with increasing the pin length, leading to the effective sheet thickness decreased accordingly. Electron backscatter diffraction analysis reveals that the weld zones, especially the nugget zone (NZ), have the much lower texture intensity than the base metal. Some new texture components are formed in the thermo-mechanical affected zone (TMAZ) and the NZ of joint. Lap shear test results show that the failure load of joints generally decreases with increasing the pin length and the rotational speed. The joints failed during the lap shear tests at three locations: the lap interface, the RS of the top sheet and the AS of the bottom sheet. The fracture locations are mainly determined by the hook defects. - Highlights: • Hook defect size mainly varies with the pin length of stirring tool. • The proportion of LAGBs and substructured grains increases from NZ to TMAZ. • Weld zones, especially the NZ, have the much lower texture intensity than the BM. • Lap shear failure load and fracture location of joints is relative to the hook defects.

  12. RIA simulation tests using driver tube for ATF cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  13. Detection of fatigue cracks in cladded blocks

    International Nuclear Information System (INIS)

    Singh, G.P.; Cervantes, R.A.; Manning, R.C.; Takama, S.

    1986-01-01

    A nuclear reactor pressure vessel (RPV) operates at high temperatures. Feedwater nozzles are susceptible to thermal fatigue; and, after a large number of plant startup/shutdown cycles, thermal fatigue cracking may be initiated at these nozzles. In order to address this problem, ultrasonic data were acquired from five cladded specimens with overall approximate 4-mm thick stainless steel cladding; the specimens contained one fatigue crack each. The study evaluates the application of signal processing and pattern recognition methods to discriminate between base metal-to-clad interface signals and fatigue crack signals. Details are presented

  14. 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.

  15. Chemical compatibility between cladding alloys and advanced fuels

    International Nuclear Information System (INIS)

    Fee, D.C.; Johnson, C.E.

    1975-05-01

    The National Advanced Fuels Program requires chemical, mechanical, and thermophysical properties data for cladding alloys. The compatibility behavior of cladding alloys with advanced fuels is critically reviewed. in carbide fuel pins, the principal compatibility problem is cladding carburization, diffusion of carbon into the cladding matrix accompanied by carbide precipitation. Carburization changes the mechanical properties of the cladding alloy. The extent of carburization increases in sodium (versus gas) bonded fuels. The depth of carburization increases with increasing sesquicarbide (M 2 C 3 ) content of the fuel. In nitride fuel pins, the principal compatibility problem is cladding nitriding, diffusion of nitrogen into the cladding matrix accompanied by nitride precipitation. Nitriding changes the mechanical properties of the cladding alloy. In both carbide and nitride fuel pins, fission products do not migrate appreciably to the cladding and do not appear to contribute to cladding attack. 77 references. (U.S.)

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

    NARCIS (Netherlands)

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

    2015-01-01

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

  17. CREEP STRAIN CORRELATION FOR IRRADIATED CLADDING

    International Nuclear Information System (INIS)

    P. Macheret

    2001-01-01

    In an attempt to predict the creep deformation of spent nuclear fuel cladding under the repository conditions, different correlations have been developed. One of them, which will be referred to as Murty's correlation in the following, and whose expression is given in Henningson (1998), was developed on the basis of experimental points related to unirradiated Zircaloy cladding (Henningson 1998, p. 56). The objective of this calculation is to adapt Murty's correlation to experimental points pertaining to irradiated Zircaloy cladding. The scope of the calculation is provided by the range of experimental parameters characterized by Zircaloy cladding temperature between 292 C and 420 C, hoop stress between 50 and 630 MPa, and test time extending to 8000 h. As for the burnup of the experimental samples, it ranges between 0.478 and 64 MWd/kgU (i.e., megawatt day per kilogram of uranium), but this is not a parameter of the adapted correlation

  18. MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION

    Directory of Open Access Journals (Sweden)

    Miroslav Cech

    2016-12-01

    Full Text Available This paper describes materials made of zirconium-based alloys used for nuclear fuel cladding fabrication. It is focused on corrosion problems their theoretical description and modeling in nuclear engineering.

  19. GSGG edge cladding development: Final technical report

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  20. Optimization of metal-clad waveguide sensors

    DEFF Research Database (Denmark)

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

    2005-01-01

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

  1. Plasmonic waveguides cladded by hyperbolic metamaterials.

    Science.gov (United States)

    Ishii, Satoshi; Shalaginov, Mikhail Y; Babicheva, Viktoriia E; Boltasseva, Alexandra; Kildishev, Alexander V

    2014-08-15

    Strongly anisotropic media with hyperbolic dispersion can be used for claddings of plasmonic waveguides (PWs). In order to analyze the fundamental properties of such waveguides, we analytically study 1D waveguides arranged from a hyperbolic metamaterial (HMM) in a HMM-Insulator-HMM (HIH) structure. We show that HMM claddings give flexibility in designing the properties of HIH waveguides. Our comparative study on 1D PWs reveals that HIH-type waveguides can have a higher performance than MIM or IMI waveguides.

  2. Flat-Cladding Fiber Bragg Grating Sensors for Large Strain Amplitude Fatigue Tests

    Directory of Open Access Journals (Sweden)

    Xijia Gu

    2010-08-01

    Full Text Available We have successfully developed a flat-cladding fiber Bragg grating sensor for large cyclic strain amplitude tests of up to ±8,000 με. The increased contact area between the flat-cladding fiber and substrate, together with the application of a new bonding process, has significantly increased the bonding strength. In the push-pull fatigue tests of an aluminum alloy, the plastic strain amplitudes measured by three optical fiber sensors differ only by 0.43% at a cyclic strain amplitude of ±7,000 με and 1.9% at a cyclic strain amplitude of ±8,000 με. We also applied the sensor on an extruded magnesium alloy for evaluating the peculiar asymmetric hysteresis loops. The results obtained were in good agreement with those measured from the extensometer, a further validation of the sensor.

  3. Clad Degradation- Summary and Abstraction for LA

    International Nuclear Information System (INIS)

    D. Stahl

    2004-01-01

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

  4. Anodic Aluminum Dissolution of LiTFSA Containing Electrolytes for Li-Ion-Batteries

    International Nuclear Information System (INIS)

    Hofmann, Andreas; Merklein, Lisa; Schulz, Michael; Hanemann, Thomas

    2014-01-01

    In this study, novel electrolyte solvents and the conducting salt lithium bis(trifluoromethanesulfonyl)azanide (LiTFSA) are compared with respect to aluminum anodic dissolution, often called in an ambiguous manner aluminum corrosion. Namely, mixtures of propylene carbonate, sulfolane, the ionic liquid N,N-diethyl-N-methyl-N-(2-methoxyethyl)ammonium bis(trifluoromethanesulfonyl)azanide (DMMA-TFSA) and various concentrations of LiTFSA are prepared and investigated in Al|Li cell configuration. It is verified that the choice of the concentration of LiTFSA and the composition of the solvent mixture affects the tendency of Al dissolution significantly. Principally, the best results (least dissolution/corrosion) are obtained in case of high LiTFSA concentrations and by adding the ionic liquid DMMA-TFSA to the organic solvents

  5. Adaptation of fuel code for light water reactor with austenitic steel rod cladding

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

  6. Adaptation of fuel code for light water reactor with austenitic steel rod cladding

    International Nuclear Information System (INIS)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira; Giovedi, Claudia

    2015-01-01

    Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

  7. Accident tolerant fuel cladding development: Promise, status, and challenges

    Science.gov (United States)

    Terrani, Kurt A.

    2018-04-01

    The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-01

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

  9. A pellet-clad interaction failure criterion

    International Nuclear Information System (INIS)

    Howl, D.A.; Coucill, D.N.; Marechal, A.J.C.

    1983-01-01

    A Pellet-Clad Interaction (PCI) failure criterion, enabling the number of fuel rod failures in a reactor core to be determined for a variety of normal and fault conditions, is required for safety analysis. The criterion currently being used for the safety analysis of the Pressurized Water Reactor planned for Sizewell in the UK is defined and justified in this paper. The criterion is based upon a threshold clad stress which diminishes with increasing fast neutron dose. This concept is consistent with the mechanism of clad failure being stress corrosion cracking (SCC); providing excess corrodant is always present, the dominant parameter determining the propagation of SCC defects is stress. In applying the criterion, the SLEUTH-SEER 77 fuel performance computer code is used to calculate the peak clad stress, allowing for concentrations due to pellet hourglassing and the effect of radial cracks in the fuel. The method has been validated by analysis of PCI failures in various in-reactor experiments, particularly in the well-characterised power ramp tests in the Steam Generating Heavy Water Reactor (SGHWR) at Winfrith. It is also in accord with out-of-reactor tests with iodine and irradiated Zircaloy clad, such as those carried out at Kjeller in Norway. (author)

  10. Transparent Aluminum Oxide Films by Edge Anodization

    Science.gov (United States)

    Stott, Jonathan; Greenwood, Thomas; Winn, David

    In this paper we present our recent work on manufacturing thin (3 - 5 μm) films of porous aluminum(III) oxide [PAO] using a novel edge-anodization technique. With this modified anodization process, we are able to create transparent PAO films on top of insulating substrates such as glass or plastic. By controlling the processing parameters, the index of refraction of PAO films can be engineered to match the substrate, which gives us a durable reflection-free and scratch-resistant coating over conventional optics or LCD displays. Eventually we hope to create ordered porous aluminum oxide cladding around an optical fiber core, which could have a number of interesting optical properties if the pore spacing can be matched to the wavelength of light in the fiber. This work was funded by Fairfield University startup funding.

  11. 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

  12. 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

  13. Inspection system for Zircaloy clad fuel rods

    International Nuclear Information System (INIS)

    Yancey, M.E.; Porter, E.H.; Hansen, H.R.

    1975-10-01

    A description is presented of the design, development, and performance of a remote scanning system for nondestructive examination of fuel rods. Characteristics that are examined include microcracking of fuel rod cladding, fuel-cladding interaction, cladding thickness, fuel rod diameter variation, and fuel rod bowing. Microcracking of both the inner and outer fuel rod surfaces and variations in wall thickness are detected by using a pulsed eddy current technique developed by Argonne National Laboratory (ANL). Fuel rod diameter variation and fuel rod bowing are detected by using two linear variable differential transformers (LVDTs) and a signal conditioning system. The system's mechanical features include variable scanning speeds, a precision indexing system, and a servomechanism to maintain proper probe alignment. Initial results indicate that the system is a very useful mechanism for characterizing irradiated fuel rods

  14. Laser cladding to select new glassy alloys

    International Nuclear Information System (INIS)

    Medrano, L.L.O.; Afonso, C.R.M.; Kiminami, C.S.; Gargarella, P.; Ramasco, B.

    2016-01-01

    A new experimental technique used to analyze the effect of compositional variation and cooling rate in the phase formation in a multicomponent system is the laser cladding. This work have evaluated the use of laser cladding to discover a new bulk metallic glass (BMG) in the Al-Co-Zr system. Coatings with composition variation have made by laser cladding using Al-Co-Zr alloys powders and the samples produced have been characterized by X ray diffraction, microscopy and energy-dispersive X-ray spectroscopy. The results did not show the composition variation as expected, because of incomplete melting during laser process. It was measured a composition variation tendency that allowed the glass forming investigation by the glass formation criterion λ+Δh 1/2 . The results have showed no glass formation in the coating samples, which prove a limited capacity of Zr-Co-Al system to form glass (author)

  15. Potential effects of gallium on cladding materials

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-10-01

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

  16. Stress corrosion testing of irradiated cladding tubes

    International Nuclear Information System (INIS)

    Lunde, L.; Olshausen, K.D.

    1980-01-01

    Samples from two fuel rods with different cladding have been stress corrosion tested by closed-end argon-iodine pressurization at 320 0 C. The fuel rods with stress relieved and recrystallized Zircaloy-2 had received burnups of 10.000 and 20.000 MWd/ton UO 2 , respectively. It was found that the SCC failure stress was unchanged or slightly higher for the irradiated than for the unirradiated control tubes. The tubes failed consistently in the end with the lowest irradiation dose. The diameter increase of the irradiated cladding during the test was 1.1% for the stress-relieved samples and 0.24% for the recrystallized samples. SEM examination revealed no major differences between irradiated and unirradiated cladding. A ''semi-ductile'' fracture zone in recrystallized material is described in some detail. (author)

  17. Potential effects of gallium on cladding materials

    International Nuclear Information System (INIS)

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

    1997-10-01

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

  18. Modelling of pellet-clad interaction during power ramps

    International Nuclear Information System (INIS)

    Zhou, G.; Lindback, J.E.; Schutte, H.C.; Jernkvist, L.O.; Massih, A.R.; Massih, A.R.

    2005-01-01

    A computational method to describe the pellet-clad interaction phenomenon is presented. The method accounts for the mechanical contact between fragmented pellets and the zircaloy clad, as well as for chemical reaction of fission products with zircaloy during power ramps. Possible pellet-clad contact states, soft, hard and friction, are taken into account in the computational algorithm. The clad is treated as an elastic-plastic-viscoplastic material with irradiation hardening. Iodine-induced stress corrosion cracking is described by using a fracture mechanics-based model for crack propagation. This integrated approach is used to evaluate two power ramp experiments made on boiling water reactor fuel rods in test reactors. The influence of the pellet-clad coefficient of friction on clad deformation is evaluated and discussed. Also, clad deformations, pellet-clad gap size and fission product gas release for one of the ramped rods are calculated and compared with measured data. (authors)

  19. Multilayer cladding with hyperbolic dispersion for plasmonic waveguides

    DEFF Research Database (Denmark)

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

    2015-01-01

    We study the properties of plasmonic waveguides with a dielectric core and multilayer metal-dielectric claddings that possess hyperbolic dispersion. The waveguides hyperbolic multilayer claddings show better performance in comparison to conventional plasmonic waveguides. © OSA 2015....

  20. Management of cladding hulls and fuel hardware

    International Nuclear Information System (INIS)

    1985-01-01

    The reprocessing of spent fuel from power reactors based on chop-leach technology produces a solid waste product of cladding hulls and other metallic residues. This report describes the current situation in the management of fuel cladding hulls and hardware. Information is presented on the material composition of such waste together with the heating effects due to neutron-induced activation products and fuel contamination. As no country has established a final disposal route and the corresponding repository, this report also discusses possible disposal routes and various disposal options under consideration at present

  1. Evaluation of fast experimental reactor claddings, (2)

    International Nuclear Information System (INIS)

    Miura, Makoto; Nagaki, Hiroshi; Koyama, Masahiro; Tanaka, Yasumasa

    1974-01-01

    Thin-walled fine tubes of Type 316 austenitic stainless steel are used for fuel cladding in Joyo (experimental FBR). The material exhibits the change of the mechanical properties in long-time annealing at high temperature, resulting from the precipitation of carbide in structure. In this connection, the experiment and the results on the changes of the microstructure and mechanical properties (proof stress and hardness) are described. The test specimens are the fuel cladding tubes produced for trial for Joyo core and those for FFTF core made in the U.S.A. They were heated between 400 0 and 850 0 C for 1000 hr in vacuum. (Mori, K.)

  2. Inpile (in PWR) testing of cladding materials

    International Nuclear Information System (INIS)

    Hahn, R.; Jeong, Y. H.; Baek, B. J.; Kim, K. H.; Kim, S. J.; Choi, B. K.; Kim, J. M.

    1999-04-01

    As an introduction, the reasons to perform inpile tests are depicted. An overview over general inpile test procedure is given, and test details which are necessary for the development of new alloys for high burnup claddings, like sample geometries and measuring techniques for inpile corrosion testing, are described in detail. Tests for the creep and length change behavior of cladding tubes are described briefly. Finally, conclusions are drawn and literature citations for further test details are given. (author). 9 refs., 2 tabs., 17 figs

  3. Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage

    International Nuclear Information System (INIS)

    Sindelar, R.L.

    1999-01-01

    The report provides an evaluation of: (1) the release rate of radionuclides through minor cladding penetrations (breaches) on aluminum-based spent nuclear fuel (AL SNF), and (2) the consequences of direct storage of breached AL SNF relative to the authorization basis for SRS basin operation

  4. Recycling of automotive aluminum

    OpenAIRE

    Cui, Jirang; Roven, Hans Jørgen

    2010-01-01

    With the global warming of concern, the secondary aluminum stream is becoming an even more important component of aluminum production and is attractive because of its economic and environmental benefits. In this work, recycling of automotive aluminum is reviewed to highlight environmental benefits of aluminum recycling, use of aluminum alloys in automotive applications, automotive recycling process, and new technologies in aluminum scrap process. Literature survey shows that newly developed t...

  5. Polarization effects in silicon-clad optical waveguides

    Science.gov (United States)

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

    1984-01-01

    By changing the thickness of a semiconductor cladding layer deposited on a planar dielectric waveguide, the TE or TM propagating modes may be selectively attenuated. This polarization effect is due to the periodic coupling between the lossless propagating modes of the dielectric slab waveguide and the lossy modes of the cladding layer. Experimental tests involving silicon claddings show high selectivity for either polarization.

  6. Electrometallurgical treatment of aluminum-matrix fuels

    International Nuclear Information System (INIS)

    Willit, J.L.; Gay, E.C.; Miller, W.E.; McPheeters, C.C.; Laidler, J.J.

    1996-01-01

    The electrometallurgical treatment process described in this paper builds on our experience in treating spent fuel from the Experimental Breeder Reactor (EBR-II). The work is also to some degree, a spin-off from applying electrometallurgical treatment to spent fuel from the Hanford single pass reactors (SPRs) and fuel and flush salt from the Molten Salt Reactor Experiment (MSRE) in treating EBR-II fuel, we recover the actinides from a uranium-zirconium fuel by electrorefining the uranium out of the chopped fuel. With SPR fuel, uranium is electrorefined out of the aluminum cladding. Both of these processes are conducted in a LiCl-KCl molten-salt electrolyte. In the case of the MSRE, which used a fluoride salt-based fuel, uranium in this salt is recovered through a series of electrochemical reductions. Recovering high-purity uranium from an aluminum-matrix fuel is more challenging than treating SPR or EBR-II fuel because the aluminum- matrix fuel is typically -90% (volume basis) aluminum

  7. Functionally graded materials produced by laser cladding

    NARCIS (Netherlands)

    Pei, Y.T.; Hosson, J.Th.M. De

    2000-01-01

    AlSi40 functionally graded materials (FGMs) were produced by a one-step laser cladding process on cast Al-alloy substrate as a possible solution for interfacial problems often present in laser coatings. The microstructure of the FGMs consists of a large amount of silicon primary particles surrounded

  8. Crud deposits on zircaloy-clad fuel

    International Nuclear Information System (INIS)

    Lister, D.H.

    1980-05-01

    The information on in-reactor corrosion of Zircaloy fuel cladding and crud deposition on fuel generated by Atomic Energy of Canada Limited up to 1979 has been reviewed elsewhere. This report is a summary of the crud deposition part of that review. (auth)

  9. The measurement of residual stresses in claddings

    International Nuclear Information System (INIS)

    Hofer, G.; Bender, N.

    1978-01-01

    The ring core method, a variation of the hole drilling method for the measurement of biaxial residual stresses, has been extended to measure stresses from depths of about 5 to 25mm. It is now possible to measure the stress profiles of clad material. Examples of measured stress profiles are shown and compared with those obtained with a sectioning technique. (author)

  10. Thermodynamics of pellet-cladding interaction

    International Nuclear Information System (INIS)

    Kyoh, Bunkei; Fuji, Kensho

    1987-01-01

    Equilibrium thermodynamic calculations are performed on the U-Zr-Cs-I-O system that is assumed to exist in the fuel-cladding gap of light water reactor (LWR) fuel under pellet-cladding interaction (PCI) failure condition. For this purpose a computer program called SOLGASMIX-PV for the calculation of complex multi-component equilibria is used, and the results of postirradiation examination are interpreted. The analysis of the thermodynamics of the system U-Zr-Cs-I-O indicates that cesium and iodine are assumed to be released from fuel pellet into the fuel-cladding gap as CsI, therefore, the Cs/I ratio in fuel-cladding bonding zone is one. The important condensed phases in this region are UO 2 , U 3 O 8 , Cs 2 U 2 O 7 , Cs 2 U 15 O 46 , ZrO 2 and CsI, and the major gaseous species are CsI, I 2 and I. Under this situation where Cs/I ratio is one, cesium-zirconate is not present. If, however, cesium rich phase is partially present then cesium will be associated with zirconium, possibly as Cs 2 ZrO 3 . (author)

  11. Composite cladding for nuclear fuel elements

    International Nuclear Information System (INIS)

    Armijo, J.S.

    1975-01-01

    The composite cladding described is for nuclear reactors and comprises a zirconium alloy substrate, a metallurgically bonded metal barrier on the inner surface of the substrate, and a metallurgically bonded internal coating on the inner surface of the metal barrier. The metal of the barrier is selected from aluminium, niobium, copper, nickel, stainless steel or iron. The internal coating is zirconium alloy [fr

  12. Study and Behaviour of Prefabricated Composite Cladding

    Science.gov (United States)

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

    2017-07-01

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

  13. Dissolution of aluminium-cladded fuel elements

    International Nuclear Information System (INIS)

    Bernhard, G.; Boessert, W.; Hladik, O.; Schwarzbach, R.

    1984-01-01

    In the molybdenum production plant at Rossendorf (AMOR) short-term irradiated aluminium-cladded fuel elements from the Rossendorf research reactor RFR are dissolved for the purpose of molybdenum 99 production. The dissolution behaviour of these fuel elements and the appropriate dissolver are described. (author)

  14. 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.

  15. Plasmonic waveguides cladded by hyperbolic metamaterials

    DEFF Research Database (Denmark)

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

    2014-01-01

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

  16. Prevention of nuclear fuel cladding materials corrosion

    International Nuclear Information System (INIS)

    Yang, K.R.; Yang, J.C.; Lee, I.C.; Kang, H.D.; Cho, S.W.; Whang, C.K.

    1983-01-01

    The only way which could be performed by the operator of nuclear power plant to minimizing the degradation of nuclear fuel cladding material is to control the water quality of primary coolant as specified standard conditions which dose not attack the cladding material. If the water quality of reactor coolant does not meet far from the specification, the failure will occure not only cladding material itself but construction material of primary system which contact with the coolant. The corrosion product of system material are circulate through the whole primary system with the coolant and activated by the neutron near the reactor core. The activated corrosion products and fission products which released from fuel rod to the coolant, so called crud, will repeate deposition and redeposition continuously on the fuel rod and construction material surface. As a result we should consider heat transfer problem. In this study following activities were performed; 1. The crud sample was taken from the spent fuel rod surface of Kori unit one and analized for radioactive element and non radioactive chemical species. 2. The failure mode of nuclear fuel cladding material was estimated by the investigation of releasing type of fission products from the fuel rod to the reactor coolant using the iodine isotopes concentration of reactor coolants. 3. A study was carried out on the sipping test results of spent fuel and a discussion was made on the water quality control records through the past three cycle operation period of Kori unit one plant. (Author)

  17. Thick tool steel coatings with laser cladding

    NARCIS (Netherlands)

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

    2007-01-01

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

  18. Cladding Effects on Structural Integrity of Nuclear Components

    International Nuclear Information System (INIS)

    Sattari-Far, Iradi; Andersson, Magnus

    2006-06-01

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

  19. Cladding Effects on Structural Integrity of Nuclear Components

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-06-15

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

  20. Influence of texture on fracture toughness of zircaloy cladding

    International Nuclear Information System (INIS)

    Grigoriev, V.; Andersson, Stefan

    1997-06-01

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

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

    Science.gov (United States)

    Zhang, Chuanwei; Yu, Zhengyang

    2017-12-01

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

  2. Membrane Purification Cell for Aluminum Recycling

    Energy Technology Data Exchange (ETDEWEB)

    David DeYoung; James Wiswall; Cong Wang

    2011-11-29

    Recycling mixed aluminum scrap usually requires adding primary aluminum to the scrap stream as a diluent to reduce the concentration of non-aluminum constituents used in aluminum alloys. Since primary aluminum production requires approximately 10 times more energy than melting scrap, the bulk of the energy and carbon dioxide emissions for recycling are associated with using primary aluminum as a diluent. Eliminating the need for using primary aluminum as a diluent would dramatically reduce energy requirements, decrease carbon dioxide emissions, and increase scrap utilization in recycling. Electrorefining can be used to extract pure aluminum from mixed scrap. Some example applications include producing primary grade aluminum from specific scrap streams such as consumer packaging and mixed alloy saw chips, and recycling multi-alloy products such as brazing sheet. Electrorefining can also be used to extract valuable alloying elements such as Li from Al-Li mixed scrap. This project was aimed at developing an electrorefining process for purifying aluminum to reduce energy consumption and emissions by 75% compared to conventional technology. An electrolytic molten aluminum purification process, utilizing a horizontal membrane cell anode, was designed, constructed, operated and validated. The electrorefining technology could also be used to produce ultra-high purity aluminum for advanced materials applications. The technical objectives for this project were to: - Validate the membrane cell concept with a lab-scale electrorefining cell; - Determine if previously identified voltage increase issue for chloride electrolytes holds for a fluoride-based electrolyte system; - Assess the probability that voltage change issues can be solved; and - Conduct a market and economic analysis to assess commercial feasibility. The process was tested using three different binary alloy compositions (Al-2.0 wt.% Cu, Al-4.7 wt.% Si, Al-0.6 wt.% Fe) and a brazing sheet scrap composition (Al-2

  3. The relative stress-corrosion-cracking susceptibility of candidate aluminum-lithium alloys for aerospace applications

    Science.gov (United States)

    Pizzo, P. P.

    1982-01-01

    Stress corrosion tests of Al-Li-Cu powder metallurgy alloys are described. Alloys investigated were Al-2.6% Li-1.4% and Al-2.6% Li-1.4% Cu-1.6% Mg. The base properties of the alloys were characterized. Process, heat treatment, and size/orientational effects on the tensile and fracture behavior were investigated. Metallurgical and electrochemical conditions are identified which provide reproducible and controlled parameters for stress corrosion evaluation. Preliminary stress corrosion test results are reported. Both Al-Li-Cu alloys appear more susceptible to stress corrosion crack initiation than 7075-T6 aluminum, with the magnesium bearing alloy being the most susceptible. Tests to determine the threshold stress intensity for the base and magnesium bearing alloys are underway. Twelve each, bolt loaded DCB type specimens are under test (120 days) and limited crack growth in these precracked specimens has been observed. General corrosion in the aqueous sodium chloride environment is thought to be obscuring results through crack tip blunting.

  4. Investigation of the Precipitation Behavior in Aluminum Based Alloys

    KAUST Repository

    Khushaim, Muna S.

    2015-11-30

    The transportation industries are constantly striving to achieve minimum weight to cut fuel consumption and improve overall performance. Different innovative design strategies have been placed and directed toward weight saving combined with good mechanical behavior. Among different materials, aluminum-based alloys play a key role in modern engineering and are widely used in construction components because of their light weight and superior mechanical properties. Introduction of different nano-structure features can improve the service and the physical properties of such alloys. For intelligent microstructure design in the complex Al-based alloy, it is important to gain a deep physical understanding of the correlation between the microstructure and macroscopic properties, and thus atom probe tomography with its exceptional capabilities of spatially resolution and quantitative chemical analyses is presented as a sophisticated analytical tool to elucidate the underlying process of precipitation phenomena in aluminum alloys. A complete study examining the influence of common industrial heat treatment on the precipitation kinetics and phase transformations of complex aluminum alloy is performed. The qualitative evaluation results of the precipitation kinetics and phase transformation as functions of the heat treatment conditions are translated to engineer a complex aluminum alloy. The study demonstrates the ability to construct a robust microstructure with an excellent hardness behavior by applying a low-energy-consumption, cost-effective method. The proposed strategy to engineer complex aluminum alloys is based on both mechanical strategy and intelligent microstructural design. An intelligent microstructural design requires an investigation of the different strengthen phases, such as T1 (Al2CuLi), θ′(Al2Cu), β′(Al3Zr) and δ′(Al3Li). Therefore, the early stage of phase decomposition is examined in different binary Al-Li and Al-Cu alloys together with different

  5. Microstructure of laser cladded martensitic stainless steel

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2006-08-01

    Full Text Available of austenitic solidification. Table 4 - Chemical composition of the laser cladded martensitic stainless steel in the dendritic and interdendritic areas Material Area C* Cr Ni Mn Si Mo Dendritic 0.3 12.8 0.15 0.7 0.65 0.02 Fe211-1 (420) Off... alloy steels and are shown in Table 4. Ms (ºC) = 550 – 350C – 40Mn - 20Cr – 10Mo – 17Ni – 8W – 35V – 10Cu + 15Co + 30Al (Eq 3) Table 5 - Ms temperatures of laser cladded martensitic stainless steel Material Ms Dendritic area (ºC) Ms...

  6. Development of metal-clad filled evacuated panel superinsulation

    Energy Technology Data Exchange (ETDEWEB)

    Wilkes, K.E.; Strizak, J.P.; Weaver, F.J. [Oak Ridge National Lab., TN (United States); Besser, J.E.; Smith, D.L. [Aladdin Industries, Inc. (United States)

    1997-03-01

    This Cooperative Research and Development Agreement (CRADA) was between Aladdin Industries, Inc. and Lockheed Martin Energy Research Corp. The purpose of the CRADA was to determine the thermal performance of various metal claddings used to encapsulate Filled Evacuated Panel (FEP) superinsulation and to optimize the cost versus thermal performance of the claddings. A FEP superinsulation is a new type of superinsulation with the potential for saving large amounts of energy in buildings, building equipment, transportation (refrigerated railcars and trucks), industrial applications, etc. The major disadvantage of metal claddings for FEPs is the heat loss through the cladding caused by the high thermal conductivity of most metals. In smaller FEPs, this heat loss can degrade the overall performance of the FEP by factors of two or more as compared with polymer-clad FEPs. On the other hand, metal claddings are essentially impermeable to ambient air, whereas polymer claddings are not. Thus, the longevity and reliability of metal-clad FEPs are much superior to polymer-clad FEPs. In addition, because of the very low vapor pressure of metals as compared to polymers, metal-clad FEPs can achieve and operate at lower internal pressures. These lower pressures allow use of less expensive and/or higher performance filler materials.

  7. Conditioning of nuclear cladding wastes by melting

    International Nuclear Information System (INIS)

    Puyou, M.; Jouan, A.; Jacquet-Francillon, N.

    1991-01-01

    This paper discusses a cold-crucible induction melting process to condition cladding waste from irradiated fast breeder reactor fuel. The process has been developed by the CEA at Marcoule (France) as part of a major R and D program. It has been qualified at industrial scale on nonradioactive waste, and at laboratory scale on radioactive waste: several radioactive ingots have been produced from actual stainless steel or zircaloy hulls. The results confirm the numerous advantages of this containment method

  8. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  9. Alloy development for high burnup cladding (PWR)

    International Nuclear Information System (INIS)

    Hahn, R.; Jeong, Y. H.; Baek, K. H.; Kim, S. J.; Choi, B. K.; Kim, J.M.

    1999-04-01

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

  10. Evolución de la fricción interna del material compuesto de matriz Al-Li 8090 reforzado con partículas de SiC

    Directory of Open Access Journals (Sweden)

    Gutiérrez-Urrutia, I.

    2001-04-01

    Full Text Available The present study has been undertaken to investigate the mechanism of thermal stress relief at the range of temperatures below room temperature for the metal matrix composite Al-Li 8090/SiC. For this aim the experimental technique of internal friction has been used which has been showed up very effective. Several thermal cycles from 453 K to 100 K were used in order to measure the internal friction as well as the elastic modulus of the material concluding that thermal stresses are relaxed by microplastic deformation around the reinforcements. It has been also related the variation in the elastic modulus with the different levels of precipitation.

    El presente trabajo investiga el mecanismo de relajación de tensiones térmicas a temperaturas por debajo de la de ambiente en el material compuesto Al-Li 8090/SiC. Para ello se ha empleado la técnica experimental de fricción interna que se ha mostrado la más eficaz para tal fin. Aplicando diferentes ciclos térmicos de 453 K a 100 K se midió tanto la fricción interna como el módulo elástico del material concluyendo que el mecanismo de relajación de tensiones térmicas es el de microdeformación plástica alrededor del reforzamiento. También se relaciona la variación del módulo elástico con los diferentes estadios de precipitación.

  11. 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)

  12. Hygrothermal performance of ventilated wooden cladding

    Energy Technology Data Exchange (ETDEWEB)

    Nore, Kristine

    2009-10-15

    This project contributes to more accurate design guidelines for high-performance building envelopes by analysis of hygrothermal performance of ventilated wooden cladding. Hygrothermal performance is defined by cladding temperature and moisture conditions, and subsequently by risk of degradation. Wood cladding is the most common facade material used in rural and residential areas in Norway. Historically, wooden cladding design varied in different regions in Norway. This was due to both climatic variations and the logistical distance to materials and craftspeople. The rebuilding of Norwegian houses in the 1950s followed central guidelines where local climate adaptation was often not evaluated. Nowadays we find some technical solutions that do not withstand all climate exposures. The demand for thermal comfort and also energy savings has changed hygrothermal condition of the building envelopes. In well-insulated wall assemblies, the cladding temperature is lower compared to traditional walls. Thus the drying out potential is smaller, and the risk of decay may be higher. The climate change scenario indicates a warmer and wetter future in Norway. Future buildings should be designed to endure harsher climate exposure than at present. Is there a need for refined climate differentiated design guidelines for building enclosures? As part of the Norwegian research programme 'Climate 2000', varieties of wooden claddings have been investigated on a test house in Trondheim. The aim of this investigation was to increase our understanding of the relation between microclimatic conditions and the responding hygrothermal performance of wooden cladding, according to orientation, design of ventilation gap, wood material quality and surface treatment. The two test facades, facing east and west have different climate exposure. Hourly measurements of in total 250 sensors provide meteorological data; temperature, radiation, wind properties, relative humidity, and test house data

  13. Pellet-clad interaction in water reactor fuels

    International Nuclear Information System (INIS)

    2004-01-01

    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.)

  14. Clad fiber capacitor and method of making same

    Science.gov (United States)

    Tuncer, Enis

    2012-12-11

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

  15. Analyses on Silicide Coating for LOCA Resistant Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings.

  16. Analyses on Silicide Coating for LOCA Resistant Cladding

    International Nuclear Information System (INIS)

    Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin

    2015-01-01

    A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings

  17. Determination of Stress-Corrosion Cracking in Aluminum-Lithium Alloy ML377

    Science.gov (United States)

    Valek, Bryan C.

    1995-01-01

    The use of aluminum-lithium alloys for aerospace applications is currently being studied at NASA Langley Research Center's Metallic Materials Branch. The alloys in question will operate under stress in a corrosive environment. These conditions are ideal for the phenomena of Stress-Corrosion Cracking (SCC) to occur. The test procedure for SCC calls for alternate immersion and breaking load tests. These tests were optimized for the lab equipment and materials available in the Light Alloy lab. Al-Li alloy ML377 specimens were then subjected to alternate immersion and breaking load tests to determine residual strength and resistance to SCC. Corrosion morphology and microstructure were examined under magnification. Data shows that ML377 is highly resistant to stress-corrosion cracking.

  18. Nuclear reactor fuel element with vanadium getter on cladding

    International Nuclear Information System (INIS)

    Johnson, C.E.; Carroll, K.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. 5 claims, 1 figure

  19. Computer analysis of elongation of the WWER fuel rod claddings

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2008-01-01

    In this paper description of mechanisms influencing changes of the WWER fuel cladding length and axial forces influencing fuel and cladding are presented. It is shown that shortening of the fuel claddings in case of high burnup can be explained by the change of the fuel and cladding reference state caused by reduction of the fuel rod power level - during reactor outages. It is noted that the presented calculated data are to be reviewed and interpreted as the preliminary results; further work is needed for their confirmation. (authors)

  20. Calculations of stresses in GCFR cladding under normal operating conditions

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Hsieh, T.C.; Billone, M.C.

    1979-11-01

    A modified version of the LIFE-III code, LIFE-GCFR, and classical stress-analysis techniques have been used to calculate the stresses in the GCFR cladding under normal reactor operating conditions. Several types of loadings on the cladding that occur during normal operation have been considered. These include fuel-cladding mechanical interaction, thermal stresses induced by radial and axial temperature gradients, and swelling gradient-induced stresses. The combined and individual effects of these loadings, as well as the effect of creep on cladding stresses, have been assessed

  1. Finite-width plasmonic waveguides with hyperbolic multilayer cladding.

    Science.gov (United States)

    Babicheva, Viktoriia E; Shalaginov, Mikhail Y; Ishii, Satoshi; Boltasseva, Alexandra; Kildishev, Alexander V

    2015-04-20

    Engineering plasmonic metamaterials with anisotropic optical dispersion enables us to tailor the properties of metamaterial-based waveguides. We investigate plasmonic waveguides with dielectric cores and multilayer metal-dielectric claddings with hyperbolic dispersion. Without using any homogenization, we calculate the resonant eigenmodes of the finite-width cladding layers, and find agreement with the resonant features in the dispersion of the cladded waveguides. We show that at the resonant widths, the propagating modes of the waveguides are coupled to the cladding eigenmodes and hence, are strongly absorbed. By avoiding the resonant widths in the design of the actual waveguides, the strong absorption can be eliminated.

  2. Fine Grain Aluminum Superplasticity

    Science.gov (United States)

    1980-02-01

    Continua on ravaraa sida H nacaaaary and identify by block numbar) Superplastic aluminum, Superplasticity, Superplastic forming. High strength aluminum...size. The presence of precipitate particles also acts to impede grain boundary migration during recrystallization, further aiding in maintaining a

  3. ALUMINUM BOX BUNDLING PRESS

    Directory of Open Access Journals (Sweden)

    Iosif DUMITRESCU

    2015-05-01

    Full Text Available In municipal solid waste, aluminum is the main nonferrous metal, approximately 80- 85% of the total nonferrous metals. The income per ton gained from aluminum recuperation is 20 times higher than from glass, steel boxes or paper recuperation. The object of this paper is the design of a 300 kN press for aluminum box bundling.

  4. All fiber cladding mode stripper with uniform heat distribution and high cladding light loss manufactured by CO2 laser ablation

    Science.gov (United States)

    Jebali, M. A.; Basso, E. T.

    2018-02-01

    Cladding mode strippers are primarily used at the end of a fiber laser cavity to remove high-power excess cladding light without inducing core loss and beam quality degradation. Conventional manufacturing methods of cladding mode strippers include acid etching, abrasive blasting or laser ablation. Manufacturing of cladding mode strippers using laser ablation consist of removing parts of the cladding by fused silica ablation with a controlled penetration and shape. We present and characterize an optimized cladding mode stripper design that increases the cladding light loss with a minimal device length and manufacturing time. This design reduces the localized heat generation by improving the heat distribution along the device. We demonstrate a cladding mode stripper written on a 400um fiber with cladding light loss of 20dB, with less than 0.02dB loss in the core and minimal heating of the fiber and coating. The manufacturing process of the designed component is fully automated and takes less than 3 minutes with a very high throughput yield.

  5. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    Energy Technology Data Exchange (ETDEWEB)

    Rudisill, T; John Mickalonis, J

    2006-09-27

    The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO{sub 2}) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO{sub 2} layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH{sub 4}F)/ammonium nitrate (NH{sub 4}NO{sub 3}) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO{sub 2} layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH{sub 4}){sub 2}ZrF{sub 6}) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination

  6. Preliminary experiments for the fabrication of clad for a spherical fuel for a research fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Almeida, L.A.A.

    1982-01-01

    A preliminary experiments using 1100 aluminum 0,5mm thick hemispheres welded on 10mm diameter steel and ceramic spheres in order to determine a method to clad spherical fuel for a research fluidized bed nuclear reactor were studied. The processes of hot press, T.I.G. and resistance we use for welding. A qualitative compression and metalographic tests of welded pieces are performed. By the analysis of the results of the tests we conclude that the resistance welding was the best. The experimental methods and the results with their analysis are presented in the paper. (Author) [pt

  7. Bragg grating induced cladding mode coupling due to asymmetrical index modulation in depressed cladding fibers

    DEFF Research Database (Denmark)

    Berendt, Martin Ole; Grüne-Nielsen, Lars; Soccolich, C.F.

    1998-01-01

    to reduce this problem. None of these designs seems to give complete solutions. In particular, the otherwise promising depressed cladding design gives a pronounced coupling to one LP01 mode, this has been referred to as a Ghost grating. To find the modes of the fiber we have established a numerical mode...

  8. Rheological evaluation of pretreated cladding removal waste

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  9. Photonic lantern with cladding-removable fibers

    Science.gov (United States)

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

    2014-07-01

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

  10. PCI resistant light water reactor fuel cladding

    International Nuclear Information System (INIS)

    Foster, J.P.; Sabol, G.P.

    1988-01-01

    A tubular nuclear fuel element cladding tube is described, the fuel element cladding tube forming the entire fuel element cladding and consisting of: a single continuous wall, the single continuous wall consisting of a single alloy selected from the group consisting of zirconium base alloys, A, B, C, D, and E; the single continuous wall characterized by a cold worked and stress relieved microstructure throughout; wherein the zirconium base alloy A contains 0.2 - 0.6 w/o Sn, 0.03 - 0.11 w/o sum of Fe and Cr, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy B contains 0.1 - 0.6 w/oo Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy C contains 1.2 - 1.7 w/o Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities; the zirconium base alloy D contains 0.15 - 0.6 w/o Sn, 0.15 - 0.5 w/o Fe, section 600 ppm O, and section 1500 ppm total impurities; and the zirconium base alloy E contains 0.4 - 0.6 w/o Sn, 0.1 - 0.3 w/o Fe, 0.03 - 0.07 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities

  11. Analysis of pellet cladding mechanical interaction using computational simulation

    International Nuclear Information System (INIS)

    Berretta, José R.; Suman, Ricardo B.; Faria, Danilo P.; Rodi, Paulo A.; Giovedi, Claudia

    2017-01-01

    During the operation of Pressurized Water Reactors (PWR), specifically under power transients, the fuel pellet experiences many phenomena, such as swelling and thermal expansion. These dimensional changes in the fuel pellet can enable occurrence of contact it and the cladding along the fuel rod. Thus, pellet cladding mechanical interaction (PCMI), due this contact, induces stress increase at the contact points during a period, until the accommodation of the cladding to the stress increases. This accommodation occurs by means of the cladding strain, which can produce failure, if the fuel rod deformation is permanent or the burst limit of the cladding is reached. Therefore, the mechanical behavior of the cladding during the occurrence of PCMI under power transients shall be investigated during the fuel rod design. Considering the Accident Tolerant Fuel program which aims to develop new materials to be used as cladding in PWR, one important design condition to be evaluated is the cladding behavior under PCMI. The purpose of this paper is to analyze the effects of the PCMI on a typical PWR fuel rod geometry with stainless steel cladding under normal power transients using computational simulation (ANSYS code). The PCMI was analyzed considering four geometric situations at the region of interaction between pellet and cladding. The first case, called “perfect fuel model” was used as reference for comparison. In the second case, it was considered the occurrence of a pellet crack with the loss of a chip. The goal for the next two cases was that a pellet chip was positioned into the gap of pellet-cladding, in the situations described in the first two cases. (author)

  12. Analysis of pellet cladding mechanical interaction using computational simulation

    Energy Technology Data Exchange (ETDEWEB)

    Berretta, José R.; Suman, Ricardo B.; Faria, Danilo P.; Rodi, Paulo A., E-mail: jose.berretta@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LabRisco/USP), São Paulo, SP (Brazil). Laboratório de Análise, Avaliação e Gerenciamento de Riscos

    2017-07-01

    During the operation of Pressurized Water Reactors (PWR), specifically under power transients, the fuel pellet experiences many phenomena, such as swelling and thermal expansion. These dimensional changes in the fuel pellet can enable occurrence of contact it and the cladding along the fuel rod. Thus, pellet cladding mechanical interaction (PCMI), due this contact, induces stress increase at the contact points during a period, until the accommodation of the cladding to the stress increases. This accommodation occurs by means of the cladding strain, which can produce failure, if the fuel rod deformation is permanent or the burst limit of the cladding is reached. Therefore, the mechanical behavior of the cladding during the occurrence of PCMI under power transients shall be investigated during the fuel rod design. Considering the Accident Tolerant Fuel program which aims to develop new materials to be used as cladding in PWR, one important design condition to be evaluated is the cladding behavior under PCMI. The purpose of this paper is to analyze the effects of the PCMI on a typical PWR fuel rod geometry with stainless steel cladding under normal power transients using computational simulation (ANSYS code). The PCMI was analyzed considering four geometric situations at the region of interaction between pellet and cladding. The first case, called “perfect fuel model” was used as reference for comparison. In the second case, it was considered the occurrence of a pellet crack with the loss of a chip. The goal for the next two cases was that a pellet chip was positioned into the gap of pellet-cladding, in the situations described in the first two cases. (author)

  13. COMPARISON OF CLADDING CREEP RUPTURE MODELS

    Energy Technology Data Exchange (ETDEWEB)

    P. Macheret

    2000-06-12

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

  14. Termination of plastic-clad fiber

    International Nuclear Information System (INIS)

    Nance, W.R.

    1982-03-01

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

  15. Stainless Steel Cladding Of Structural Steels By CO2 Laser Welding Techniques

    Science.gov (United States)

    Ludovico, A.; Daurelio, G.; Arcamone, O.

    1989-01-01

    Steel cladding processes are usually performed in different ways: hot roll cladding, strip cladding, weld cladding, explosion forming. For the first time, a medium power (2 KW c.w.) CO2 laser was used to clad structural steels (Fe 37C), 3 and 5 mm thick, with austenitic stainless steels (AISI 304 and AISI 316), 0.5 and 1.5 mm thick. The cladding technique we have developed uses the laser penetration welding process.

  16. Aluminum reference electrode

    Science.gov (United States)

    Sadoway, Donald R.

    1988-01-01

    A stable reference electrode for use in monitoring and controlling the process of electrolytic reduction of a metal. In the case of Hall cell reduction of aluminum, the reference electrode comprises a pool of molten aluminum and a solution of molten cryolite, Na.sub.3 AlF.sub.6, wherein the electrical connection to the molten aluminum does not contact the highly corrosive molten salt solution. This is accomplished by altering the density of either the aluminum (decreasing the density) or the electrolyte (increasing the density) so that the aluminum floats on top of the molten salt solution.

  17. Qualification of submerged-arc narrow strip cladding process

    International Nuclear Information System (INIS)

    Ayres, P.S.; Gottschling, J.D.; Jeffers, G.K.

    1975-08-01

    An unique narrow strip cladding process for use on both plate and forging material for nuclear components was developed. The qualification testing of this low-heat input process for cladding nuclear components, including those of SA508 Class 2 material is described. The theory that explains the acceptable results of these tests is also given. (auth)

  18. Qualification of submerged-arc narrow strip cladding process

    International Nuclear Information System (INIS)

    Ayres, P.S.; Gottschling, J.D.; Jeffers, G.K.

    1976-03-01

    Babcock and Wilcox has developed an unique narrow strip cladding process for use on both plate and forging material for nuclear components. The qualification testing of this low-heat input process for cladding nuclear components is described, including those of SA508 Class 2 material. The theory that explains the acceptable results of these tests is also given

  19. Laser cladding process development for high carbon steel substrates

    CSIR Research Space (South Africa)

    Lengopeng, T

    2014-11-01

    Full Text Available with an increase in laser power at constant scan speed. The hardness values taken on the HAZ showed a decrease with the increase in the number of clad layers during build-up. Meanwhile, four clad layer build up samples showed similar hardness trend independent...

  20. Interfacial adhesion of laser clad functionally graded materials

    NARCIS (Netherlands)

    Ocelik, V.; Pei, Y.T.; de Hosson, J.T.M.; Popoola, O; Dahotre, NB; Midea, SJ; Kopech, HM

    2003-01-01

    Two functionally graded coatings were prepared by different laser surface engineering techniques. Laser cladding of AlSi40 powder leads to the formation of functionally graded material (FGM) coating on AI-Si cast alloy substrate. Mapping of strain fields near the laser clad track using the digital

  1. Long-range plasmonic waveguides with hyperbolic cladding

    DEFF Research Database (Denmark)

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

    2015-01-01

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

  2. Material Selection for Accident Tolerant Fuel Cladding

    International Nuclear Information System (INIS)

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

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H 2 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 800°C could not be reproduced. Both FeCrAl and MAX phase Ti 2 AlC form a protective alumina scale in steam. However, commercial Ti 2 AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO 2 , and therefore Ti 2 AlC 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.

  3. 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.

  4. The ballooning of fuel cladding tubes: theory and experiment

    International Nuclear Information System (INIS)

    Shewfelt, R.S.W.

    1988-01-01

    Under some conditions, fuel clad ballooning can result in considerable strain before rupture. If ballooning were to occur during a loss-of-coolant accident (LOCA), the resulting substantial blockage of the sub-channel would restrict emergency core cooling. However, circumferential temperature gradients that would occur during a LOCA may significantly limit the average strain at failure. Understandably, the factors that control ballooning and rupture of fuel clad are required for the analysis of a LOCA. Considerable international effort has been spent on studying the deformation of Zircaloy fuel cladding under conditions that would occur during a LOCA. This effort has established a reasonable understanding of the factors that control the ballooning, failure time, and average failure strain of fuel cladding. In this paper, both the experimental and theoretical studies of the fuel clad ballooning are reviewed. (author)

  5. Test system to simulate transient overpower LMFBR cladding failure

    International Nuclear Information System (INIS)

    Barrus, H.G.; Feigenbutz, L.V.

    1981-01-01

    One of the HEDL programs has the objective to experimentally characterize fuel pin cladding failure due to cladding rupture or ripping. A new test system has been developed which simulates a transient mechanically-loaded fuel pin failure. In this new system the mechanical load is prototypic of a fuel pellet rapidly expanding against the cladding due to various causes such as fuel thermal expansion, fuel melting, and fuel swelling. This new test system is called the Fuel Cladding Mechanical Interaction Mandrel Loading Test (FCMI/MLT). The FCMI/MLT test system and the method used to rupture cladding specimens very rapidly to simulate a transient event are described. Also described is the automatic data acquisition and control system which is required to control the startup, operation and shutdown of the very fast tests, and needed to acquire and store large quantities of data in a short time

  6. Compact cladding-pumped planar waveguide amplifier and fabrication method

    Science.gov (United States)

    Bayramian, Andy J.; Beach, Raymond J.; Honea, Eric; Murray, James E.; Payne, Stephen A.

    2003-10-28

    A low-cost, high performance cladding-pumped planar waveguide amplifier and fabrication method, for deployment in metro and access networks. The waveguide amplifier has a compact monolithic slab architecture preferably formed by first sandwich bonding an erbium-doped core glass slab between two cladding glass slabs to form a multi-layer planar construction, and then slicing the construction into multiple unit constructions. Using lithographic techniques, a silver stripe is deposited and formed at a top or bottom surface of each unit construction and over a cross section of the bonds. By heating the unit construction in an oven and applying an electric field, the silver stripe is then ion diffused to increase the refractive indices of the core and cladding regions, with the diffusion region of the core forming a single mode waveguide, and the silver diffusion cladding region forming a second larger waveguide amenable to cladding pumping with broad area diodes.

  7. Effect of Rotation Rate on Microstructure and Properties of Friction Stir Welded Joints of Al/Cu Clad Plates

    Directory of Open Access Journals (Sweden)

    QIAO Ke

    2017-10-01

    Full Text Available Al/Cu clad plates were joined by friction stir welding (FSW, and the effect of rotation rate on microstructure and mechanical properties of joints was investigated. The results show that the laminar structure of aluminum and copper is generated in the weld. With increase the of rotation rate, the grain sizes of aluminum and copper are increased respectively. The average microhardness of the Al/Cu plates exceeds that of the as-received metal of 33.0 HV, and ultimate tensile strength is 127.21 MPa in the nugget zone when rotation rate is 1180 r/min. The microhardness of copper in the nugget zone is 99.7 HV, reached 82.05% of the microhardness of received metal, and void defect is main reason responsible for the decrease of mechanical properties of joints.

  8. 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.

  9. Production of aluminum-lithium near net shape extruded cylinders

    Science.gov (United States)

    Hartley, Paula J.

    1995-01-01

    In the late 1980's, under funding from the Advanced Launch System Program, numerous near net shape technologies were investigated as a means for producing high quality, low cost Aluminum-Lithium (Al-Li) hardware. Once such option was to extrude near net shape barrel panels instead of producing panels by machining thick plate into a final tee-stiffened configuration (which produced up to 90% scrap). This method offers a reduction in the volume of scrap and consequently reduces the buy-to-fly cost. Investigation into this technology continued under Shuttle-C funding where four Al alloys 2219, 2195, 2096, and RX 818 were extruded. Presented herein are the results of that program. Each alloy was successfully extruded at Wyman Gordon, opened and flattened at Ticorm, and solution heat treated and stretched at Reynolds Metals Company. The first two processes were quite successful while the stretching process did offer some challenges. Due to the configuration of the panels and the stretch press set-up, it was difficult to induce a consistent percentage of cold work throughout the length and width of each panel. The effects of this variation will be assessed in the test program to be conducted at a future date.

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

    Science.gov (United States)

    Garnier, John E.; Griffith, George W.

    2016-03-01

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

  11. Aluminum powder metallurgy processing

    Energy Technology Data Exchange (ETDEWEB)

    Flumerfelt, J.F.

    1999-02-12

    The objective of this dissertation is to explore the hypothesis that there is a strong linkage between gas atomization processing conditions, as-atomized aluminum powder characteristics, and the consolidation methodology required to make components from aluminum powder. The hypothesis was tested with pure aluminum powders produced by commercial air atomization, commercial inert gas atomization, and gas atomization reaction synthesis (GARS). A comparison of the GARS aluminum powders with the commercial aluminum powders showed the former to exhibit superior powder characteristics. The powders were compared in terms of size and shape, bulk chemistry, surface oxide chemistry and structure, and oxide film thickness. Minimum explosive concentration measurements assessed the dependence of explosibility hazard on surface area, oxide film thickness, and gas atomization processing conditions. The GARS aluminum powders were exposed to different relative humidity levels, demonstrating the effect of atmospheric conditions on post-atomization processing conditions. The GARS aluminum powders were exposed to different relative humidity levels, demonstrating the effect of atmospheric conditions on post-atomization oxidation of aluminum powder. An Al-Ti-Y GARS alloy exposed in ambient air at different temperatures revealed the effect of reactive alloy elements on post-atomization powder oxidation. The pure aluminum powders were consolidated by two different routes, a conventional consolidation process for fabricating aerospace components with aluminum powder and a proposed alternative. The consolidation procedures were compared by evaluating the consolidated microstructures and the corresponding mechanical properties. A low temperature solid state sintering experiment demonstrated that tap densified GARS aluminum powders can form sintering necks between contacting powder particles, unlike the total resistance to sintering of commercial air atomization aluminum powder.

  12. Materials considerations in accelerator targets

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Rebak, Raul B.

    2014-01-01

    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

  14. 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

  15. Elimination of aluminum adjuvants.

    Science.gov (United States)

    Hem, Stanley L

    2002-05-31

    In vitro dissolution experiments although perhaps not at typical body concentrations and temperatures demonstrated that the alpha-hydroxycarboxylic acids present in interstitial fluid (citric acid, lactic acid, and malic acid) are capable of dissolving aluminum-containing adjuvants. Amorphous aluminum phosphate adjuvant dissolved more rapidly than crystalline aluminum hydroxide adjuvant. Intramuscular administration in New Zealand White rabbits of aluminum phosphate and aluminum hydroxide adjuvants, which were labelled with 26Al, revealed that 26Al was present in the first blood sample (1 h) for both adjuvants. The area under the blood level curve for 28 days indicated that three times more aluminum was absorbed from aluminum phosphate adjuvant than aluminum hydroxide adjuvant. In vivo studies using 26Al-labelled adjuvants are relatively safe because accelerator mass spectrometry (AMS) can quantify quantities of 26Al as small as 10(-17) g. A similar study in humans would require a whole-body exposure of 0.7 microSv per year compared to the natural background exposure of 3000 microSv per year. The in vitro dissolution and in vivo absorption studies indicate that aluminum-containing adjuvants which are administered intramuscularly are dissolved by alpha-hydroxycarboxylic acids in interstitial fluid, absorbed into the blood, distributed to tissues, and eliminated in the urine.

  16. Corrosion surveillance program of aluminum spent fuel elements in wet storage sites

    International Nuclear Information System (INIS)

    Linardi, E; Haddad, R

    2012-01-01

    Due to different degradation issues observed in aluminum-clad spent fuel during long term storage in water, the IAEA implemented in 1996 a Coordinated Research Project (CRP) and a Regional Project for Latin America, on Corrosion of Research Reactor Aluminum Clad Spent Fuel in Water. Argentine has been among the participant countries of these projects, carrying out spent fuel corrosion surveillance activities in its storage facilities. As a result of the research a large database on corrosion of aluminum-clad fuel has been generated. It was determined that the main types of corrosion affecting the spent fuel are pitting and galvanic corrosion due to contact with stainless steel. It was concluded that the quality of the water is the critical factor to control in a spent fuel storage facility. Another phase of the program is being conducted currently, which began in 2011 with the immersion of test racks in the RA1 reactor pool, and in the Research Reactor Spent Fuel Storage Facility (FACIRI), located in Ezeiza Atomic Center. This paper presents the results of the chemical analysis of the water performed so far, and its relationship with the examination of the coupons extracted from the sites (author)

  17. Hollow Core Photonic Crystal Fibre Comprising a Fibre Grating in the Cladding and its Applications

    DEFF Research Database (Denmark)

    2010-01-01

    An optical fibre is provided having a fibre cladding around a longitudinally extending optical propagation core. The cladding has a reflection region of a varying refractive index in the longitudinal direction.......An optical fibre is provided having a fibre cladding around a longitudinally extending optical propagation core. The cladding has a reflection region of a varying refractive index in the longitudinal direction....

  18. Cladding development and characterization in the United States

    International Nuclear Information System (INIS)

    Lobsinger, R.J.

    1977-01-01

    During the past decade, an extensive program of cladding development has been carried out at HEDL in support of the FFTF/Breeder Reactor Program. This activity has resulted in the establishment of specification, fabrication, and characterization requirements for breeder reactor fuel cladding. In establishing these requirements, input has been derived from vendor development programs, evaluation studies, and reviews with other national laboratories. The major fabrication requirements of the current specification as derived from the development program and the testing required to assure product conformance and characterize cladding behavior are described

  19. Fabrication of oxide dispersion strengthened ferritic clad fuel pins

    International Nuclear Information System (INIS)

    Zirker, L.R.; Bottcher, J.H.; Shikakura, S.; Tsai, C.L.

    1991-01-01

    A resistance butt welding procedure was developed and qualified for joining ferritic fuel pin cladding to end caps. The cladding are INCO MA957 and PNC ODS lots 63DSA and 1DK1, ferritic stainless steels strengthened by oxide dispersion, while the end caps are HT9 a martensitic stainless steel. With adequate parameter control the weld is formed without a residual melt phase and its strength approaches that of the cladding. This welding process required a new design for fuel pin end cap and weld joint. Summaries of the development, characterization, and fabrication processes are given for these fuel pins. 13 refs., 6 figs., 1 tab

  20. Deep-probe metal-clad waveguide biosensors

    DEFF Research Database (Denmark)

    Skivesen, Nina; Horvath, Robert; Thinggaard, S.

    2007-01-01

    -clad waveguide sensor is shown to be the best all-round alternative to the surface-plasmon resonance biosensor. Both metal-clad waveguides are tested experimentally for cell detection, showing a detection linut of 8-9 cells/mm(2). (c) 2006 Elsevier B.V. All rights reserved.......Two types of metal-clad waveguide biosensors, so-called dip-type and peak-type, are analyzed and tested. Their performances are benchmarked against the well-known surface-plasmon resonance biosensor, showing improved probe characteristics for adlayer thicknesses above 150-200 nm. The dip-type metal...

  1. Long-range plasmonic waveguides with hyperbolic cladding.

    Science.gov (United States)

    Babicheva, Viktoriia E; Shalaginov, Mikhail Y; Ishii, Satoshi; Boltasseva, Alexandra; Kildishev, Alexander V

    2015-11-30

    We study plasmonic waveguides with dielectric cores and hyperbolic multilayer claddings. The proposed design provides better performance in terms of propagation length and mode confinement in comparison to conventional designs, such as metal-insulator-metal and insulator-metal-insulator plasmonic waveguides. We show that the proposed structures support long-range surface plasmon modes, which exist when the permittivity of the core matches the transverse effective permittivity component of the metamaterial cladding. In this regime, the surface plasmon polaritons of each cladding layer are strongly coupled, and the propagation length can be on the order of a millimeter.

  2. Is the Aluminum Hypothesis Dead?

    Science.gov (United States)

    2014-01-01

    The Aluminum Hypothesis, the idea that aluminum exposure is involved in the etiology of Alzheimer disease, dates back to a 1965 demonstration that aluminum causes neurofibrillary tangles in the brains of rabbits. Initially the focus of intensive research, the Aluminum Hypothesis has gradually been abandoned by most researchers. Yet, despite this current indifference, the Aluminum Hypothesis continues to attract the attention of a small group of scientists and aluminum continues to be viewed with concern by some of the public. This review article discusses reasons that mainstream science has largely abandoned the Aluminum Hypothesis and explores a possible reason for some in the general public continuing to view aluminum with mistrust. PMID:24806729

  3. Characteristics Of KRS-5 Fiber With Crystalline Cladding

    Science.gov (United States)

    Kimura, M.; Kachi, S.; Shiroyama, K.

    1986-05-01

    KRS-5 polycrystalline fibers having a core-cladding structure have been fabricated by complex extrusion. The core material is KRS-5 (T1Br-TlI), and the cladding material is KRS-6 (T1Br-T1C1). Through thermal treatment after fiber fabrication, the refractive index profile is changed to the graded index (GI) type. The attenuation loss of these core-cladding GI fibers is 0.2 dB/m at about 10.6 μm (lauching NA = 0.05). Fibers with this core-cladding structure will enable wide applications using far infrared wavelength light, for example in CO2, laser processing machines and spectroscopy equipment, because of their ease in handling.

  4. Cladding axial elongation models for FRAP-T6

    International Nuclear Information System (INIS)

    Shah, V.N.; Carlson, E.R.; Berna, G.A.

    1983-01-01

    This paper presents a description of the cladding axial elongation models developed at the Idaho National Engineering Laboratory (INEL) for use by the FRAP-T6 computer code in analyzing the response of fuel rods during reactor transients in light water reactors (LWR). The FRAP-T6 code contains models (FRACAS-II subcode) that analyze the structural response of a fuel rod including pellet-cladding-mechanical-interaction (PCMI). Recently, four models were incorporated into FRACAS-II to calculate cladding axial deformation: (a) axial PCMI, (b) trapped fuel stack, (c) fuel relocation, and (d) effective fuel thermal expansion. Comparisons of cladding axial elongation measurements from two experiments with the corresponding FRAP-T6 calculations are presented

  5. High Temperature Resistance Claddings for Nuclear Thermal Rockets, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This program will develop a series of nano-/micro-composite coated nuclear reactor facing components using MesoCoat's CermaCladTM process. This proposed SBIR program...

  6. Oxidation during reflood of reactor core with melting cladding

    Energy Technology Data Exchange (ETDEWEB)

    Siefken, L.J.; Allison, C.M.; Davis, K.L. [and others

    1995-09-01

    Models were recently developed and incorporated into the SCDAP/RELAP5 code for calculating the oxidation of fuel rods during cladding meltdown and reflood. Experiments have shown that a period of intense oxidation may occur when a hot partially oxidized reactor core is reflooded. This paper offers an explanation of the cladding meltdown and oxidation processes that cause this intense period of oxidation. Models for the cladding meltdown and oxidation processes are developed. The models are assessed by simulating a severe fuel damage experiment that involved reflood. The models for cladding meltdown and oxidation were found to improve calculation of the temperature and oxidation of fuel rods during the period in which hot fuel rods are reflooded.

  7. A Multi-Scale Modeling of Laser Cladding Process (Preprint)

    National Research Council Canada - National Science Library

    Cao, J; Choi, J

    2006-01-01

    Laser cladding is an additive manufacturing process that a laser generates a melt-pool on the substrate material while a second material, as a powder or a wire form, is injected into that melt-pool...

  8. Finite-width plasmonic waveguides with hyperbolic multilayer cladding

    DEFF Research Database (Denmark)

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

    2015-01-01

    Engineering plasmonic metamaterials with anisotropic optical dispersion enables us to tailor the properties of metamaterial-based waveguides. We investigate plasmonic waveguides with dielectric cores and multilayer metal-dielectric claddings with hyperbolic dispersion. Without using any homogeniz...

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

    Science.gov (United States)

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

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

  10. Effect of Localized Corrosion on Fatigue-Crack Growth in 2524-T3 and 2198-T851 Aluminum Alloys Used as Aircraft Materials

    Science.gov (United States)

    Moreto, J. A.; Broday, E. E.; Rossino, L. S.; Fernandes, J. C. S.; Bose Filho, W. W.

    2018-03-01

    Corrosion and fatigue of aluminum alloys are major issues for the in-service life assessment of aircraft structures and for the management of aging air fleets. The aim of this work was to evaluate the effect of localized corrosion on fatigue crack growth (FCG) resistance of the AA2198-T851 Al-Li alloy (Solution Heat Treated, Cold Worked, and Artificially Aged), comparing it with the FCG resistance of AA2524-T3 (Solution Heat Treated and Cold Worked), considering the effect of seawater fog environment. Before fatigue tests, the corrosion behavior of 2198-T851 and 2524-T3 aluminum alloys was verified using open circuit potential and potentiodynamic polarization techniques. Fatigue in air and corrosion fatigue tests were performed applying a stress ratio (R) of 0.1, 15 Hz (air) and 0.1 Hz (seawater fog) frequencies, using a sinusoidal waveform in all cases. The results showed that the localized characteristics of the 2198-T851 and 2524-T3 aluminum alloys are essentially related to the existence of intermetallic compounds, which, due to their different nature, may be cathodic or anodic in relation to the aluminum matrix. The corrosive medium has affected the FCG rate of both aluminum alloys, in a quite similar way.

  11. 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.

  12. Interim report on the creepdown of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Hobson, D.O.; Dodd, C.V.

    1977-01-01

    This report describes the creepdown phenomenon in Zircaloy fuel cladding and the methods by which it will be measured and analyzed. Instrumentation for monitoring radial deformation in the cladding is described in detail--in terms of theory, design, and stability. The programs that control the microcomputer are listed, both to document the level of sophistication of the instrumentation and to indicate the flexibility of the test equipment

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

    International Nuclear Information System (INIS)

    McCoy, K.

    2000-01-01

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

  14. Special techniques for tensile tests of irradiated zirconium claddings

    International Nuclear Information System (INIS)

    Prokhorov, V.I.; Makarov, O.Ju.; Smirnov, V.P.

    2002-01-01

    Irradiated zirconium alloy claddings possessing property anisotropy should be tested in transverse and longitudinal directions. Such mechanical tests can be performed in conditions of large variety of geometric peculiarities of specimens, supports or grips. The objective of the work is the development of the unified complex of updated special techniques that allow investigation of mechanical pre- and post-irradiation properties of VVER claddings including radiation effect of property anisotropy changes in the same way. (author)

  15. Laser cladding crack repair of austenitic stainless steel

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2009-06-01

    Full Text Available of pressurised vessel Fig. 2: Overlay positional welding Fig. 3: Crack sealing without peening prior to cladding 3 Hammer peening was introduced in order to mechanically seal of squirting water prior to crack sealing. Leaks were reduced from.... Simulated crack repair and overlay cladding with 316L is shown in Fig. 6. Three small leaking pores were observed in the first layer, indicated by arrows. Hammer peening was applied to the first layer to mechanically seal the leaking pores prior...

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

    International Nuclear Information System (INIS)

    Kim, Kyu-Tae

    2013-01-01

    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 −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

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

    Science.gov (United States)

    Sexton, Leo

    2003-03-01

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

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

    Science.gov (United States)

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

    2014-03-01

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

  19. A model for hydrogen pickup for BWR cladding materials

    International Nuclear Information System (INIS)

    Hede, G.; Kaiser, U.

    2001-01-01

    It has been observed that rod elongation is driven by the hydrogen pickup but not by corrosion as such. Based on this a non-destructive method to determine clad hydrogen concentration has been developed. The method is based on the observation that there are three different mechanisms behind the rod growth: the effect of neutron irradiation on the Zircaloy microstructure, the volume increase of the cladding as an effect of hydride precipitation and axial pellet-cladding-mechanical-interaction (PCMI). The derived correlation is based on the experience of older cladding materials, inspected at hot-cell laboratories, that obtained high hydrogen levels (above 500 ppm) at lower burnup (assembly burnup below 50 MWd/kgU). Now this experience can be applied, by interpolation, on more modern cladding materials with a burnup beyond 50 MWd/kgU by analysis of the rod growth database of the respective cladding materials. Hence, the method enables an interpolation rather than an extrapolation of present day hydrogen pickup database, which improves the reliability and accuracy. Further, one can get a good estimate of the hydrogen pickup during an ongoing outage based on a non-destructive method. Finally, rod growth measurements are normally performed for a large population of rods, hence giving a good statistics compared to examination of a few rods at a hot cell. (author)

  20. Development of aluminum (Al5083)-clad ternary Ag-In-Cd alloy for JSNS decoupled moderator

    International Nuclear Information System (INIS)

    Teshigawara, M.; Harada, M.; Saito, S.; Oikawa, K.; Maekawa, F.; Futakawa, M.; Kikuchi, K.; Kato, T.; Ikeda, Y.; Naoe, T.; Koyama, T.; Ooi, T.; Zherebtsov, S.; Kawai, M.; Kurishita, H.; Konashi, K.

    2006-01-01

    To develop Ag (silver)-In (indium)-Cd (cadmium) alloy decoupler, a method is needed to bond the decoupler between Al alloy (Al5083) and the ternary Ag-In-Cd alloy. We found that a better HIP condition was temperature, pressure and holding time at 803 K, 100 MPa and 10 min. for small test pieces (φ22 mm in dia. x 6 mm in height). Hardened layer due to the formation of AlAg 2 was found in the bonding layer, however, the rupture strength of the bonding layer is more than 30 MPa, the calculated design stress. Bonding tests of a large size piece (200 x 200 x 30 mm 3 ), which simulated the real scale, were also performed according to the results of small size tests. The result also gave good bonding and enough required-mechanical-strength

  1. Cladding Attachment Over Thick Exterior Insulating Sheathing

    Energy Technology Data Exchange (ETDEWEB)

    Baker, P. [Building Science Corporation, Somerville, MA (United States); Eng, P. [Building Science Corporation, Somerville, MA (United States); Lepage, R. [Building Science Corporation, Somerville, MA (United States)

    2014-01-01

    The addition of insulation to the exterior of buildings is an effective means of increasing the thermal resistance of both wood framed walls as well as mass masonry wall assemblies. For thick layers of exterior insulation (levels greater than 1.5 inches), the use of wood furring strips attached through the insulation back to the structure has been used by many contractors and designers as a means to provide a convenient cladding attachment location (Straube and Smegal 2009, Pettit 2009, Joyce 2009, Ueno 2010). The research presented in this report is intended to help develop a better understanding of the system mechanics involved and the potential for environmental exposure induced movement between the furring strip and the framing. BSC sought to address the following research questions: 1.What are the relative roles of the mechanisms and the magnitudes of the force that influence the vertical displacement resistance of the system? 2.Can the capacity at a specified deflection be reliably calculated using mechanics based equations? 3.What are the impacts of environmental exposure on the vertical displacement of furring strips attached directly through insulation back to a wood structure?

  2. Cladding Attachment Over Thick Exterior Insulating Sheathing

    Energy Technology Data Exchange (ETDEWEB)

    Baker, P. [Building Science Corporation, Somerville, MA (United States); Eng, P. [Building Science Corporation, Somerville, MA (United States); Lepage, R. [Building Science Corporation, Somerville, MA (United States)

    2014-01-01

    The addition of insulation to the exterior of buildings is an effective means of increasing the thermal resistance of both wood framed walls as well as mass masonry wall assemblies. For thick layers of exterior insulation (levels greater than 1.5 inches), the use of wood furring strips attached through the insulation back to the structure has been used by many contractors and designers as a means to provide a convenient cladding attachment location (Straube and Smegal 2009, Pettit 2009, Joyce 2009, Ueno 2010). The research presented in this report is intended to help develop a better understanding of the system mechanics involved and the potential for environmental exposure induced movement between the furring strip and the framing. BSC sought to address the following research questions: 1. What are the relative roles of the mechanisms and the magnitudes of the force that influence the vertical displacement resistance of the system? 2. Can the capacity at a specified deflection be reliably calculated using mechanics based equations? 3. What are the impacts of environmental exposure on the vertical displacement of furring strips attached directly through insulation back to a wood structure?

  3. Advances in aluminum pretreatment

    Energy Technology Data Exchange (ETDEWEB)

    Sudour, Michel; Maintier, Philippe [PPG Industries France, 3 Z.A.E. Les Dix Muids, B.P. 89, F-59583 Marly (France); Simpson, Mark [PPG Industries Inc., 1200 Piedmont Troy, Michigan 48083 (United States); Quaglia, Paolo [PPG Industries Italia, Via Garavelli 21, I-15028 Quattordio (Italy)

    2004-07-01

    As automotive manufacturers continue to look for ways to reduce vehicle weight, aluminum is finding more utility as a body panel component. The substitution of cold-rolled steel and zinc-coated substrates with aluminum has led to new challenges in vehicle pretreatment. As a result, changes to traditional pretreatment chemistries and operating practices are necessary in order to produce an acceptable coating on aluminum body panels. These changes result in increased sludging and other undesirable characteristics. In addition to the chemistry changes, there are also process-related problems to consider. Many existing automotive pretreatment lines simply were not designed to handle aluminum and its increased demands on filtration and circulation equipment. To retrofit such a system is capital intensive and in addition to requiring a significant amount of downtime, may not be totally effective. Thus, the complexities of pre-treating aluminum body panels have actually had a negative effect on efforts to introduce more aluminum into new vehicle design programs. Recent research into ways of reducing the negative effects has led to a new understanding of the nature of zinc phosphate bath -aluminum interactions. Many of the issues associated with the pretreatment of aluminum have been identified and can be mitigated with only minor changes to the zinc phosphate bath chemistry. The use of low levels of soluble Fe ions, together with free fluoride, has been shown to dramatically improve the efficiency of a zinc phosphate system processing aluminum. Appearance of zinc phosphate coatings, coating weights and sludge are all benefited by this chemistry change. (authors)

  4. Effects of cold worked and fully annealed claddings on fuel failure behaviour

    International Nuclear Information System (INIS)

    Saito, Shinzo; Hoshino, Hiroaki; Shiozawa, Shusaku; Yanagihara, Satoshi

    1979-12-01

    Described are the results of six differently heat-treated Zircaloy clad fuel rod tests in NSRR experiments. The purpose of the test is to examine the extent of simulating irradiated claddings in mechanical properties by as-cold worked ones and also the effect of fully annealing on the fuel failure bahaviour in a reactivity initiated accident (RIA) condition. As-cold worked cladding does not properly simulated the embrittlement of the irradiated one in a RIA condition, because the cladding is fully annealed before the fuel failure even in the short transient. Therefore, the fuel behaviour such as fuel failure threshold energy, failure mechanism, cladding deformation and cladding oxidation of the fully annealed cladding fuel, as well as that of the as-cold worked cladding fuel, are not much different from that of the standard stress-relieved cladding fuel. (author)

  5. Estimation of aluminum and argon activation sources in the HANARO coolant

    International Nuclear Information System (INIS)

    Jun, Byung Jin; Lee, Byung Chul; Kim, Myong Seop

    2010-01-01

    The activation products of aluminum and argon are key radionuclides for operational and environmental radiological safety during the normal operation of open-tank-in-pool type research reactors using aluminum-clad fuels. Their activities measured in the primary coolant and pool surface water of HANARO have been consistent. We estimated their sources from the measured activities and then compared these values with their production rates obtained by a core calculation. For each aluminum activation product, an equivalent aluminum thickness (EAT) in which its production rate is identical to its release rate into the coolant is determined. For the argon activation calculation, the saturated argon concentration in the water at the temperature of the pool surface is assumed. The EATs are 5680, 266 and 1.2 nm, respectively, for Na-24, Mg-27 and Al-28, which are much larger than the flight lengths of the respective recoil nuclides. These values coincide with the water solubility levels and with the half-lives. The EAT for Na-24 is similar to the average oxide layer thickness (OLT) of fuel cladding as well; hence, the majority of them in the oxide layer may be released to the coolant. However, while the average OLT clearly increases with the fuel burn-up during an operation cycle, its effect on the pool-top radiation is not distinguishable. The source of Ar-41 is in good agreement with the calculated reaction rate of Ar-40 dissolved in the coolant

  6. Aluminum 2195 T8 Gore Development for Space Launch System Core and Upper Stage

    Science.gov (United States)

    Volz, Martin

    2015-01-01

    Gores are pie-shaped panels that are welded together to form the dome ends of rocket fuel tanks as shown in figure 1. Replacing aluminum alloy 2219 with aluminum (Al)-lithium (Li) alloy 2195 as the Space Launch System (SLS) cryogenic tank material would save enormous amounts of weight. In fact, it has been calculated that simply replacing Al 2219 gores with Al 2195 gores on the SLS core stage domes could save approximately 3,800 pound-mass. This is because the Al-Li 2195 alloy exhibits both higher mechanical properties and lower density than the SLS baseline Al 2219 alloy. Indeed, the known advantages of Al 2195 led to its use as a replacement for Al 2219 in the shuttle external tank program. The required thicknesses of Al 2195 gores for either SLS core stage tanks or upper stage tanks will depend on the specific design configurations. The required thicknesses or widths may exceed the current experience base in the manufacture of such gores by the stretch-forming process. Accordingly, the primary objective of this project was to enhance the formability of Al 2195 by optimizing the heat treatment and stretch-forming process for gore thicknesses up to 0.75 inches, which envelop the maximum expected gore thicknesses for SLS tank configurations.

  7. Corrosion Resistance Evaluation of HANA Claddings in Commercial PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hee-Hun; Kwon, Oh-Hyun; Kim, Hong-Jin; Yoo, Jong-Sung; Kim, Yong-Hwan [KEPCO NF, Daejeon (Korea, Republic of)

    2014-10-15

    Korea Atomic Energy Research Institute (KAERI) in collaboration with KEPCO Nuclear Fuel (KNF) developed newly-advanced alloy which are named HANA (High-performance Alloy for Nuclear Application) for high burnup PWR nuclear fuel, showed an excellent out-pile corrosion resistance in PWR simulating loop conditions. And in-pile corrosion resistance of HANA claddings, which was examined at the first provisional inspection after -185 FPD of irradiation in the Halden Reactor, and also shown superior to the other references alloy. Also, other researches showed a much better corrosion resistance when compared to the other Zr-based alloy in various corrosion conditions. In this study, the LTA program for newly-developed fuel assembly (HIPER) with the HANA claddings was implemented to justify the performance for 3 cycles of operation schedule in Hanul nuclear power plant. The objective of this study is to compare corrosion properties of reference alloy with HANA claddings loaded in Hanul nuclear power plant.. For the examination procedures, the oxide thickness measurements method and equipment of PSE are described in detail as follow in measurement methods chapter. Finally, based on the above mentioned measurements method, the summarized oxide thickness data obtained from PSE are evaluated for the corrosion resistance in commercial nuclear power plant and some discussion for the corrosion resistance are described. In the past, corrosion resistance of HANA claddings was successfully conducted in test reactor. In this study, the corrosion characteristic of HANA claddings which are applied to HIPER is examined in the commercial nuclear power plant. HANA claddings in the HIPER showed a more improved corrosion resistance than reference alloy claddings and are evaluated well with meeting the oxide thickness criteria.

  8. Corrosion Inhibitors for Aluminum.

    Science.gov (United States)

    Muller, Bodo

    1995-01-01

    Describes a simple and reliable test method used to investigate the corrosion-inhibiting effects of various chelating agents on aluminum pigments in aqueous alkaline media. The experiments that are presented require no complicated or expensive electronic equipment. (DDR)

  9. Development of advanced LWR fuel cladding

    International Nuclear Information System (INIS)

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

    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

  10. 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.

  11. Corrosion Protection of Aluminum

    Science.gov (United States)

    Dalrymple, R. S.; Nelson, W. B.

    1963-07-01

    Treatment of aluminum-base metal surfaces in an autoclave with an aqueous chromic acid solution of 0.5 to 3% by weight and of pH below 2 for 20 to 50 hrs at 160 to 180 deg C produces an extremely corrosion-resistant aluminum oxidechromium film on the surface. A chromic acid concentration of 1 to 2% and a pH of about 1 are preferred.

  12. Modeling Thermal and Stress Behavior of the Fuel-clad Interface in Monolithic Fuel Mini-plates

    International Nuclear Information System (INIS)

    Miller, Gregory K.; Medvedev, Pavel G.; Burkes, Douglas E.; Wachs, Daniel M.

    2010-01-01

    As part of the Global Threat Reduction Initiative, a fuel development and qualification program is in process with the objective of qualifying very high density low enriched uranium fuel that will enable the conversion of high performance research reactors with operational requirements beyond those supported with currently available low enriched uranium fuels. The high density of the fuel is achieved by replacing the fuel meat with a single monolithic low enriched uranium-molybdenum fuel foil. Doing so creates differences in the mechanical and structural characteristics of the fuel plate because of the planar interface created by the fuel foil and cladding. Furthermore, the monolithic fuel meat will dominate the structural properties of the fuel plate rather than the aluminum matrix, which is characteristic of dispersion fuel types. Understanding the integrity and behavior of the fuel-clad interface during irradiation is of great importance for qualification of the new fuel, but can be somewhat challenging to determine with a single technique. Efforts aimed at addressing this problem are underway within the fuel development and qualification program, comprised of modeling, as-fabricated plate characterization, and post-irradiation examination. An initial finite element analysis model has been developed to investigate worst-case scenarios for the basic monolithic fuel plate structure, using typical mini-plate irradiation conditions in the Advanced Test Reactor. Initial analysis shows that the stress normal to the fuel-clad interface dominates during irradiation, and that the presence of small, rounded delaminations at the interface is not of great concern. However, larger and/or fuel-clad delaminations with sharp corners can create areas of concern, as maximum principal cladding stress, strain, displacement, and peak fuel temperature are all significantly increased. Furthermore, stresses resulting from temperature gradients that cause the plate to bow or buckle in

  13. EFFECT OF THE SCREENS RADIANT REFLECTANCE ON THERMAL TRANSPORT PROCESS IN THE CLADDING STRUCTURES

    Directory of Open Access Journals (Sweden)

    V. D. Sizov

    2016-01-01

    Full Text Available The article analyses variants of the heat insulating layers disposition in relation to the cladding load-carrying structures and demonstrates prime advantages and drawbacks of the three variants. The authors notice that from the heat-engineering viewpoint the variant with exterior side winterization is the most favourable. However, utilizing micromodules as heat-insulating layers screened with leafing aluminum makes it necessary to account for the screens reflecting power. It allows reducing the irradiating component in the combined value of thermal transport through the enclosure and consequently raises the structure thermal resistance or, with parity of these values, leads to lower thickness of the heat-insulating layer. The known data applied for calculating the total heat transmission helps demonstrate reduction of the general heat flux value by 1.4 times, and the heat transmission resistance by 1.76 m2 deg./W. This allows reducing thickness of the heat-insulating layer (with regard of two screens by 0.07 m. Computations illustrate the fact that account for the radiant reflectance of screening enables lowering the rated heat flux passing through the enclosure. Which again allows decreasing the structure thermal resistance and its general thickness (by 70 mm at the expense of small thickness of the heat insulation of micromodules. The humidity regime calculations establish good acceptability of the enclosure service conditions in winter. The period will see no real water vapour condensation. The plotted diagrams of the cladding heat-and-humidity conditions demonstrate that condensation zones do not affect the layer of thermal insulation (micromodules. And the condensation zone with reduction of the heat-insulating layer appears only during ‘severe’ outside temperature conditions of a cold month. Reduced to 230 mm thickness of the wall construction allows utilizing ‘old’ stock of forms with prefabricated panels in parallel with energy

  14. 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.

  15. The quest for safe and reliable fuel cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Pino, Eddy S.; Abe, Alfredo Y., E-mail: eddypino132@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    The tragic Fukushima Daiichi Nuclear Plant accident of March, 2011, has brought great unrest and challenge to the nuclear industry, which, in collaboration with universities and nuclear research institutes, is making great efforts to improve the safety in nuclear reactors developing accident tolerant fuels (ATF). This involves the study of different materials to be applied as cladding and, also, the improvement in the fuel properties in order to enhance the fuel performance and safety, specifically under accident conditions. Related to the cladding, iron based alloys and silicon carbide (SiC) materials have been studied as a good alternative. In the case of austenitic stainless steel, there is the advantage that the austenitic stainless steel 304 was used as cladding material in the first PWR (Pressurized Water Reactor) registering a good performance. Then, alternated cladding materials such as iron based alloys (304, 310, 316, 347) should be used to replace the zirconium-based alloys in order to improve safety. In this paper, these cladding materials are evaluated in terms of their physical and chemical properties; among them, strength and creep resistance, thermal conductivity, thermal stability and corrosion resistance. Additionally, these properties are compared with those of conventional zirconium-based alloys, the most used material in actual PWR, to assess the advantages and disadvantages of each material concerning to fuel performance and safety contribution. (author)

  16. POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

    Directory of Open Access Journals (Sweden)

    Vojtěch Caha

    2016-12-01

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

  17. Fuel-cladding chemical interaction in mixed-oxide fuels

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Weber, J.W.; Devary, J.L.

    1978-10-01

    The character and extent of fuel-cladding chemical interaction (FCCI) was established for UO 2 -25 wt% PuO 2 clad with 20% cold worked Type 316 stainless steel irradiated at high cladding temperatures to peak burnups greater than 8 atom %. The data base consists of 153 data sets from fuel pins irradiated in EBR-II with peak burnups to 9.5 atom %, local cladding inner surface temperatures to 725 0 C, and exposure times to 415 equivalent full power days. As-fabricated oxygen-to-metal ratios (O/M) ranged from 1.938 to 1.984 with the bulk of the data in the range 1.96 to 1.98. HEDL P-15 pins provided data at low heat rates, approx. 200 W/cm, and P-23 series pins provided data at higher heat rates, approx. 400 W/cm. A design practice for breeder reactors is to consider an initial reduction of 50 microns in cladding thickness to compensate for possible FCCI. This approach was considered to be a conservative approximation in the absence of a comprehensive design correlation for extent of interaction. This work provides to the designer a statistically based correlation for depth of FCCI which reflects the influences of the major fuel and operating parameters on FCCI

  18. Cladding Alloys for Fluoride Salt Compatibility Final Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-05-01

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

  19. The quest for safe and reliable fuel cladding materials

    International Nuclear Information System (INIS)

    Pino, Eddy S.; Abe, Alfredo Y.; Giovedi, Claudia

    2015-01-01

    The tragic Fukushima Daiichi Nuclear Plant accident of March, 2011, has brought great unrest and challenge to the nuclear industry, which, in collaboration with universities and nuclear research institutes, is making great efforts to improve the safety in nuclear reactors developing accident tolerant fuels (ATF). This involves the study of different materials to be applied as cladding and, also, the improvement in the fuel properties in order to enhance the fuel performance and safety, specifically under accident conditions. Related to the cladding, iron based alloys and silicon carbide (SiC) materials have been studied as a good alternative. In the case of austenitic stainless steel, there is the advantage that the austenitic stainless steel 304 was used as cladding material in the first PWR (Pressurized Water Reactor) registering a good performance. Then, alternated cladding materials such as iron based alloys (304, 310, 316, 347) should be used to replace the zirconium-based alloys in order to improve safety. In this paper, these cladding materials are evaluated in terms of their physical and chemical properties; among them, strength and creep resistance, thermal conductivity, thermal stability and corrosion resistance. Additionally, these properties are compared with those of conventional zirconium-based alloys, the most used material in actual PWR, to assess the advantages and disadvantages of each material concerning to fuel performance and safety contribution. (author)

  20. Performance of HT9 clad metallic fuel at high temperature

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Hayes, S.L.

    1992-01-01

    Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area

  1. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    International Nuclear Information System (INIS)

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong

    2012-01-01

    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. 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. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced

  2. Compatibility of niobium, titanium, and vanadium metals with LMFBR cladding

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1975-10-01

    A series of laboratory capsule annealing experiments were conducted to assess the compatibility of niobium, vanadium, and titanium with 316 stainless steel cladding in the temperature range of 700 to 800 0 C. Niobium, vanadium, and titanium are cantidate oxygen absorber materials for control of oxygen chemistry in LMFBR fuel pins. Capsule examination indicated good compatibility between niobium and 316 stainless steel at 800 0 C. Potential compatibility problems between cladding and vanadium or titanium were indicated at 800 0 C under reducing conditions. In the presence of Pu/sub 0.25/U/sub 0.75/O/sub 1.98/ fuel (Δanti G 02 congruent to -160 kcal/mole) no reaction was observed between vanadium or titanium and cladding at 800 0 C

  3. Chemical Dissolution of Simulant FCA Cladding and Plates

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-08

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

  4. Characterization Of Cladding Hull Wastes From Used Nuclear Fuels

    Directory of Open Access Journals (Sweden)

    Kang K.H.

    2015-06-01

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

  5. Performance of refractory alloy-clad fuel pins

    International Nuclear Information System (INIS)

    Dutt, D.S.; Cox, C.M.; Millhollen, M.K.

    1984-12-01

    This paper discusses objectives and basic design of two fuel-cladding tests being conducted in support of SP-100 technology development. Two of the current space nuclear power concepts use conventional pin type designs, where a coolant removes the heat from the core and transports it to an out-of-core energy conversion system. An extensive irradiation testing program was conducted in the 1950's and 1960's to develop fuel pins for space nuclear reactors. The program emphasized refractory metal clad uranium nitride (UN), uranium carbide (UC), uranium oxide (UO 2 ), and metal matrix fuels (UCZr and BeO-UO 2 ). Based on this earlier work, studies presented here show that UN and UO 2 fuels in conjunction with several refractory metal cladding materials demonstrated high potential for meeting space reactor requirements and that UC could serve as an alternative but higher risk fuel

  6. Clad photon sieve for generating localized hollow beams

    Science.gov (United States)

    Cheng, Yiguang; Tong, Junmin; Zhu, Jiangping; Liu, Junbo; Hu, Song; He, Yu

    2016-02-01

    A novel photon sieve structure called clad photon sieve is proposed to generate localized hollow beams and its design principle and focusing properties are studied. The clad photon sieve is composed of the internal zone and external zone with pinholes being positioned on the dark zones. Pinholes in the internal zone and in the external zone give destructive interference to the focus, leading to localized hollow beams being generated on the focal plane. Focusing properties of clad photon sieve with different focal lengths, zone numbers and modulation factors are also studied by theoretical calculations, numerical simulations and experiments, showing that the central dark spot size can be controlled by the focal length and rings number, and the intensity of the central dark spot varies with different modulation factors related with the internal zone and the external zone. This photon sieve can be useful for trapping and manipulating of particles and cooling of atoms.

  7. Compatibility of niobium, titanium, and vanadium metals with LMFBR cladding

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, C.N.

    1975-10-01

    A series of laboratory capsule annealing experiments were conducted to assess the compatibility of niobium, vanadium, and titanium with 316 stainless steel cladding in the temperature range of 700 to 800/sup 0/C. Niobium, vanadium, and titanium are cantidate oxygen absorber materials for control of oxygen chemistry in LMFBR fuel pins. Capsule examination indicated good compatibility between niobium and 316 stainless steel at 800/sup 0/C. Potential compatibility problems between cladding and vanadium or titanium were indicated at 800/sup 0/C under reducing conditions. In the presence of Pu/sub 0.25/U/sub 0.75/O/sub 1.98/ fuel (..delta..anti G/sub 02/ congruent to -160 kcal/mole) no reaction was observed between vanadium or titanium and cladding at 800/sup 0/C.

  8. Absorptivity Measurements and Heat Source Modeling to Simulate Laser Cladding

    Science.gov (United States)

    Wirth, Florian; Eisenbarth, Daniel; Wegener, Konrad

    The laser cladding process gains importance, as it does not only allow the application of surface coatings, but also additive manufacturing of three-dimensional parts. In both cases, process simulation can contribute to process optimization. Heat source modeling is one of the main issues for an accurate model and simulation of the laser cladding process. While the laser beam intensity distribution is readily known, the other two main effects on the process' heat input are non-trivial. Namely the measurement of the absorptivity of the applied materials as well as the powder attenuation. Therefore, calorimetry measurements were carried out. The measurement method and the measurement results for laser cladding of Stellite 6 on structural steel S 235 and for the processing of Inconel 625 are presented both using a CO2 laser as well as a high power diode laser (HPDL). Additionally, a heat source model is deduced.

  9. Influence Of The Laser Cladding Strategies On The Mechanical Properties Of Inconel 718

    International Nuclear Information System (INIS)

    Lamikiz, A.; Tabernero, I.; Ukar, E.; Lopez de Lacalle, L. N.; Delgado, J.

    2011-01-01

    This work presents different experimental results of the mechanical properties of Inconel registered 718 test parts built-up by laser cladding. Recently, turbine manufacturers for aeronautical sector have presented high interest on laser cladding processes. This process allows building fully functional structures on superalloys, such as Inconel registered 718, with high flexibility on complex shapes. However, there is limited data on mechanical properties of the laser cladding structures. Moreover, the available data do not include the influence of process parameters and laser cladding strategies. Therefore, a complete study of the influence of the laser cladding parameters and mainly, the variation of the tensile strength with the laser cladding strategy is presented. The results show that there is a high directionality of mechanical properties, depending on the strategies of laser cladding process. In other words, the test parts show a fiber -like structure that should be considered on the laser cladding strategy selection.

  10. Low-Stress Silicon Cladding for Surface Finishing Large UVOIR Mirrors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this Phase I research, ZeCoat Corporation demonstrated a low-stress silicon cladding process for surface finishing large UVOIR mirrors. A polishable cladding is...

  11. Gap conductance in Zircaloy-clad LWR fuel rods

    International Nuclear Information System (INIS)

    Ainscough, J.B.

    1982-04-01

    This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO 2 in Zircaloy cladding, under both steady-state and transient conditions. The relevant theory is discussed together with some of the approximations usually made in performance modelling codes. The state of the physical property data which are needed for heat transfer calculations is examined and some of the relevant in- and out-of-reactor experimental work on fuel rod conductance is reviewed

  12. Comparison of models discribing cladding deformations during LOCA

    International Nuclear Information System (INIS)

    Chakraborty, A.K.; Zipper, R.

    1981-05-01

    This report compares the important models for the determination of cladding deformations during LOCA. In addition to the comparisons of underlying assumptions of different models the same is done for the coefficients applied for the models. In order to assess the predictive capability of the models the calculated results are compared with the experimental results of the individual claddings. It was found out that the results of temperature ramp tests could be calculated better than that of the pressure ramp tests. The calculations revealed that even with the simplified assumption of the model used in TESPA the agreement of the calculated results with those of model NORA was relatively good. (orig.) [de

  13. Cladding dimensional changes in mixed oxide fuel pins

    International Nuclear Information System (INIS)

    Makenas, B.J.; Jost, J.W.; Hales, J.W.

    1980-01-01

    Several types of stainless steel (304, 316, 316-20% Cold Worked, and 316 Titanium stabilized 20% CW) have been used as cladding for mixed oxide (U, Pu)O 2 fuel pins irradiated in EBR-II. All of the materials have performed satisfactorily in their respective experimental subassemblies but significant differences in swelling and inelastic strain behavior have been found at high fast fluences among the different materials and among different heats of the same material. Cladding diameter increases were measured for 622 developmental and FFTF reference design fuel pins which were irradiated in 19 experimental subassemblies. Fuel pin diameters were determined from multiangle axial trace profilometry measurements

  14. Technical committee meeting on fuel and cladding interaction. Summary report

    International Nuclear Information System (INIS)

    1977-04-01

    Experiments and experiences concerning fuel-cladding interaction in thermal and fast neutron flux burnup are dealt with. A number of results from in-pile and out-of pile experiments with different fuel pins with cladding made of different stainless steels showed the importance of corrosion process, dependent on the burnup, core temperature, metal-oxide ratio, and other steady state parameters in the core of fast reactors (most frequently LMFBRs). This is of importance for fuel pins design and fabrication. Mixed oxide fuel is treated in many cases

  15. Synthesis of clad motion experiments interpretation: codes and validation

    International Nuclear Information System (INIS)

    Papin, J.; Fortunato, M.; Seiler, J.M.

    1983-04-01

    This communication deals with clad melting and relocation phenomena related to LMFBR safety analysis of loss of flow accidents. We present: - the physical models developed at DSN/CEN Cadarache in single channel and bundle geometry. The interpretation with these models of experiments performed by the STT (CEN Grenoble). It comes out that we have now obtained a good understanding of the involved phenomena in single channel geometry. On the other hand, further studies are necessary for a better knowledge of clad motion phenomena in bundle cases with conditions close to reactor ones

  16. Irradiation experience with HT9-clad metallic fuel

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  17. Modelling of pellet-cladding interaction in PWR's

    International Nuclear Information System (INIS)

    Esteves, A.M.; Silva, A.T. e.

    1992-01-01

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyses the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. (author)

  18. Aluminum industry options paper

    International Nuclear Information System (INIS)

    1999-10-01

    In 1990, Canada's producers of aluminum (third largest in the world) emitted 10 million tonnes of carbon dioxide and equivalent, corresponding to 6.4 tonnes of greenhouse gas intensity per tonne of aluminum. In 2000, the projection is that on a business-as-usual (BAU) basis Canadian producers now producing 60 per cent more aluminum than in 1990, will emit 10.7 million tonnes of carbon dioxide and equivalent, corresponding to a GHG intensity of 4.2 tonnes per tonne of aluminum. This improvement is due to production being based largely on hydro-electricity, and partly because in general, Canadian plants are modern, with technology that is relatively GHG-friendly. The Aluminum Association of Canada estimates that based on anticipated production, and under a BAU scenario, GHG emissions from aluminum production will rise by 18 per cent by 2010 and by 30 per cent by 2020. GHG emissions could be reduced below the BAU forecast first, by new control and monitoring systems at some operations at a cost of $4.5 to 7.5 million per smelter. These systems could reduce carbon dioxide equivalent emissions by 0.8 million tonnes per year. A second alternative would require installation of breaker feeders which would further reduce perfluorocarbon (PFC) emissions by 0.9 million tonnes of carbon dioxide equivalent. Cost of the breakers feeders would be in the order of $200 million per smelter. The third option calls for the the shutting down of some of the smelters with older technology by 2015. In this scenario GHG emissions would be reduced by 2010 by 0.8 million tonnes per year of carbon dioxide equivalent. However, the cost in this case would be about $1.36 billion. The industry would support measures that would encourage the first two sets of actions, which would produce GHG emissions from aluminum production in Canada of about 10.2 million tonnes per year of carbon dioxide equivalent, or about two per cent above 1990 levels with double the aluminum production of 1990. Credit for

  19. Engineering of metal-clad optical nanocavity to optimize coupling with integrated waveguides

    OpenAIRE

    Kim, Myung-Ki; Li, Zheng; Huang, Kun; Going, Ryan; Wu, Ming C.; Choo, Hyuck

    2013-01-01

    We propose a cladding engineering method that flexibly modifies the radiation patterns and rates of metal-clad nanoscale optical cavity. Optimally adjusting the cladding symmetry of the metal-clad nanoscale optical cavity modifies the modal symmetry and produces highly directional radiation that leads to 90% coupling efficiency into an integrated waveguide. In addition, the radiation rate of the cavity mode can be matched to its absorption rate by adjusting the thickness of the bottom-claddin...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-14

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

  1. Gradient microstructure in laser clad TiC-reinforced Ni-alloy composite coating

    NARCIS (Netherlands)

    Pei, Y.T.; Zuo, T.C.

    1998-01-01

    A gradient TiC–(Ni alloy) composite coating was produced by one step laser cladding with pre-placed mixture powder on a 1045 steel substrate. The clad layers consisted of TiC particles, γ-Ni primary dendrites and interdendritic eutectics. From the bottom to the top of the clad layer produced at 2000

  2. Aluminum for plasmonics.

    Science.gov (United States)

    Knight, Mark W; King, Nicholas S; Liu, Lifei; Everitt, Henry O; Nordlander, Peter; Halas, Naomi J

    2014-01-28

    Unlike silver and gold, aluminum has material properties that enable strong plasmon resonances spanning much of the visible region of the spectrum and into the ultraviolet. This extended response, combined with its natural abundance, low cost, and amenability to manufacturing processes, makes aluminum a highly promising material for commercial applications. Fabricating Al-based nanostructures whose optical properties correspond with theoretical predictions, however, can be a challenge. In this work, the Al plasmon resonance is observed to be remarkably sensitive to the presence of oxide within the metal. For Al nanodisks, we observe that the energy of the plasmon resonance is determined by, and serves as an optical reporter of, the percentage of oxide present within the Al. This understanding paves the way toward the use of aluminum as a low-cost plasmonic material with properties and potential applications similar to those of the coinage metals.

  3. Aluminum Hydroxide and Magnesium Hydroxide

    Science.gov (United States)

    Aluminum Hydroxide, Magnesium Hydroxide are antacids used together to relieve heartburn, acid indigestion, and upset stomach. They ... They combine with stomach acid and neutralize it. Aluminum Hydroxide, Magnesium Hydroxide are available without a prescription. ...

  4. Regeneration of aluminum hydride

    Science.gov (United States)

    Graetz, Jason Allan; Reilly, James J.

    2009-04-21

    The present invention provides methods and materials for the formation of hydrogen storage alanes, AlH.sub.x, where x is greater than 0 and less than or equal to 6 at reduced H.sub.2 pressures and temperatures. The methods rely upon reduction of the change in free energy of the reaction between aluminum and molecular H.sub.2. The change in free energy is reduced by lowering the entropy change during the reaction by providing aluminum in a state of high entropy, by increasing the magnitude of the change in enthalpy of the reaction or combinations thereof.

  5. Regeneration of aluminum hydride

    Science.gov (United States)

    Graetz, Jason Allan; Reilly, James J; Wegrzyn, James E

    2012-09-18

    The present invention provides methods and materials for the formation of hydrogen storage alanes, AlH.sub.x, where x is greater than 0 and less than or equal to 6 at reduced H.sub.2 pressures and temperatures. The methods rely upon reduction of the change in free energy of the reaction between aluminum and molecular H.sub.2. The change in free energy is reduced by lowering the entropy change during the reaction by providing aluminum in a state of high entropy, and by increasing the magnitude of the change in enthalpy of the reaction or combinations thereof.

  6. Fundamental metallurgical aspects of axial splitting in zircaloy cladding

    International Nuclear Information System (INIS)

    Chung, H. M.

    2000-01-01

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

  7. Advances in appendage joining techniques for PHWR fuel cladding

    International Nuclear Information System (INIS)

    Desai, P.B.; Ray, T.K.; Date, V.G.; Purushotham, D.S.C.

    1995-01-01

    This paper describes work carried out at the BARC on the development of a technique to join tiny appendages (spacers and bearing pads) to thin cladding (before loading of UO 2 pellets) by resistance welding for PHWR fuel assemblies. The work includes qualifying the process for production environment, designing prototype equipment for regular production and quality monitoring. In the first phase of development, welding of appendages on UO 2 loaded elements was successfully developed, and is being used in production. Welding of appendages on to empty clad tubes is a superior technique for several reasons. Many problems associated with development of welding on empty tubes were resolved. work was initiated, in the second phase of the development task, to select a suitable technique to join appendages on empty clad tubes without any collapse of thin clad. Several alternatives were reviewed and assessed such as laser, full face welding, shim welding and shrink fitting ring spacers. Selection of a method using a mandrel and a modified electrode geometry was fully developed. Results were optimized and process development successfully completed. Appropriate weld monitoring techniques were also reviewed for their adaptation. This technique is useful for 19, 22 as well as 37 element assemblies. (author)

  8. Prediction of cladding life in waste package environments

    International Nuclear Information System (INIS)

    McCoy, J.K.; Doering, T.W.

    1994-01-01

    Fuel cladding can potentially provide longer containment or slower release of radionuclides from spent fuel after geologic disposal. To predict the amount of benefit that cladding can provide, we surveyed degradation modes and developed a model for creep rupture by diffusion-controlled cavity growth, the mechanism that several authors have concluded is the most important. In this mechanism, voids nucleate on the grain boundaries and grow by diffusion of vacancies along the grain boundaries to the voids. When a certain fraction of the grain boundary area is covered with voids, the material fails. An analytic expression for cladding lifetime is developed. Besides materials constants, the predicted lifetime depends on the temperature history, the hoop stress in the cladding, the spacing between void nuclei, and the micro-structure. The inclusion of microstructure is a significant new feature of the model; this feature is used to help avoid excessive conservatism. The model is applied in a sample calculation for disposal of spent fuel, and the practice of using temperature limits to evaluate repository designs is examined

  9. Production and quality control of fuel cladding tubes for LWRs

    International Nuclear Information System (INIS)

    Matsuda, Katsuhiko; Hagi, Shigeki; Anada, Hiroyuki; Abe, Hideaki; Hyodo, Shigetoshi

    1994-01-01

    This paper reviews the recent fabrication technology and corrosion resistance study of fuel cladding tubes for LWRs conducted by Sumitomo Metal Industries Ltd. started the research on zircaloy in 1957. In 1980, the factory exclusively for the production of cladding tubes was founded, and the mass production system on full scale was established. Thereafter, the various improvement of the production technology, the development of new products, and the heightening of the performance mainly on the corrosion resistance have been tested and studied. Recently, the works in the production processes were almost automated, and the installation of the production lines advanced, and the stabilization of product quality and the rationalization of costs are promoted. Moreover, the development of the zircaloy cladding tubes having high corrosion resistance has been advanced to cope with the long term cycle operation of LWRs hereafter. The features of zircaloy cladding tubes, the manufacturing processes, the improvement of the manufacturing technology, the improvement of the corrosion resistance and so on are reported. (K.I.)

  10. Clad failure detection in G 3 - operational feedback

    International Nuclear Information System (INIS)

    Plisson, J.

    1964-01-01

    After briefly reviewing the role and the principles of clad failure detection, the author describes the working conditions and the conclusions reached after 4 years operation of this installation on the reactor G 3. He mentions also the modifications made to the original installation as well as the tests carried out and the experiments under way. (author) [fr

  11. Radioluminescent clad optical fibre X-ray sensor

    NARCIS (Netherlands)

    Fitzpatrick, C.; O'Donoghue, C.; Lewis, E.; Schöbel, J.; Bastaiens, B.; Bastiaens, Hubertus M.J.; van der Slot, Petrus J.M.

    2003-01-01

    An optical fibre X-ray sensor using cladding to core coupling of radioluminescent emissions from X-ray phosphors is presented. The sensor is tested using a line source normally used in the pre-ionisation of an excimer laser and is calibrated against a pendosimeter and the emission intensity of a

  12. Microstrain Determination in Individual Grains of Laser Deposited Cladding Layers

    NARCIS (Netherlands)

    de Oliveira, Uazir O. B.; Ocelik, Vaclav; De Hosson, Jeff T. M.; Chandra, T; Tsuzaki, K; Militzer, M; Ravindran, C

    2007-01-01

    The laser cladding technique makes the deposition of thick metallic, wear and corrosion resistant coatings feasible on weaker substrates. During the process, localized high thermal gradients generate internal stresses that may cause cracking when these overcome the fracture stress. To explain the

  13. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. First, a thermal creep damage index is set up through a sufficiently sophisticated clad physical analysis including arbitrary time dependence of power and neutron flux as well as effects of sodium temperature, burnup and steel mechanical behavior. Although this strain limit approach implies a more general but time consuming model., on the counterpart the net output is improved and e.g. clad temperature, stress and strain maxima may be easily assessed. A full spectrum of variables are statistically treated to account for their probability distributions. Creep damage probability may be obtained and can contribute to a quantitative fuel probability estimation

  14. Achilles tests finally nail PWR fuel clad ballooning fears

    International Nuclear Information System (INIS)

    Dore, P.; McMinn, K.

    1992-01-01

    A conclusive series of experiments carried out by AEA Reactor Services at its Achilles rig in the UK has finally allayed fears that fuel clad ballooning is a major safety problem for Sizewell B, Britain's first Pressurized Water Reactor. The experiments are described in this article. (author)

  15. Viscoelastic modelling of Zircaloy cladding in-pile transient creep

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-15

    In fuel behaviour modelling accurate description of the cladding stress response is important for both operational and safety considerations. The cladding creep determines in part the width of the gas gap, the duration to pellet-cladding contact and the stresses to the cladding due to the pellet expansion. Conventionally the strain hardening rule has been used to describe the creep response to transient loads in engineering applications. However, it has been well documented that the strain hardening rule does not describe well results of tests with load drops or reversals. In our earlier work we have developed a model for primary creep which can be used to simulate the in- and out-of-pile creep tests. Since then several creep experiments have entered into public domain. In this paper we develop the model formulation based on the theory of viscoelasticity, and show that this model can reproduce the new experimental results. We also show that the creep strain recovery encountered in experimental measurements can be explained by viscoelastic behaviour.

  16. Delayed hydride cracking of Zircaloy-4 fuel cladding

    International Nuclear Information System (INIS)

    Pizarro, Luis M.; Fernandez, Silvia; Lafont, Claudio; Mizrahi, Rafael; Haddad, Roberto

    2007-01-01

    Crack propagation rates, grown by the delayed hydride cracking mechanism, were measured in Zircaloy-4 fuel cladding, according to a Coordinated Research Project (CRP) sponsored by the International Atomic Energy Agency (IAEA). During the first stage of the program a Round Robin Testing was performed on fuel cladding samples provided by Studsvik (Sweden), of the type used in PWR reactors. Crack growth in the axial direction is obtained through the specially developed 'pin load testing' (PLT) device. In these tests, crack propagation rates were determined at 250 C degrees on several samples of the material described above, obtaining a mean value of about 2.5 x 10 -8 m s -1 . The results were analyzed and compared satisfactorily with those obtained by the other laboratories participating in the CRP. At the present moment, similar tests on CANDU and Atucha I type fuel cladding are being performed. It is thought that the obtained results will give valuable information concerning the analysis of possible failures affecting fuel cladding under reactor operation. (author) [es

  17. An Innovative Ceramic Corrosion Protection System for Zircaloy Cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-02-19

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

  18. Hydrogenation and high temperature oxidation of Zirconium claddings

    International Nuclear Information System (INIS)

    Novotny, T.; Perez-Feró, E.; Horváth, M.

    2015-01-01

    In the last few years a new series of experiments started for supporting the new LOCA criteria, considering the proposals of US NRC. The effects which can cause the embrittlement of VVER fuel claddings were reviewed and evaluated in the framework of the project. The purpose of the work was to determine how the fuel cladding’s hydrogen uptake under normal operating conditions, effect the behavior of the cladding under LOCA conditions. As a first step a gas system equipment with gas valves and pressure gauge was built, in which the zirconium alloy can absorb hydrogen under controlled conditions. In this apparatus E110 (produced by electrolytic method, currently used at Paks NPP) and E110G (produced by a new technology) alloys were hydrogenated to predetermined hydrogen contents. According the results of ring compression tests the E110G alloys lose their ductility above 3200 ppm hydrogen content. This limit can be applied to determine the ductile-brittle transition of the nuclear fuel claddings. After the hydrogenation, high temperature oxidation experiments were carried out on the E110G and E110 samples at 1000 °C and 1200 °C. 16 pieces of E110G and 8 samples of E110 with 300 ppm and 600 ppm hydrogen content were tested. The oxidation of the specimens was performed in steam, under isothermal conditions. Based on the ring compression tests load-displacement curves were recorded. The main objective of the compression tests was to determine the ductile-brittle transition. These results were compared to the results of our previous experiments where the samples did not contain hydrogen. The original claddings showed more ductile behavior than the samples with hydrogen content. The higher hydrogen content resulted in a more brittle mechanical behavior. However no significant difference was observed in the oxidation kinetics of the same cladding types with different hydrogen content. The experiments showed that the normal operating hydrogen uptake of the fuel claddings

  19. Steam attack studies in SiC clad and advanced steel clad

    International Nuclear Information System (INIS)

    Terrani, Kurt; Snead, Lance; Pint, Bruce

    2013-01-01

    The most recent generation of zirconium alloys used as nuclear fuel cladding in LWRs offer exceptional performance under normal operating conditions with failure rates below 1 ppm. This level of performance is due to six decades of active research and development by nuclear operators, vendors, government laboratories, and academia. During this same period, iron alloys, which were successfully used in the early days of nuclear power but supplanted by zirconium alloys, have undergone broad-based development including specialty specialised steels for high-temperature corrosion and steam resistance. This coupled with the recent trends in fuel handling, reactor operation and the extreme control of coolant chemistry begs for re-examination of the application of advanced iron-alloys as LWR fuel cladding, specifically, oxidation resistant iron alloys that offer large margins of safety under severe accident conditions. Recently, steam oxidation tests have been performed on a wide range of austenitic and ferritic iron alloys at ORNL where the dependence of the oxidation kinetics on alloy chemistry has been determined. As the temperature increases the minimum Cr content in the ferritic and austenitic alloys needs to be increased for oxidation resistance. At temperatures ≥1200 deg. C ∼25% Cr content is necessary for adequate oxidation resistance. At the fixed Cr content, an increase in Ni content in the austenitic alloys enhances the oxidation resistance. Note that the 25% minimum Cr content disqualifies the conventional austenitic (18Cr-8Ni) alloys such as 304L as oxidation resistant materials in steam. The most promising alloys are ferritic alloys that contain some Al in addition to Cr. FeCrAl alloys with 20% Cr and 5% Al exhibit exceptionally low oxidation in high-temperature steam up to 1300 deg. C due to formation of a protective alumina film on the surface of the material. High-pressure steam oxidation tests (up to 2 MPa) were also performed that showed the effect of

  20. Analysis of steam explosions in plate-type, uranium-aluminum fuel test reactors

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1989-01-01

    The concern over steam explosions in nuclear reactors can be traced to prompt critical nuclear excursions in aluminum-clad/fueled test reactors, as well as to explosive events in aluminum, pulp, and paper industries. The Reactor Safety Study prompted an extensive analytical and experimental effort for over a decade. This has led to significant improvements in their understanding of the steam explosion issue for commercial light water reactors. However, little progress has been made toward applying the lessons learned from this effort to the understanding and modeling of steam explosion phenomena in aluminum-clad/fueled research and test reactors. The purposes of this paper are to (a) provide a preliminary analysis of the destructive events in test reactors, based on current understandings of steam explosions; (b) provide a proposed approach for determining the likelihood of a steam explosion event under scenarios in which molten U-Al fuel drops into a water-filled cavity; and (c) present a benchmarking study conducted to estimate peak pressure pulse magnitudes

  1. Thermal gradient effects on the oxidation of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Klein, A.C.; Reyes, J.N. Jr.; Maguire, M.A.

    1990-01-01

    A Thermal Gradient Test Facility (TGTF) has been designed and constructed to measure the thermal gradient effect on pressurized water reactor (PWR) fuel rod cladding. The TGTF includes a heat flux simulator assembly capable of producing a wide range of PWR operating conditions including water flow velocities and temperatures, water chemistry conditions, cladding temperatures, and heat fluxes ranging to 160 W/cm 2 . It is fully instrumented including a large number of thermocouples both inside the water flow channel and inside the cladding. Two test programs are in progress. First, cladding specimens are pre-oxidized in air at 500 deg. C and in 400 deg. C steam for various lengths of time to develop a range of uniform oxide thicknesses from 1 to 60 micrometers. The pre-oxidized specimens are placed in the TGTF to characterize the oxide thermal conductivity under a variety of water flow and heat flux conditions. Second, to overcome the long exposure times required under typical PWR conditions a series of tests with the addition of high concentrations of lithium hydroxide to the water are being considered. Static autoclave tests have been conducted with lithium hydroxide concentrations ranging from 0 to 2 moles per liter at 300, 330, and 360 deg. C for up to 36 hours. Results for zircaloy-4 show a considerable increase in the weight gain for the exposed samples with oxidation rate enhancement factors as high as 70 times that of pure water. Operation of the TGTF with elevated lithium hydroxide levels will yield real-time information concerning the effects of a heat flux on the oxidation kinetics of zircaloy fuel rod cladding. (author). 5 refs, 5 figs, 2 tabs

  2. Aluminum Sulfate 18 Hydrate

    Science.gov (United States)

    Young, Jay A.

    2004-01-01

    A chemical laboratory information profile (CLIP) of the chemical, aluminum sulfate 18 hydrate, is presented. The profile lists physical and harmful properties, exposure limits, reactivity risks, and symptoms of major exposure for the benefit of teachers and students using the chemical in the laboratory.

  3. Applied Electrochemistry of Aluminum

    DEFF Research Database (Denmark)

    Li, Qingfeng; Qiu, Zhuxian

    Electrochemistry of aluminum is of special importance from both theoretical and technological point of view. It covers a wide range of electrolyte systems from molten fluoride melts at around 1000oC to room temperature molten salts, from aqueous to various organic media and from liquid to solid...

  4. Invisible Display in Aluminum

    DEFF Research Database (Denmark)

    Prichystal, Jan Phuklin; Hansen, Hans Nørgaard; Bladt, Henrik Henriksen

    2005-01-01

    for an integrated display in a metal surface is often ruled by design and functionality of a product. The integration of displays in metal surfaces requires metal removal in order to clear the area of the display to some extent. The idea behind an invisible display in Aluminum concerns the processing of a metal...

  5. Aluminum for Plasmonics

    Science.gov (United States)

    2014-01-01

    mini - mize the deleterious effects of the bulk metal oxide. Conversely, the optical scattering spectrum of an Al nanodisk can serve as a reporter of Al...Nanoparticles. J. Phys. Chem. C 2008, 112, 13958–13963. 22. Chowdhury, M. H.; Ray, K.; Gray, S. K.; Pond , J.; Lakowicz, J. R. Aluminum Nanoparticles as

  6. Aluminum battery alloys

    Science.gov (United States)

    Thompson, David S.; Scott, Darwin H.

    1985-01-01

    Aluminum alloys suitable for use as anode structures in electrochemical cs are disclosed. These alloys include iron levels higher than previously felt possible, due to the presence of controlled amounts of manganese, with possible additions of magnesium and controlled amounts of gallium.

  7. Borated aluminum alloy manufacturing technology

    International Nuclear Information System (INIS)

    Shimojo, Jun; Taniuchi, Hiroaki; Kajihara, Katsura; Aruga, Yasuhiro

    2003-01-01

    Borated aluminum alloy is used as the basket material of cask because of its light weight, thermal conductivity and superior neutron absorbing abilities. Kobe Steel has developed a unique manufacturing process for borated aluminum alloy using a vacuum induction melting method. In this process, aluminum alloy is melted and agitated at higher temperatures than common aluminum alloy fabrication methods. It is then cast into a mold in a vacuum atmosphere. The result is a high quality aluminum alloy which has a uniform boron distribution and no impurities. (author)

  8. An allowable cladding peak temperature for spent nuclear fuels in interim dry storage

    Science.gov (United States)

    Cha, Hyun-Jin; Jang, Ki-Nam; Kim, Kyu-Tae

    2018-01-01

    Allowable cladding peak temperatures for spent fuel cladding integrity in interim dry storage were investigated, considering hydride reorientation and mechanical property degradation behaviors of unirradiated and neutron irradiated Zr-Nb cladding tubes. Cladding tube specimens were heated up to various temperatures and then cooled down under tensile hoop stresses. Cool-down specimens indicate that higher heat-up temperature and larger tensile hoop stress generated larger radial hydride precipitation and smaller tensile strength and plastic hoop strain. Unirradiated specimens generated relatively larger radial hydride precipitation and plastic strain than did neutron irradiated specimens. Assuming a minimum plastic strain requirement of 5% for cladding integrity maintenance in interim dry storage, it is proposed that a cladding peak temperature during the interim dry storage is to keep below 250 °C if cladding tubes are cooled down to room temperature.

  9. Effect of Co - based Alloy on Properties of Laser Cladding Layer

    Science.gov (United States)

    Yang, Y.; Jiang, Z. P.; Li, H. Z.

    2017-11-01

    A large number of laser cladding experiments have been carried out using 20CrMnTi steel as substrate and Co-based alloy as cladding material. The influence of Co-based alloy on the laser cladding properties of 20CrMnTi steel was studied by analyzing the macroscopic and microscopic characteristics of cladding crack susceptibility, dilution rate, microstructure and friction and wear properties. The results show that the high-power laser cladding of Co-based material can obtain a flat defect-free cladding layer with compact structure and low crack susceptibility. A multi-layer cladding strategy with variable power can be used to fabricate thin wall structures without collapse Parts, the surface smooth without pores.

  10. Corrosion of research reactor aluminium clad spent fuel in water

    International Nuclear Information System (INIS)

    2009-12-01

    A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

  11. Abundance of interstellar aluminum

    Science.gov (United States)

    Barker, E. S.; Lugger, P. M.; Weiler, E. J.; York, D. G.

    1984-01-01

    New observations of Al II 1670 A, the only line of the dominant ionization stage of interstellar aluminum detected to date, are presented. Observations of ionized silicon are used to define an empirical curve of growth from which aluminum depletions can be derived. The depletion ranges from a factor of 10 in alpha Vir, with E(B-V) of about 0.04, to a factor of 1000 in omicron Per. The depletion is similar to that of iron, but a factor of 2-10 lower than that for silicon in the same stars. The observations of near-UV lines using the Copernicus V1 tubes with removal of a high cosmic-ray-induced fluorescent background are described.

  12. Behavior of aluminum in aluminum welders and manufacturers of aluminum sulfate--impact on biological monitoring.

    Science.gov (United States)

    Riihimäki, Vesa; Valkonen, Sinikka; Engström, Bernt; Tossavainen, Antti; Mutanen, Pertti; Aitio, Antero

    2008-12-01

    The suitability of determining aluminum in serum or urine as a form of biological monitoring was critically assessed. Airborne and internal aluminum exposure was assessed for 12 aluminum welders in a shipyard and 5 manufacturers of aluminum sulfate. Particles were characterized with X-ray diffraction and scanning electron microscopy. Aluminum in air and biological samples was analyzed using electrothermal atomic absorption spectrometry. Basic toxicokinetic features were inferred from the data. The mean 8-hour time-weighted average concentration of aluminum was 1.1 (range 0.008-6.1) mg/m(3) for the shipyard and 0.13 (range 0.02-0.5) mg/m(3) for the aluminum sulfate plant. Welding fume contained aluminum oxide particles aluminum sulfate particles ranged from 1 to 10 microm in diameter. The shipyard welders' mean postshift serum and urinary concentrations of aluminum (S-Al and U-Al, respectively) were 0.22 and 3.4 micromol/l, respectively, and the aluminum sulfate workers' corresponding values were 0.13 and 0.58 micromol/l. Between two shifts, the welders' S-Al concentration decreased by about 50% (Paluminum sulfate workers. After aluminum welding at the shipyard had ceased, the median S-Al concentration decreased by about 50% (P=0.007) within a year, but there was no change (P=0.75) in the corresponding U-Al concentration. About 1% of aluminum in welding fume appears to be rapidly absorbed from the lungs, whereas an undetermined fraction is retained and forms a lung burden. A higher fractional absorption of aluminum seems possible for aluminum sulfate workers without evidence of a lung burden. After rapid absorption, aluminum is slowly mobilized from the lung burden and dominates the S-Al and U-Al concentrations of aluminum welders. For kinetic reasons, S-Al or U-Al concentrations cannot be used to estimate the accumulation of aluminum in the target organs of toxicity. However, using U-Al analysis to monitor aluminum welders' lung burden seems practical.

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

    International Nuclear Information System (INIS)

    Kawai, Masayoshi; Kikuchi, Kenji; Kurishita, Hiroaki; Li, J.-F.; Furusaka, Michihiro

    2001-01-01

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

  14. Development of performance improvement of the cladding by ion beam

    International Nuclear Information System (INIS)

    Jung, Ki Sok; Kim, W.; Lee, J. H.; Choi, B. H.; Han, J. K.; Yoon, Y. H.; Lee, J. S.; Kwon, H. S.

    1997-07-01

    Ion implantation for the enhancement of the mechanical properties and the corrosion properties of the nuclear fuel cladding material has been conducted. Nitrogen implantations at 120 kV, > 5x10 17 ions/cm 2 , and >500 deg C resulted in the ZrN formation at the substrate surfaces. Nitrogen implantation at the elevated temperature in the oxygen atmosphere resulted in the great enhancement of the mechanical properties such as wear hardnesses. The hardness enhancement was especially significant up to 3 times at or above 500 deg C and 3x10 17 ions/cm 2 . The fretting wear resistance could be enhanced by 2 times at a dose of 8x10 17 ions/cm 2 . For the dedicated implanter of the cladding, main parts of the system were fabricated, e.g. ion sources, analyzing magnet, acceleration tube, vacuum lines, and the control programs. (author). 96 refs., 88 figs

  15. Modeling of realistic cladding structures for photonic bandgap fibers

    DEFF Research Database (Denmark)

    Mortensen, Niels Asger; Nielsen, Martin Dybendal

    2004-01-01

    Cladding structures of photonic bandgap fibers often have airholes of noncircular shape, and, typically, close-to-hexagonal airholes with curved corners are observed. We study photonic bandgaps in such structures by aid of a two-parameter representation of the size and curvature. For the fundamen......Cladding structures of photonic bandgap fibers often have airholes of noncircular shape, and, typically, close-to-hexagonal airholes with curved corners are observed. We study photonic bandgaps in such structures by aid of a two-parameter representation of the size and curvature....... For the fundamental bandgap we find that the bandgap edges (the intersections with the air line) shift toward shorter wavelengths when the air-filling fraction f is increased. The bandgap also broadens, and the relative bandwidth increases exponentially with f2. Compared with recent experiments [Nature 424, 657 (2003...

  16. Fracture resistance of irradiated stainless steel clad vessels

    International Nuclear Information System (INIS)

    McCabe, D.E.

    1988-01-01

    The surface crack embedded in the clad layer of a reactor vessel has been identified as a critical safety assessment condition relative to the pressurized thermal shock accident scenario. This project was initiated to determine the severity of such cracks experimentally, using irradiated material, and to identify the material property and stress conditions in the local region of the crack that are significant to the analysis. Bend bar tests provided the experimental simulation of the subject RPV surface crack. This report covers analysis techniques used and presents the findings indicated by the experimental results for irradiated and unirradiated materials. The irradiated clad specimens are the first of this type aimed at representing actual conditions in the wall of an irradiated pressure vessel. (author)

  17. Solution-mediated cladding doping of commercial polymer optical fibers

    Science.gov (United States)

    Stajanca, Pavol; Topolniak, Ievgeniia; Pötschke, Samuel; Krebber, Katerina

    2018-03-01

    Solution doping of commercial polymethyl methacrylate (PMMA) polymer optical fibers (POFs) is presented as a novel approach for preparation of custom cladding-doped POFs (CD-POFs). The presented method is based on a solution-mediated diffusion of dopant molecules into the fiber cladding upon soaking of POFs in a methanol-dopant solution. The method was tested on three different commercial POFs using Rhodamine B as a fluorescent dopant. The dynamics of the diffusion process was studied in order to optimize the doping procedure in terms of selection of the most suitable POF, doping time and conditions. Using the optimized procedure, longer segment of fluorescent CD-POF was prepared and its performance was characterized. Fiber's potential for sensing and illumination applications was demonstrated and discussed. The proposed method represents a simple and cheap way for fabrication of custom, short to medium length CD-POFs with various dopants.

  18. Pulsed Magnetic Welding for Advanced Core and Cladding Steel

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-19

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

  19. Modeling of Zircaloy cladding degradation under repository conditions

    International Nuclear Information System (INIS)

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

    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

  20. Modeling of MOX Fuel Pellet-Clad Interaction Using ABAQUS

    International Nuclear Information System (INIS)

    Ambrosek, Richard G.; Pedersen, Robert C.; Maple, Amanda

    2002-01-01

    Post-irradiation examination (PIE) has indicated an increase in the outer diameter of fuel pins being irradiated in the Advanced Test Reactor (ATR) for the MOX irradiation program. The diameter increase is the largest in the region between fuel pellets. The fuel pellet was modeled using PATRAN and the model was evaluated using ABAQUS, version 6.2. The results from the analysis indicate the non-uniform clad diameter is caused by interaction between the fuel pellet and the clad. The results also demonstrate that the interaction is not uniform over the pellet axial length, with the largest interaction occurring in the region of the pellet-pellet interface. Results were obtained for an axisymmetric model and for a 1/8 pie shaped segment, using the coupled temperature-displacement solution technique. (authors)

  1. The characterization of activities associated with irradiated fuel element claddings

    International Nuclear Information System (INIS)

    Jenkins, I.L.; Bolus, D.J.; Glover, K.M.; Haynes, J.W.; Mapper, D.; Marwick, A.D.; Waterman, M.J.

    1982-01-01

    The object of the present work was to characterise the natures and amounts of the various α and βγ activities associated with cladding hulls. The claddings studied were stainless steel from a Fast Reactor and from an Advanced Gas Reactor and Zircaloy from a Boiling Water Reactor, from a Pressurized Water Reactor and from a Steam Generating Heavy Water Reactor. The hulls were examined by the following methods: alpha spectrometry to identify and quantify the α emitters and to estimate their depths of penetration, partial and complete dissolution of hulls followed by gross α counting, α spectrometry and γ spectrometry, fission track autoradiography to determine the distribution of fissile material associated with hulls, neutron activation to determine the total fissile content of the hulls, chemical separations followed by β counting and chemical treatment with various reagents to examine the ease of decontamination

  2. Chemical interaction at the FBR cladding fuel interfaces

    International Nuclear Information System (INIS)

    Delbrassine, A.; Retels, J.; Dirven, P.

    1978-01-01

    Pins containing UO 2 -30 wt.%PuO 2 and/or Caesium and/or Telluriom as doping elements have been irradiated for about 40 days in the BR2 reactor. The effects of two Cs/Te ratios, namely 1.3 and 4 and a wide range of O/M ratios on the inner corrosion of the clad have been investigated. The influence of Tellurium on the attack of the cladding has been pointed out. It may be responsible for the Chromium NS Nickel depletion in the grain boundaries of the steel. It is necessary to measure the effective Ts/Te ratio associated with the local corrosion layers. This local Cs/Te ratio should be more useful than the initial mean Cs/Te ratio in a pin for understanding the corrosion phenomene. (author)

  3. Pulsed Magnetic Welding for Advanced Core and Cladding Steel

    International Nuclear Information System (INIS)

    Cao, Guoping; Yang, Yong

    2013-01-01

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

  4. Multiresponse Optimization of Laser Cladding Steel + VC Using Grey Relational Analysis in the Taguchi Method

    Science.gov (United States)

    Zhang, Zhe; Kovacevic, Radovan

    2016-07-01

    Laser cladding of metal matrix composite coatings (MMCs) has become an effective and economic method to improve the wear resistance of mechanical components. The clad quality characteristics such as clad height, carbide fraction, carbide dissolution, and matrix hardness in MMCs determine the wear resistance of the coatings. These clad quality characteristics are influenced greatly by the laser cladding processing parameters. In this study, American Iron and Steel Institute (AISI) 420 + 20% vanadium carbide (VC) was deposited on mild steel with a high powder direct diode laser. The Taguchi-based Grey relational method was used to optimize the laser cladding processing parameters (laser power, scanning speed, and powder feed rate) with the consideration of multiple clad characteristics related to wear resistance (clad height, carbide volume fraction, and Fe-matrix hardness). A Taguchi L9 orthogonal array was designed to study the effects of processing parameters on each response. The contribution and significance of each processing parameter on each clad characteristic were investigated by the analysis of variance (ANOVA). The Grey relational grade acquired from Grey relational analysis was used as the performance characteristic to obtain the optimal combination of processing parameters. Based on the optimal processing parameters, the phases and microstructure of the laser-cladded coating were characterized by using x-ray diffraction (XRD) and scanning electron microscopy (SEM) with energy-dispersive spectroscopy (EDS).

  5. Radiation hardness of new Kuraray double cladded optical fibers

    International Nuclear Information System (INIS)

    Bedeschi, F.; Menzione, A.; Budagov, Yu.; Chirikov-Zorin, I.; Solov'ev, A.; Turchanovich, L.; Vasil'chenko, V.

    1996-01-01

    The radiation hardness of the new plastic scintillating and clear fibers irradiated by 137 Cs γ-flux and by pulsed reactor fast neutrons were investigated. All the studied fibers were of S-type (with S=70) and had a double cladding. Optical fibers degradation study after irradiation shows that the level of radiation hardness lower that what is expected from results of previous studies. 9 refs., 6 figs

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

    International Nuclear Information System (INIS)

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

    1998-06-01

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

  7. In-situ crack repair by laser cladding

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2010-09-01

    Full Text Available Hammer peening was introduced in order to mechanically seal of squirting water prior to crack sealing. Leaks were reduced from squirting to oozing. During cladding, water started squirting out again resulting in pore formation in some areas. Hammer... is shown in Figure 6. Three small leaking pores were observed in the first layer, indicated by arrows. Hammer peening was applied to the first layer to mechanically seal the leaking pores prior to deposition of the second sealing layer. Successful...

  8. Technology for High Pure Aluminum Oxide Production from Aluminum Scrap

    Science.gov (United States)

    Ambaryan, G. N.; Vlaskin, M. S.; Shkolnikov, E. I.; Zhuk, A. Z.

    2017-10-01

    In this study a simple ecologically benign technology of high purity alumina production is presented. The synthesis process consists of three steps) oxidation of aluminum in water at temperature of 90 °C) calcinations of Al hydroxide in atmosphere at 1100 °C) high temperature vacuum processing of aluminum alpha oxide at 1750 °C. Oxidation of aluminum scrap was carried out under intensive mixing in water with small addition of KOH as a catalyst. It was shown that under implemented experimental conditions alkali was continuously regenerated during oxidation reaction and synergistic effect of low content alkali aqueous solution and intensive mixing worked. The product of oxidation of aluminum scrap is the powder of Al(OH)3. Then it can be preliminary granulated or directly subjected to thermal treatment deleting the impurities from the product (aluminum oxide). It was shown the possibility to produce the high-purity aluminum oxide of 5N grade (99.999 %). Aluminum oxide, synthesized by means of the proposed method, meets the requirements of industrial manufacturers of synthetic sapphire (aluminum oxide monocrystals). Obtained high pure aluminum oxide can be also used for the manufacture of implants, artificial joints, microscalpels, high-purity ceramics and other refractory shapes for manufacture of ultra-pure products.

  9. Aluminum Carbothermic Technology

    Energy Technology Data Exchange (ETDEWEB)

    Bruno, Marshall J.

    2005-03-31

    This report documents the non-proprietary research and development conducted on the Aluminum Carbothermic Technology (ACT) project from contract inception on July 01, 2000 to termination on December 31, 2004. The objectives of the program were to demonstrate the technical and economic feasibility of a new carbothermic process for producing commercial grade aluminum, designated as the ''Advanced Reactor Process'' (ARP). The scope of the program ranged from fundamental research through small scale laboratory experiments (65 kW power input) to larger scale test modules at up to 1600 kW power input. The tasks included work on four components of the process, Stages 1 and 2 of the reactor, vapor recovery and metal alloy decarbonization; development of computer models; and economic analyses of capital and operating costs. Justification for developing a new, carbothermic route to aluminum production is defined by the potential benefits in reduced energy, lower costs and more favorable environmental characteristics than the conventional Hall-Heroult process presently used by the industry. The estimated metrics for these advantages include energy rates at approximately 10 kWh/kg Al (versus over 13 kWh/kg Al for Hall-Heroult), capital costs as low as $1250 per MTY (versus 4,000 per MTY for Hall-Heroult), operating cost reductions of over 10%, and up to 37% reduction in CO2 emissions for fossil-fuel power plants. Realization of these benefits would be critical to sustaining the US aluminum industries position as a global leader in primary aluminum production. One very attractive incentive for ARP is its perceived ability to cost effectively produce metal over a range of smelter sizes, not feasible for Hall-Heroult plants which must be large, 240,000 TPY or more, to be economical. Lower capacity stand alone carbothermic smelters could be utilized to supply molten metal at fabrication facilities similar to the mini-mill concept employed by the steel industry

  10. Microbial biofilm growth on irradiated, spent nuclear fuel cladding

    Science.gov (United States)

    Bruhn, D. F.; Frank, S. M.; Roberto, F. F.; Pinhero, P. J.; Johnson, S. G.

    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 × 10 3 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.

  11. 21 CFR 73.1645 - Aluminum powder.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 1 2010-04-01 2010-04-01 false Aluminum powder. 73.1645 Section 73.1645 Food and... ADDITIVES EXEMPT FROM CERTIFICATION Drugs § 73.1645 Aluminum powder. (a) Identity. (1) The color additive aluminum powder shall be composed of finely divided particles of aluminum prepared from virgin aluminum. It...

  12. 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.

  13. Cladding of Cr- Modified NiAl Coating on 310 Stainless Steel Weld Cladding and Evaluation of its Wear Behavior.

    Directory of Open Access Journals (Sweden)

    S. Pourmohamadi

    2017-02-01

    Full Text Available In this study, a Cr-modified NiAl coating was fabricated by weld cladding technique using Gas- Tungsten Arc Welding (GTAW process on 310 steel. Chemical composition and microstructure of the coating was studied by X-Ray Diffraction (XRD, optical microscopy and scanning electron microscopy equipped with an Energy Dispersive Spectroscopy (EDS. The wear behavior of the coated steel was examined through pin-on-disc tests at ambient temperature and 400 °C. The results showed that the hardness of coated steel increased remarkably due to the formation of Cr-modified NiAl on the surface. Furthermore, the wear experiments showed that the presence of Cr-modified NiAl coating caused significant improvement in wear resistance of cladding 310 steel at both ambient temperature and 400 °C. These results were discussed based on the wear mechanism obtained from examination of the worn surfaces using SEM.

  14. Fundamental Studies on Aluminum Soaps

    Science.gov (United States)

    1944-06-01

    loosely bound lauric acid in aluminum dilaurate giving results accurate probably to 0.1 - 0.2# and reproducible to ubdtit ,𔃺.05^, The method...proceeds at the steady rate quoted above» Therefore the lauric acid is not hold in the form of solid solution which would give a constantly...both from lauric acid and aluminum dilaurate. It is extremely unlikely that aluminum trilaurate, AIL3, would rapidly yield dilaurate with dry acetone

  15. Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

    International Nuclear Information System (INIS)

    Boltax, A.; Biancheria, A.

    1977-01-01

    Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

  16. Tailoring nonlinearity and dispersion of photonic crystal fibers using hybrid cladding

    International Nuclear Information System (INIS)

    Zhao-lun, Liu; Lan-tian, Hou; Wei, Wang

    2009-01-01

    We present a hybrid cladding photonic crystal fiber for shaping high nonlinear and flattened dispersion in a wide range of wavelengths. The new structure adopts hybrid cladding with different pitches, air-holes diameters and air-holes arrayed fashions. The full-vector finite element method with perfectly matched layer is used to investigate the characteristics of the hybrid cladding photonic crystal fiber such as nonlinearity and dispersion properties. The influence of the cladding structure parameters on the nonlinear coefficient and geometric dispersion is analyzed. High nonlinear coefficient and the dispersion properties of fibers are tailored by adjusting the cladding structure parameters. A novel hybrid cladding photonic crystal fiber with high nonlinear coefficient and dispersion flattened which is suited for super continuum generation is designed. (author)

  17. On LMFBR corrosion. Part II: Consideration of the in-reactor fuel-cladding system

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Walker, C.T.; Whitlow, W.H.

    1976-05-01

    The scientific and technological aspects of LMFBR cladding corrosion are discussed in detail. Emphasis is placed on the influence of the irradiation environment and the effect of fuel and filler-gas impurities on the corrosion process. These studies are complemented by a concise review of out-of-pile simulation experiments that endeavour to clarify the role of the aggressive fission products cesium, tellurium and iodine. The principal models for cladding corrosion are presented and critically assessed. Areas of uncertainty are exposed and some pertinent experiments are suggested. Consideration is also given to some new observations regarding the role of stress in fuel-cladding reactions and the formation of ferrite in the corrosion zone of the cladding during irradiation. Finally, two technological solutions to the problem of cladding corrosion are proposed. These are based on the use of an oxygen buffer in the fuel and the application of a protective coating to the inner surface of the cladding

  18. Cladding failure margins for metallic fuel in the integral fast reactor

    International Nuclear Information System (INIS)

    Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

    1987-01-01

    The reference fuel for Integral Fast Reactor (IFR) is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel

  19. Springback of aluminum alloy brazing sheet in warm forming

    Science.gov (United States)

    Han, Kyu Bin; George, Ryan; Kurukuri, Srihari; Worswick, Michael J.; Winkler, Sooky

    2017-10-01

    The use of aluminum is increasing in the automotive industry due to its high strength-to-weight ratio, recyclability and corrosion resistance. However, aluminum is prone to significant springback due to its low elastic modulus coupled with its high strength. In this paper, a warm forming process is studied to improve the springback characteristics of 0.2 mm thick brazing sheet with an AA3003 core and AA4045 clad. Warm forming decreases springback by lowering the flow stress. The parts formed have complex features and geometries that are representative of automotive heat exchangers. The key objective is to utilize warm forming to control the springback to improve the part flatness which enables the use of harder temper material with improved strength. The experiments are performed by using heated dies at several different temperatures up to 350 °C and the blanks are pre-heated in the dies. The measured springback showed a reduction in curvature and improved flatness after forming at higher temperatures, particularly for the harder temper material conditions.

  20. Thermocurrent dosimetry with high purity aluminum oxide

    International Nuclear Information System (INIS)

    Fullerton, G.D.; Cameron, J.R.; Moran, P.R.

    1976-01-01

    The application of thermocurrent (TC) to ionizing radiation dosimetry was studied. It was shown that TC in alumina (Al 2 O 3 ) has properties that are suited to personnel dosimetry and environmental monitoring. TC dosimeters were made from thin disks of alumina. Aluminum electrodes were evaporated on each side: on one face a high voltage electrode and on the opposite face a measuring electrode encircled by a guard ring. Exposure to ionizing radiation resulted in stored electrons and holes in metastable trapping sites. The signal was read-out by heating the dosimeter with a voltage source and picnometer connected in series between the opposite electrodes. The thermally remobilized charge caused a transient TC. The thermogram, TC versus time or temperature, is similar to a TL glow curve. Either the peak current or the integrated current is a measure of absorbed dose. Six grades of alumina were studied from a total of four commercial suppliers. All six materials displayed radiation induced TC signals. Sapphire of uv-grade quality from the Adolf Meller Co. (AM) had the best dosimetry properties of those investigated. Sources of interference were studied. Thermal fading, residual signal and radiation damage do not limit TC dosimetry. Ultraviolet light can induce a TC response but it is readily excluded with uv-opaque cladding. Improper surface preparation prior to electrode evaporation was shown to cause interference. A spurious TC signal resulted from polarization of surface contaminants. Spurious TC was reduced by improved cleaning prior to electrode application. Polished surfaces resulted in blocking electrodes and caused a sensitivity shift due to radiation induced thermally activated polarization. This was not observed with rough cut surfaces

  1. Reflood behavior at low initial clad temperature in Slab Core Test Facility Core-II

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Sobajima, Makoto; Abe, Yutaka; Iwamura, Takamichi; Ohnuki, Akira; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Adachi, Hiromichi.

    1990-07-01

    In order to study the reflood behavior with low initial clad temperature, a reflood test was performed using the Slab Core Test Facility (SCTF) with initial clad temperature of 573 K. The test conditions of the test are identical with those of SCTF base case test S2-SH1 (initial clad temperature 1073 K) except the initial clad temperature. Through the comparison of results from these two tests, the following conclusions were obtained. (1) The low initial clad temperature resulted in the low differential pressures through the primary loops due to smaller steam generation in the core. (2) The low initial clad temperature caused the accumulated mass in the core to be increased and the accumulated mass in the downcomer to be decreased in the period of the lower plenum injection with accumulator (before 50s). In the later period of the cold leg injection with LPCI (after 100s), the water accumulation rates in the core and the downcomer were almost the same between both tests. (3) The low initial clad temperature resulted in the increase of the core inlet mass flow rate in the lower plenum injection period. However, the core inlet mass flow rate was almost the same regardless of the initial clad temperature in the later period of the cold leg injection period. (4) The low initial clad temperature resulted in the low turnaround temperature, high temperature rise and fast bottom quench front propagation. (5) In the region apart from the quench front, low initial clad temperature resulted in the lower heat transfer. In the region near the quench front, almost the same heat transfer coefficient was observed between both tests. (6) No flow oscillation with a long period was observed in the SCTF test with low initial clad temperature of 573 K, while it was remarkable in the Cylindrical Core Test Facility (CCTF) test which was performed with the same initial clad temperature. (J.P.N.)

  2. Solid-Core Photonic Bandgap Fibers for Cladding-Pumped Raman Amplification

    Science.gov (United States)

    2011-06-03

    that the cladding is uniformly-pumped at its maximum numerical aperture and that the photonic band gap ( PBG ) structure comprised of a triangular array...published 3 Jun 2011 (C) 2011 OSA 6 June 2011 / Vol. 19, No. 12 / OPTICS EXPRESS 11855 beyond the PBG structure. In this case, the high pump cladding...exhibit lower loss [16]. For high order bandgaps to exist in the cladding, the inclusions forming the PBG structure must support a large number of guided

  3. Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor (SFR)

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock

    2013-01-01

    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

  4. The significance of cladding material on the integrity of nuclear pressure vessels with cracks

    International Nuclear Information System (INIS)

    Sattari-Far, Iradj.

    1989-05-01

    The significance of the austenitic cladding layer is reviewed in this literature study. The cladding induced stresses are generally not considered when evaluating the severity of flaws in reactor pressure vessels. It has been shown that this emission may be misleading. The necessity to consider the cladding induced stresses is also emphasized in the latest edition of ASME XI. Contrary to what is commonly assumed, the austenitic cladding displays a charpy V transition region with a low ductility. The interface material (HAZ) is the most influenced region by irradiation, and a transition shift of over 100 degree C may be expected. Because of the significant difference in the thermal expansion coefficients of the cladding and the base metal, cladding induced stresses can be set up. Even after PWHT, residual stresses of yield magnitude remain in the cladding and the HAZ at ambient temperature. The cladding induced stresses are temperature dependent and decrease as the temperature increases. The cladding induced stresses have a significant influence on small defects near the inside surface of a pressure vessel. For semielliptical surface cracks, the maximum CTOD-value along the crack front is not found at the deepest point, but in the cladding/base metal interface, having a magnitude three times higher than the value in the deepest point. It implies that this type of crack would propagate along the clad/base material interface. At some point in time, the crack will reach a geometry which may cause such a severe condition at the deepest point that it will start to grow in the depth direction as well. The initiation and growth behaviour of such cracks need to be investigated to be able to assess the significance of cladding on the integrity of nuclear pressure vessels. (author) (50 figs., 33 refs.)

  5. 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...

  6. Production of aluminum metal by electrolysis of aluminum sulfide

    Science.gov (United States)

    Minh, N.Q.; Loutfy, R.O.; Yao, N.P.

    1982-04-01

    Metallic aluminum may be produced by the electrolysis of Al/sub 2/S/sub 3/ at 700 to 800/sup 0/C in a chloride melt composed of one or more alkali metal chlorides, and one or more alkaline earth metal chlorides and/or aluminum chloride to provide improved operating characteristics of the process.

  7. Dynamic Property of Aluminum Foam

    Directory of Open Access Journals (Sweden)

    S Irie

    2016-09-01

    Full Text Available Aluminum in the foam of metallic foam is in the early stage of industrialization. It has various beneficial characteristics such as being lightweight, heat resistance, and an electromagnetic radiation shield. Therefore, the use of aluminum foam is expected to reduce the weight of equipment for transportation such as the car, trains, and aircraft. The use as energy absorption material is examined. Moreover aluminum foam can absorb the shock wave, and decrease the shock of the blast. Many researchers have reported about aluminum foam, but only a little information is available for high strain rates (103 s-1 or more. Therefore, the aluminum foam at high strain rates hasn't been not characterized yet. The purpose in this research is to evaluate the behavior of the aluminum form in the high-strain rate. In this paper, the collision test on high strain rate of the aluminum foam is investigated. After experiment, the numerical analysis model will be made. In this experiment, a powder gun was used to generate the high strain rate in aluminum foam. In-situ PVDF gauges were used for measuring pressure and the length of effectiveness that acts on the aluminum foam. The aluminum foam was accelerated to about 400 m/s from deflagration of single component powder and the foam were made to collide with the PVDF gauge. The high strain rate deformation of the aluminum form was measured at two collision speeds. As for the result, pressure was observed to go up rapidly when about 70% was compressed. From this result, it is understood that complete crush of the cell is caused when the relative volume is about 70%. In the next stage, this data will be compared with the numerical analysis.

  8. 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.

  9. Simulation of a pellet-clad mechanical interaction with ABAQUS and its verification

    International Nuclear Information System (INIS)

    Cheon, J.-S.; Lee, B.-H.; Koo, Y.-H.; Sohn, D.-S.; Oh, J.-Y.

    2003-01-01

    Pellet-clad mechanical interaction (PCMI) during power transients for MOX fuel is modelled by a FE method. The PCMI model predicts well clad elongation during power ramp and relaxation during power hold except the fuel behaviour during a power decrease. Higher fiction factor results in the earlier occurrence of PCMI and more enhanced clad elongation. The relaxation is dependent on the irradiation creep rate of the pellet and axial compressive force. Verification of the PCMI model was done using recent MOX experimental data. Temperature and clad elongation for the fuel rod can be evaluated in a reasonable way

  10. Simulation of high burn-up fuel cladding and its safety assessment under LOCA condition

    International Nuclear Information System (INIS)

    Park, Dong Jun; Won, Sung Bin; Choi, Byoung Kwon; Park, Jeong Yong; Koo, Yang Hyun

    2011-01-01

    Current LOCA safety criteria was established in the beginning of 1970s and based on the results obtained from non-irradiated Zircaloy-4 claddings. Because of major advantages in fuel-cycle costs, reactor operation, and waste management, the increase in fuel discharge burn-up is current worldwide trend in the nuclear industry. As the fuel burn-up increases, various phenomena unexpected have been reported due to changes in the condition of reactor operation and in-core environment. Since, it should be considered whether the current Loss-of-coolant accident (LOCA) criteria is suitable for high burn-up fuel cladding or not. In addition, many fuel vendors have recently developed new cladding alloys superior to Zircaloy-4 cladding. The performance of these advanced cladding alloys under LOCA, especially at high burn-up, is not well understood at this time. To better understand high burn-up effects and commercialize new cladding alloys, study of LOCA-related behavior of various types of high burn-up fuel cladding and their data base is essentially required. In this background, postulated LOCA test has been carried out with prehydrided Zircaloy-4 cladding as a surrogate for high burn-up cladding and the relevant results obtained are discussed

  11. Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design

    International Nuclear Information System (INIS)

    Roake, W.E.

    1977-01-01

    Fuel-cladding-chemical-interaction (FCCI) is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals

  12. Formation quality optimization of laser hot wire cladding for repairing martensite precipitation hardening stainless steel

    Science.gov (United States)

    Wen, Peng; Feng, Zhenhua; Zheng, Shiqing

    2015-01-01

    Laser cladding is an advantaged repairing technology due to its low heat input and high flexibility. With preheating wire by resistance heat, laser hot wire cladding shows better process stability and higher deposition efficiency compared to laser cold wire/powder cladding. Multi-pass layer were cladded on the surface of martensite precipitation hardening stainless steel FV520B by fiber laser with ER410NiMo wire. Wire feed rate and preheat current were optimized to obtain stable wire transfer, which guaranteed good formation quality of single pass cladding. Response surface methodology (RSM) was used to optimize processing parameters and predict formation quality of multi-pass cladding. Laser power P, scanning speed Vs, wire feed rate Vf and overlap ratio η were selected as the input variables, while flatness ratio, dilution and incomplete fusion value as the responses. Optimal clad layer with flat surface, low dilution and no incomplete fusion was obtained by appropriately reducing Vf, and increasing P, Vs and η. No defect like pore or crack was found. The tensile strength and impact toughness of the clad layer is respectively 96% and 86% of those of the substrate. The clad layer showed nonuniform microstructure and was divided into quenched areas with coarse lath martensite and tempered areas with tempered martensite due to different thermal cycles in adjacent areas. The tempered areas showed similar hardness to the substrate.

  13. Raman probes based on optically-poled double-clad fiber and coupler

    DEFF Research Database (Denmark)

    Brunetti, Anna Chiara; Margulis, Walter; Rottwitt, Karsten

    2012-01-01

    Two fiber Raman probes are presented, one based on an optically-poled double-clad fiber and the second based on an optically-poled double-clad fiber coupler respectively. Optical poling of the core of the fiber allows for the generation of enough 532nm light to perform Raman spectroscopy...... of a sample of dimethyl sulfoxide (DMSO), when illuminating the waveguide with 1064nm laser light. The Raman signal is collected in the inner cladding, from which it is retrieved with either a bulk dichroic mirror or a double-clad fiber coupler. The coupler allows for a substantial reduction of the fiber...

  14. Corrosion Resistance of Laser Clads of Inconel 625 and Metco 41C

    Science.gov (United States)

    Němeček, Stanislav; Fidler, Lukáš; Fišerová, Pavla

    The present paper explores the impact of laser cladding parameters on the corrosion behaviour of the resulting surface. Powders of Inconel 625 and austenitic Metco 41C steel were deposited on steel substrate. It was confirmed that the level of dilution has profound impact on the corrosion resistance and that dilution has to be minimized. However, the chemical composition of the cladding is altered even in the course of the cladding process, a fact which is related to the increase in the substrate temperature. The cladding process was optimized to achieve maximum corrosion resistance. The results were verified and validated using microscopic observation, chemical analysis and corrosion testing.

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

    Science.gov (United States)

    Saltsman, J. F.

    1973-01-01

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

  16. Results of U-xMo (x=7, 10, 12 wt.%) Alloy versus Al-6061 Cladding Diffusion Couple Experiments Performed at 500, 550 and 600 Degrees C

    Energy Technology Data Exchange (ETDEWEB)

    Emmanuel Perez; Dennis D. Keiser, Jr.; Yongho Sohn

    2013-04-01

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has been developing low enrichment fuel systems encased in Al 6061 for use in research and test reactors. U–Mo alloys in contact with Al and Al alloys can undergo diffusional interactions that can result in the development of interdiffusion zones with complex fine-grained microstructures composed of multiple phases. A monolithic fuel currently being developed by the RERTR program has local regions where the U–Mo fuel plate is in contact with the Al 6061 cladding and, as a result, the program finds information about interdiffusion zone development at high temperatures of interest. In this study, the microstructural development of diffusion couples consisting of U-7wt.%Mo, U-10wt.%Mo, and U-12wt.%Mo vs. Al 6061 (or 6061 aluminum) cladding, annealed at 500, 550, 600 degrees C for 1, 5, 20, 24, or 132 hours, was analyzed by backscatter electron microscopy and x-ray energy dispersive spectroscopy on a scanning electron microscope. Concentration profiles were determined by standardized wavelength dispersive spectroscopy and standardless x-ray energy dispersive spectroscopy. The results of this work shows that the presence of surface layers at the U–Mo/Al 6061 interface can dramatically impact the overall interdiffusion behavior in terms of rate of interaction and uniformity of the developed interdiffusion zones. It further reveals that relatively uniform interaction layers with higher Si concentrations can develop in U–Mo/Al 6061 couples annealed at shorter times and that longer times at temperature result in the development of more non-uniform interaction layers with more areas that are enriched in Al. At longer annealing times and relatively high temperatures, U–Mo/Al 6061 couples can exhibit more interaction compared to U–Mo/pure Al couples. The minor alloying constituents in Al 6061 cladding can result in the development of many complex phases in the interaction layer of U

  17. Experiments to understand the corrosion process of fuel rod claddings

    International Nuclear Information System (INIS)

    Groeschel, F.; Hermann, A.

    1997-01-01

    Fuel rods in light water reactors have to respond to the trends in increased burn-up and extended dwelling time in reactor. Waterside corrosion of the cladding affecting wall thickness, mechanical stability due to hydriding and the heat transfer due to the low thermal conductivity of the oxide scale may become the limiting factors. The corrosion process is complex and involves a large variety of mechanisms. Understanding of the process is important for safe operation and a prerequisite for development of improved materials. A variety of analytical techniques and mechanical tests, including examination of irradiated pathfinder rods, are used to tackle the different aspects. (author) 6 figs., 1 tab., 17 refs

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  19. Standard specification for architectural flat glass clad polycarbonate

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This specification covers the quality requirements for cut sizes of glass clad polycarbonate (GCP) for use in buildings as security, detention, hurricane/cyclic wind-resistant, and blast and ballistic-resistant glazing applications. 1.2 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  20. Embrittlement of zircaloy cladding due to oxygen uptake (CBRTTL)

    International Nuclear Information System (INIS)

    Reymann, G.A.

    1979-02-01

    A model for embrittlement of zircaloy due to oxygen uptake at high temperatures is described. The model defines limits for oxygen content and temperature which, if exceeded, give rise to zircaloy cladding which is sufficiently embrittled to cause failure either on quenching or normal handling following a transient. A significant feature of this model is that the onset of embrittlement is dependent on the cooling rate. A distinction is made between slow and fast cooling, with the boundary at 100 K/s. The material property correlations and computer subcodes described in MATPRO are developed for use in Light Water Reactor (LWR) codes

  1. Ultrasonic signal processing for sizing under-clad flaws

    International Nuclear Information System (INIS)

    Shankar, R.; Paradiso, T.J.; Lane, S.S.; Quinn, J.R.

    1985-01-01

    Ultrasonic digital data were collected from underclad cracks in sample pressure vessel specimen blocks. These blocks were weld cladded under different processes to simulate actual conditions in US Pressure Water Reactors. Each crack was represented by a flaw-echo dynamic curve which is a plot of the transducer motion on the surface as a function of the ultrasonic response into the material. Crack depth sizing was performed by identifying in the dynamic curve the crack tip diffraction signals from the upper and lower tips. This paper describes the experimental procedure, digital signal processing methods used and algorithms developed for crack depth sizing

  2. Widely tunable femtosecond solitonic radiation in photonic crystal fiber cladding

    DEFF Research Database (Denmark)

    Peng, J. H.; Sokolov, A. V.; Benabid, F.

    2010-01-01

    We report on a means to generate tunable ultrashort optical pulses. We demonstrate that dispersive waves generated by solitons within the small-core features of a photonic crystal fiber cladding can be used to obtain femtosecond pulses tunable over an octave-wide spectral range. The generation...... process is highly efficient and occurs at the relatively low laser powers available from a simple Ti:sapphire laser oscillator. The described phenomenon is general and will play an important role in other systems where solitons are known to exist....

  3. Laser cladding of wear resistant metal matrix composite coatings

    International Nuclear Information System (INIS)

    Yakovlev, A.; Bertrand, Ph.; Smurov, I.

    2004-01-01

    A number of coatings with wear-resistant properties as well as with a low friction coefficient are produced by laser cladding. The structure of these coatings is determined by required performance and realized as metal matrix composite (MMC), where solid lubricant serves as a ductile matrix (e.g. CuSn), reinforced by appropriate ceramic phase (e.g. WC/Co). One of the engineered coating with functionally graded material (FGM) structure has a dry friction coefficient 0.12. Coatings were produced by coaxial injection of powder blend into the zone of laser beam action. Metallographic and tribological examinations were carried out confirming the advanced performance of engineered coatings

  4. Properties of light water reactor spent fuel cladding. Interim report

    International Nuclear Information System (INIS)

    Farwick, D.G.; Moen, R.A.

    1979-08-01

    The Commercial Waste and Spent Fuel Packaging Program will provide containment packages for the safe storage or disposal of spent Light Water Reactor (LWR) fuel. Maintaining containment of radionuclides during transportation, handling, processing and storage is essential, so the best understanding of the properties of the materials to be stored is necessary. This report provides data collection, assessment and recommendations for spent LWR fuel cladding materials properties. Major emphasis is placed on mechanical properties of the zircaloys and austenitic stainless steels. Limited information on elastic constants, physical properties, and anticipated corrosion behavior is also provided. Work is in progress to revise these evaluations as the program proceeds

  5. Aluminum anode for aluminum-air battery - Part I: Influence of aluminum purity

    Science.gov (United States)

    Cho, Young-Joo; Park, In-Jun; Lee, Hyeok-Jae; Kim, Jung-Gu

    2015-03-01

    2N5 commercial grade aluminum (99.5% purity) leads to the lower aluminum-air battery performances than 4N high pure grade aluminum (99.99% purity) due to impurities itself and formed impurity complex layer which contained Fe, Si, Cu and others. The impurity complex layer of 2N5 grade Al declines the battery voltage on standby status. It also depletes discharge current and battery efficiency at 1.0 V which is general operating voltage of aluminum-air battery. However, the impurity complex layer of 2N5 grade Al is dissolved with decreasing discharge voltage to 0.8 V. This phenomenon leads to improvement of discharge current density and battery efficiency by reducing self-corrosion reaction. This study demonstrates the possibility of use of 2N5 grade Al which is cheaper than 4N grade Al as the anode for aluminum-air battery.

  6. Aluminum plasmonic photocatalysis

    Science.gov (United States)

    Hao, Qi; Wang, Chenxi; Huang, Hao; Li, Wan; Du, Deyang; Han, Di; Qiu, Teng; Chu, Paul K.

    2015-01-01

    The effectiveness of photocatalytic processes is dictated largely by plasmonic materials with the capability to enhance light absorption as well as the energy conversion efficiency. Herein, we demonstrate how to improve the plasmonic photocatalytic properties of TiO2/Al nano-void arrays by overlapping the localized surface plasmon resonance (LSPR) modes with the TiO2 band gap. The plasmonic TiO2/Al arrays exhibit superior photocatalytic activity boasting an enhancement of 7.2 folds. The underlying mechanisms concerning the radiative energy transfer and interface energy transfer processes are discussed. Both processes occur at the TiO2/Al interface and their contributions to photocatalysis are evaluated. The results are important to the optimization of aluminum plasmonic materials in photocatalytic applications. PMID:26497411

  7. Determination of Sandoz Black Aluminum Coloring Dye Olive Aluminum Coloring Dye and Sodium Dichromate Aluminum Sealing Solutions by UV-Visible Spectrophotometry

    National Research Council Canada - National Science Library

    Sopok, Samuel

    1992-01-01

    The chemical literature lacks an acceptable method to determine and adequately control Sandoz black aluminum coloring dye, olive aluminum coloring dye, and sodium dichromate aluminum sealing solutions...

  8. Spray Rolling Aluminum Strip

    Energy Technology Data Exchange (ETDEWEB)

    Lavernia, E.J.; Delplanque, J-P; McHugh, K.M.

    2006-05-10

    Spray forming is a competitive low-cost alternative to ingot metallurgy for manufacturing ferrous and non-ferrous alloy shapes. It produces materials with a reduced number of processing steps, while maintaining materials properties, with the possibility of near-net-shape manufacturing. However, there are several hurdles to large-scale commercial adoption of spray forming: 1) ensuring strip is consistently flat, 2) eliminating porosity, particularly at the deposit/substrate interface, and 3) improving material yield. Through this program, a new strip/sheet casting process, termed spray rolling, has been developed, which is an innovative manufacturing technique to produce aluminum net-shape products. Spray rolling combines the benefits of twin-roll casting and conventional spray forming, showing a promising potential to overcome the above hurdles associated with spray forming. Spray rolling requires less energy and generates less scrap than conventional processes and, consequently, enables the development of materials with lower environmental impacts in both processing and final products. Spray Rolling was developed as a collaborative project between the University of California-Davis, the Colorado School of Mines, the Idaho National Engineering and Environmental Laboratory, and an industry team. The following objectives of this project were achieved: (1) Demonstration of the feasibility of the spray rolling process at the bench-scale level and evaluation of the materials properties of spray rolled aluminum strip alloys; and (2) Demonstration of 2X scalability of the process and documentation of technical hurdles to further scale up and initiate technology transfer to industry for eventual commercialization of the process.

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

    International Nuclear Information System (INIS)

    Park, Jung-Hwan; Kim, Eui-Jung; Jung, Yang-Il; Park, Dong-Jun; Kim, Hyun-Gil; Park, Jeong-Yong; Koo, Yang-Hyun

    2015-01-01

    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

  10. Conical reflection of light during free-space coupling into a symmetrical metal-cladding waveguide.

    Science.gov (United States)

    Zheng, Yuanlin; Cao, Zhuangqi; Chen, Xianfeng

    2013-09-01

    Novel conical reflection of light by a thick three-layered metal-clad optical waveguide is observed. A symmetrical metal-cladding optical waveguide is used, which exhibits extraordinary conical reflection during free-space coupling of light to the waveguide. The phenomenon is attributed to the leakage of excited ultrahigh-order guided modes and their inter- and intramode coupling interaction.

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

    International Nuclear Information System (INIS)

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

    1984-03-01

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

  12. Core temperature in super-Gaussian pumped air-clad photonic ...

    Indian Academy of Sciences (India)

    In this paper we investigate the core temperature of air-clad photonic crystal fiber (PCF) lasers pumped by a super-Gaussian (SG) source of order four. The results are compared with conventional double-clad fiber (DCF) lasers pumped by the same super-Gaussian and by top-hat pump profiles.

  13. Technique Comparison of the Fracture Toughness Tests for Irradiated Fuel Claddings in a Hot Cell

    International Nuclear Information System (INIS)

    Ahn, Sangbok; Kim, Dosik; Jung, Yanghong; Choo, Yongsun; Ryu, Wooseog

    2007-01-01

    The degradation of a fracture toughness in a fuel cladding is a important factor to restrict the operation safety in nuclear power plants. The fracture properties of claddings were traditionally measured through a rubber bung test, a burst test, etc. Those results were the qualitative fracture characteristics, and could not be used as design or operation safety evaluation data. We need to evaluate the quantitative characteristics of claddings under normal operation and in accidents. The application of a fracture mechanics concept in testing a fuel cladding is restricted by the cladding geometry and creating the correct stress-state conditions. The geometry of claddings does not meet the requirement of the ASTM Standards for a specimen configuration and an applied load. The specimen may be produced from previously flattened claddings, but the flattening causes some uncertainties in the results due to changes in the microstructure of the material and a new distribution of the internal stresses. Therefore many efforts have been devoted to developing new test techniques, to quantify the fracture characteristics of claddings. Researchers from JAEA and NFI in Japan, Studsvik Company Ltd in Sweden, IAEA in Australia, and KAERI in Korea have independently developed fracture test techniques. This study is designed to review the independently developed techniques and to compare of their merits. Finally we shall apply the other techniques to upgrade our developing techniques

  14. Development of an observation and control system for industrial laser cladding

    NARCIS (Netherlands)

    Hofman, Johannes Tjaard

    2009-01-01

    Laser cladding has become an important surface modification technique in today’s industry. It is not only applied for coating new products but also for repair and refurbishment as well as in rapid prototyping. A laser clad workstation has been developed. It uses a 4 kW Nd:YAG fibre coupled laser as

  15. Experimental verification of microbending theory using mode coupling to discrete cladding modes

    DEFF Research Database (Denmark)

    Probst, C. B.; Bjarklev, Anders Overgaard; Andreasen, S. B.

    1989-01-01

    a microbending theory in which coupling between the guided mode and a number of discrete cladding modes is considered. Very good agreement between theory and measurement is achieved. The consequences of the existence of discrete cladding modes with regard to the proper choice of artificial microbending spectrum...

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

    Energy Technology Data Exchange (ETDEWEB)

    Dalcher, A.W.

    1969-04-15

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

  17. Air-clad fibers: pump absorption assisted by chaotic wave dynamics?

    DEFF Research Database (Denmark)

    Mortensen, Niels Asger

    2007-01-01

    Wave chaos is a concept which has already proved its practical usefulness in design of double-clad fibers for cladding-pumped fiber lasers and fiber amplifiers. In general, classically chaotic geometries will favor strong pump absorption and we address the extent of chaotic wave dynamics in typical...

  18. Method of evaluation of stress corrosion cracking susceptibility of clad fuel tubes

    International Nuclear Information System (INIS)

    Takase, Iwao; Yoshida, Toshimi; Ikeda, Shinzo; Masaoka, Isao; Nakajima, Junjiro.

    1986-01-01

    Purpose: To determine, by an evaluation in out-pile test, the stress corrosion cracking susceptibility of clad fuel tubes in the reactor environment. Method: A plurality of electrodes are mounted in the circumferential direction on the entire surface of cladding tubes. Of the electrodes, electrodes at two adjacent places are used as measuring terminals and electrodes at another two places adjacent thereto are used as constant-current terminals. With a specific current flowing in the constant-current terminals, measurements are made of a potential difference between the terminals to be measured, and from a variation in the potential difference the depth of cracking of the cladding tube surface is presumed to determine the stress corrosion cracking susceptibility of the cladding tube. To check the entire surface of the cladding tube, the cladding tube is moved by each block in the circumferential direction by a contact changeover system, repeating the measurements of the potential difference. Contact type electrodes are secured with an insulator and held in uniform contact with the cladding tube by a spring. It is detachable by use of a locking system and movable as desired. Thus the stress corrosion cracking susceptibility can be determined without mounting the cladding tube through and also a fuel failure can be prevented. (Horiuchi, T.)

  19. An Examination of Collaborative Learning Assessment through Dialogue (CLAD) in Traditional and Hybrid Human Development Courses

    Science.gov (United States)

    McCarthy, Wanda C.; Green, Peter J.; Fitch, Trey

    2010-01-01

    This investigation assessed the effectiveness of using Collaborative Learning Assessment through Dialogue (CLAD) (Fitch & Hulgin, 2007) with students in undergraduate human development courses. The key parts of CLAD are student collaboration, active learning, and altering the role of the instructor to a guide who enhances learning opportunities.…

  20. 78 FR 10265 - Pricing for the 2013 Commemorative Coin Programs-Silver and Clad Coin Options

    Science.gov (United States)

    2013-02-13

    ... DEPARTMENT OF THE TREASURY United States Mint Pricing for the 2013 Commemorative Coin Programs--Silver and Clad Coin Options AGENCY: United States Mint, Department of the Treasury. ACTION: Notice... Dollar and the 2013 5-Star Generals Commemorative Coin Program for the silver and clad coin options...

  1. Elastic plastic analysis of fuel element assemblies - hexagonal claddings and fuel rods

    International Nuclear Information System (INIS)

    Mamoun, M.M.; Wu, T.S.; Chopra, P.S.; Rardin, D.C.

    1979-01-01

    Analytical studies have been conducted to investigate the structural, thermal, and mechanical behavior of fuel rods, claddings and fuel element assemblies of several designs for a conceptual Safety Test Facility (STF). One of the design objectives was to seek a geometrical configuration for a clad by maximizing the volume fraction of fuel and minimizing the resultant stresses set-up in the clad. The results of studies conducted on various geometrical configurations showed that the latter design objective can be achieved by selecting a clad of an hexagonal geometry. The analytical studies necessitated developing solutions for determining the stresses, strains, and displacements experienced by fuel rods and an hexagonal cladding subjected to thermal fuel-bowing loads acting on its internal surface, the external pressure of the coolant, and elevated temperatures. This paper presents some of the initially formulated analytical methods and results. It should be emphasized that the geometrical configuration considered in this paper may not necessarily be similar to that of the final design. Several variables have been taken into consideration including cladding thickness, the dimensions of the fuel rod, the temperature of the fuel and cladding, the external pressure of the cooling fluid, and the mechanical strength properties of fuel and cladding. A finite-element computer program, STRAW Code, has also been employed to generate several numerical results which have been compared with those predicted by employing the initially formulated solutions. The theoretically predicted results are in good agreement with those of the STRAW Code. (orig.)

  2. 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)

  3. Effects of Zr-hydride distribution of irradiated Zircaloy-2 cladding in RIA-simulating pellet-clad mechanical interaction testing

    Directory of Open Access Journals (Sweden)

    Per Magnusson

    2018-03-01

    Full Text Available A series of simulated reactivity-initiated accident (RIA tests on irradiated fully recrystallized boiling water reactor Zircaloy-2 cladding has been performed by means of the expansion-due-to-compression (EDC test method. The EDC method reproduces fuel pellet–clad mechanical interaction (PCMI conditions for the cladding during RIA transients with respect to temperature and loading rates by out-of-pile mechanical testing. The tested materials had a large variation in burnup and hydrogen content (up to 907 wppm. The results of the EDC tests showed variation in the PCMI resistance of claddings with similar burnup and hydrogen content, making it difficult to clearly identify ductile-to-brittle transition temperatures. The EDC-tested samples of the present and previous work were investigated by light optical and scanning electron microscopy to study the influence of factors such as azimuthal variation of the Zr-hydrides and the presence of hydride rims and radially oriented hydrides. Two main characteristics were identified in samples with low ductility with respect to hydrogen content and test temperature: hydride rims and radial hydrides at the cladding outer surface. Crack propagation and failure modes were also studied, showing two general modes of crack propagation depending on distribution and amount of radially oriented hydrides. It was concluded that the PCMI resistance of irradiated cladding under normal conditions with homogenously distributed circumferential hydrides is high, with good margin to the RIA failure limits. To further improve safety, focus should be on conditions causing nonfavorable hydride distribution, such as hydride reorientation and formation of hydride blisters at the cladding outer surface. Keywords: Failure, Hydrides, Hydrogen Content, Pellet–Clad Mechanical Interaction, Reactivity-Initiated Accident, Transient

  4. Development of eutectic free cladding materials for metallic fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-11-01

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

  5. Treatment of cladding hulls by the HIPOW process

    International Nuclear Information System (INIS)

    Larker, H.T.; Tegman, R.

    1981-01-01

    The conditions for densifying and bonding Zircaloy cladding hulls from spent LWR fuel to blocks by the HIPOW (hot isostatic pressing of waste) process have been studied. Fully dense and mechanically strong blocks of Zircaloy can be made without additives at temperatures around 1000 0 C. A volume reduction of about seven times and surface area reduction of more than 300 times, compared to typical loose-filled cladding hulls remaining after the chop-leach operations in a reprocessing plant, can be obtained. A study of a possible process for industrial scale has been made. Handling under water can prevent any fire hazard in the preparation sequence. The use of a special hermetically sealed double-wall metal container encasing the hulls during the densification in the hot isostatic press virtually eliminates the problem of lasting contamination of this equipment, thus greatly simplifying service and maintenance. One hot isostatic press can serve a reprocessing line with an LWR fuel capacity of 800 tons/year. Fines (residues) from fuel dissolution and alpha-contaminated ashes from incinerated organic materials in the plant may also be incorporated in the Zircaloy blocks. Tritium can quantitatively be contained in these blocks

  6. Cladding and Structural Materials for Advanced Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-06-30

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

  7. Reactor water chemistry relevant to coolant-cladding interaction

    International Nuclear Information System (INIS)

    1987-09-01

    The report is a summary of the work performed in a frame of a Coordinated Research Program organized by the IAEA and carried out from 1981 till 1986. It consists of a survey on our knowledge on coolant-cladding interaction: the basic phenomena, the relevant parameters, their control and the modelling techniques implemented for their assessment. Based upon the results of this Coordinated Research Program, the following topics are reviewed on the report: role of water chemistry in reliable operation of nuclear power plants; water chemistry specifications and their control; behaviour of fuel cladding materials; corrosion product behaviour and crud build-up in reactor circuits; modelling of corrosion product behaviour. This report should be of interest to water chemistry supervisors at the power plants, to experts in utility engineering departments, to fuel designers, to R and D institutes active in the field and to the consultants of these organizations. A separate abstract was prepared for each of the 3 papers included in the Annex of this document. Refs, figs, tabs

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  9. 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.

  10. SiC/SiC Cladding Materials Properties Handbook

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-01

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

  11. FRAPCON analysis of cladding performance during dry storage operations

    Energy Technology Data Exchange (ETDEWEB)

    Richmond, David J.; Geelhood, Kenneth J.

    2018-03-01

    There is an increasing need in the U.S. and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations (ISFSI) or interim storage sites. The NRC limits cladding temperature to 400°C while maintaining cladding hoop stress below 90 MPa in an effort to avoid radial hydride reorientation. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at 400 °C. Results were representative of the majority of U.S. LWR fuel. They conservatively showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.

  12. Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2008-01-01

    Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

  13. Aluminum hydroxide issue closure package

    International Nuclear Information System (INIS)

    Bergman, T.B.

    1998-01-01

    Aluminum hydroxide coatings on fuel elements stored in aluminum canisters in K West Basin were measured in July and August 1998. Good quality data was produced that enabled statistical analysis to determine a bounding value for aluminum hydroxide at a 99% confidence level. The updated bounding value is 10.6 kg per Multi-Canister Overpack (MCO), compared to the previously estimated bounding value of 8 kg/MCO. Thermal analysis using the updated bounding value, shows that the MCO generates oxygen concentrate that are below the lower flammability limits during the 40-year interim storage period and are, therefore, acceptable

  14. PREPARATION OF URANIUM-ALUMINUM ALLOYS

    Science.gov (United States)

    Moore, R.H.

    1962-09-01

    A process is given for preparing uranium--aluminum alloys from a solution of uranium halide in an about equimolar molten alkali metal halide-- aluminum halide mixture and excess aluminum. The uranium halide is reduced and the uranium is alloyed with the excess aluminum. The alloy and salt are separated from each other. (AEC)

  15. Annealing studies of Zircaloy-2 cladding at 580-850 deg C

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1983-01-01

    For fuel rod cladding it is important to determine if prior metallurgical condition combined with irradiation damage can influence high temperature deformation, because studies of such deformation are required to produce data for the cladding ballooning models which are used in analysing loss-of-coolant (LOCA). If the behaviour of all cladding conditions during a LOCA can be represented by, say, the annealed condition, then a great deal of experimental work on a multiplicity of cladding conditions can be avoided. By examining the metallographic structure and hardness, the present study determines the time required in the range 580 to 850 deg C for returning Zircaloy cladding to the annealed condition, so that for any transient a point can be specified where the material should have annealed. An equation has been derived to give this information. (author)

  16. Annealing studies of zircaloy-2 cladding at 580-8500C

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1978-05-01

    For fuel element cladding it is important to determine if prior metallurgical condition combined with irradiation damage can influence high temperature deformation, because studies of such deformation are required to produce data for the cladding ballooning models which are used in analysing loss-of-coolant accidents (LOCA). If the behaviour of all cladding conditions during a LOCA can be represented by, say, the annealed condition, then much experimental work on a multiplicity of cladding conditions can be avoided. By examining the metallographic structure and hardness, the present study determines the time required in the range 580 to 850 0 C for returning Zircaloy cladding to the annealed condition, so that for any transient, a point can be specified where the material should have annealed. An equation has been derived to give this information. (author)

  17. 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.

  18. Sliding wear studies of microwave clad versus unclad surface of stainless steel 304

    Directory of Open Access Journals (Sweden)

    Akshata M. K.

    2018-01-01

    Full Text Available Small and large scale (gas power plant, hydro power plant, automobile industries are suffering by failure of component. Sometimes, it is also observed that the component which was failed due to these reasons are very much costly and replacement of those also very difficult due to the complex geometry. By using Microwave hybrid heating, WC-12Co based clads were developed on austenitic stainless steel (SS304. Microwave clads were developed by introducing the preplaced, preheated powder for a duration of 15 min to microwave radiation at 2.45GHz frequency and 900 W power in domestic microwave applicator. By using optical microscope and scanning electron microscope (SEM, the developed clads were characterized. By using pin-on-disk, wear performance of the WC-12Co based clads and unclad samples were tested. It is observed that developed clad samples performed superior wear resistance than unclad samples.

  19. Deposition of Co-Ti alloy on mild steel substrate using laser cladding

    International Nuclear Information System (INIS)

    Alemohammad, Hamidreza; Esmaeili, Shahrzad; Toyserkani, Ehsan

    2007-01-01

    Laser cladding of a Co-Ti alloy on a mild steel substrate is studied. Premixed powders with the composition of 85 wt% cobalt and 15 wt% titanium are pre-placed on the substrate and a moving laser beam at different velocities is used to produce clad layers well bounded to the substrate. Characteristics of the clad are investigated using optical microscopy, X-ray diffraction (XRD), energy dispersive spectroscopy (EDS) and microhardness tests. The results reveal that the intermetallic phase TiCo 3 and β (i.e. fcc) cobalt are formed in the clad layer. The clad layer can also have major dilution from the substrate depending on the laser scanning velocity. It is observed that a finer microstructure is achievable with higher laser velocities whereas higher hardness is achieved using lower velocities. The latter is due to the formation of a larger fraction of TiCo 3 phase

  20. Introduction of an Innovative Cladding Panel System for Multi-Story Buildings

    Directory of Open Access Journals (Sweden)

    Hathairat Maneetes

    2014-08-01

    Full Text Available An Energy Dissipating Cladding System has been developed for use in buildings designed based on the concept of damage-controlled structure in seismic design. This innovative cladding panel system is capable of functioning both as a structural brace, as well as a source of energy dissipation, without demanding inelastic action and ductility from the basic lateral force resisting system. The structural systems of many modern buildings typically have large openings to accommodate glazing systems, and a popular type of construction uses spandrel precast cladding panels at each floor level that supports strip window systems. The present study focuses on developing spandrel type precast concrete cladding panels as supplementary energy dissipating devices that are added to the basic structural system. Through a series of analytical studies, the result of evaluating the ability of the proposed Energy Dissipating Cladding system to improve the earthquake resistance of the buildings is presented here.

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

    International Nuclear Information System (INIS)

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

    1983-09-01

    The technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel are summarized. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. 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 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 about 15,000 fuel rods, and about 5600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570 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 about 270 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 US. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380 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 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

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Siefken, L.J.

    1999-01-01

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

  4. Chrome - Free Aluminum Coating System

    Science.gov (United States)

    Bailey, John H.; Gugel, Jeffrey D.

    2010-01-01

    This slide presentation concerns the program to qualify a chrome free coating for aluminum. The program was required due to findings by OSHA and EPA, that hexavalent chromium, used to mitigate corrosion in aerospace aluminum alloys, poses hazards for personnel. This qualification consisted of over 4,000 tests. The tests revealed that a move away from Cr+6, required a system rather than individual components and that the maximum corrosion protection required pretreatment, primer and topcoat.

  5. Optomechanics of Single Aluminum Nanodisks.

    Science.gov (United States)

    Su, Man-Nung; Dongare, Pratiksha D; Chakraborty, Debadi; Zhang, Yue; Yi, Chongyue; Wen, Fangfang; Chang, Wei-Shun; Nordlander, Peter; Sader, John E; Halas, Naomi J; Link, Stephan

    2017-04-12

    Aluminum nanostructures support tunable surface plasmon resonances and have become an alternative to gold nanoparticles. Whereas gold is the most-studied plasmonic material, aluminum has the advantage of high earth abundance and hence low cost. In addition to understanding the size and shape tunability of the plasmon resonance, the fundamental relaxation processes in aluminum nanostructures after photoexcitation must be understood to take full advantage of applications such as photocatalysis and photodetection. In this work, we investigate the relaxation following ultrafast pulsed excitation and the launching of acoustic vibrations in individual aluminum nanodisks, using single-particle transient extinction spectroscopy. We find that the transient extinction signal can be assigned to a thermal relaxation of the photoexcited electrons and phonons. The ultrafast heating-induced launching of in-plane acoustic vibrations reveals moderate binding to the glass substrate and is affected by the native aluminum oxide layer. Finally, we compare the behavior of aluminum nanodisks to that of similarly prepared and sized gold nanodisks.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-29

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

  7. Aluminum and Alzheimer's Disease.

    Science.gov (United States)

    Colomina, Maria Teresa; Peris-Sampedro, Fiona

    2017-01-01

    Aluminum (Al) is one of the most extended metals in the Earth's crust. Its abundance, together with the widespread use by humans, makes Al-related toxicity particularly relevant for human health.Despite some factors influence individual bioavailability to this metal after oral, dermal, or inhalation exposures, humans are considered to be protected against Al toxicity because of its low absorption and efficient renal excretion. However, several factors can modify Al absorption and distribution through the body, which may in turn progressively contribute to the development of silent chronic exposures that may lately trigger undesirable consequences to health. For instance, Al has been recurrently shown to cause encephalopathy, anemia, and bone disease in dialyzed patients. On the other hand, it remains controversial whether low doses of this metal may contribute to developing Alzheimer's disease (AD), probably because of the multifactorial and highly variable presentation of the disease.This chapter primarily focuses on two key aspects related to Al neurotoxicity and AD, which are metabolic impairment and iron (Fe) alterations. We discuss sex and genetic differences as a plausible source of bias to assess risk assessment in human populations.

  8. Managing aluminum phosphide poisonings

    Science.gov (United States)

    Gurjar, Mohan; Baronia, Arvind K; Azim, Afzal; Sharma, Kalpana

    2011-01-01

    Aluminum phosphide (AlP) is a cheap, effective and commonly used pesticide. However, unfortunately, it is now one of the most common causes of poisoning among agricultural pesticides. It liberates lethal phosphine gas when it comes in contact either with atmospheric moisture or with hydrochloric acid in the stomach. The mechanism of toxicity includes cellular hypoxia due to the effect on mitochondria, inhibition of cytochrome C oxidase and formation of highly reactive hydroxyl radicals. The signs and symptoms are nonspecific and instantaneous. The toxicity of AlP particularly affects the cardiac and vascular tissues, which manifest as profound and refractory hypotension, congestive heart failure and electrocardiographic abnormalities. The diagnosis of AlP usually depends on clinical suspicion or history, but can be made easily by the simple silver nitrate test on gastric content or on breath. Due to no known specific antidote, management remains primarily supportive care. Early arrival, resuscitation, diagnosis, decrease the exposure of poison (by gastric lavage with KMnO4, coconut oil), intensive monitoring and supportive therapy may result in good outcome. Prompt and adequate cardiovascular support is important and core in the management to attain adequate tissue perfusion, oxygenation and physiologic metabolic milieu compatible with life until the tissue poison levels are reduced and spontaneous circulation is restored. In most of the studies, poor prognostic factors were presence of acidosis and shock. The overall outcome improved in the last decade due to better and advanced intensive care management. PMID:21887030

  9. The Resource-Saving Technology of Aluminum Nitride Obtaining During Combustion of Aluminum Nanopowder in Air

    OpenAIRE

    Ilyin, Aleksandr Petrovich; Mostovshchikov, Andrey Vladimirovich; Root, Lyudmila Olegovna

    2016-01-01

    The resource-saving technology of aluminum nitride obtaining during the combustion of aluminum nanopowder in air has been analyzed in the article. The investigation of the crystal phases of aluminum nanopowder combustion products obtained under the magnetic field exposure has been made. The experimental results showed the increase of aluminum nitride content up to 86 wt. % in comparison with the aluminum nitride content in combustion products without any exposure. The mechanism of aluminum ni...

  10. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    1985-11-01

    Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

  11. Experimental database of E110 claddings exposed to accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Perez-Fero, Erzsebet, E-mail: pereze@aeki.kfki.h [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, P.O. Box 49, H-1525 Budapest (Hungary); Gyori, Csaba [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Matus, Lajos; Vasaros, Laszlo; Hozer, Zoltan; Windberg, Peter; Maroti, Laszlo; Horvath, Marta; Nagy, Imre; Pinter-Csordas, Anna; Novotny, Tamas [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, P.O. Box 49, H-1525 Budapest (Hungary)

    2010-02-15

    An experimental database of E110 alloy has been developed on the basis of about 600 separate and combined effect tests of the Hungarian Academy of Sciences KFKI Atomic Energy Research Institute. It contains the data of oxidation, ballooning, tensile and compression tests, the results of post-test investigations, photos, figures, information concerning the test conditions and the corresponding English-language publications. The aim of this database is to give adequate information on the E110 cladding behaviour (oxidation, hydrogen uptake, mechanical performance) under accident conditions and to provide valuable experimental data for model development and code validation. This database is a part of the International Fuel Performance Experimental Database. It is accessible on-line, via the internet. This paper gives an overview of the experiments, the test facilities and conditions involved in the database. It presents the most important results and consequences and introduces the directory structure of the database.

  12. Composite fuel-cladding tubes and its fabrication

    International Nuclear Information System (INIS)

    Higashinakagawa, Emiko; Kawashima, Junko; Sato, Kanemitsu; Kuwae, Ryosho.

    1985-01-01

    Purpose: To reduce stress-corrosion cracks in a fuel cladding tube. Method: By inserting the sleeve of pure zirconium forming a liner layer into the hollow billet of zirconium alloy as an outer tube, then the composed tube is completed through hot-extruding. Then, the composite tube is pressed, by cold working through several passes, into a tube with a predetermined smaller inner-diameter and thinner wall. Heat treatment is applied between each of the passes of the cold working to recrystallize the zirconium-alloy outer tube substantially, as well as to integrally join the outer tube and the liner layer metal-lurgically. It serves as a buffer for moderating the mechanical interaction with the fuel pellets, whereby the resistance to stress-corrosion cracks can be increased. (Moriyama, K.)

  13. Method to produce carbon-cladded nuclear fuel particles

    International Nuclear Information System (INIS)

    Sturge, D.W.; Meaden, G.W.

    1978-01-01

    In the method charges of micro-spherules of fuel element are designed to have two carbon layers, whereby a one aims to achieve a uniform granulation (standard measurement). Two drums are used for this purpose connected behind one another. The micro-spherules coated with the first layer (phenolformaldehyde resin coated graphite particles) leave the first drum and enter the second one. Following the coating with a second layer, the micro-spherules are introduced into a grain size separator. The spherules that are too small are directly recycled into the second drum and those ones that are too large are recycled into the first drum after removing the graphite layers. The method may also be applied to metal cladded particles to manufacture cermet fuels. (RW) [de

  14. Bioactivity of calcium phosphate bioceramic coating fabricated by laser cladding

    Science.gov (United States)

    Zhu, Yizhi; Liu, Qibin; Xu, Peng; Li, Long; Jiang, Haibing; Bai, Yang

    2016-05-01

    There were always strong expectations for suitable biomaterials used for bone regeneration. In this study, to improve the biocompatiblity of titanium alloy, calcium phosphate bioceramic coating was obtained by laser cladding technology. The microstructure, phases, bioactivity, cell differentiation, morphology and resorption lacunae were investigated by optical microscope (OM), x-ray diffraction (XRD), methyl thiazolyl tetrazolium (MTT) assay, tartrate-resistant acid phosphatase (TRAP) staining and scanning electronic microscope (SEM), respectively. The results show that bioceramic coating consists of three layers, which are a substrate, an alloyed layer and a ceramic layer. Bioactive phases of β-tricalcium phosphate (β-TCP) and hydroxyapatite (HA) were found in ceramic coating. Osteoclast precursors have excellent proliferation on the bioceramic surface. The bioceramics coating could be digested by osteoclasts, which led to the resorption lacunae formed on its surface. It revealed that the gradient bioceramic coating has an excellent bioactivity.

  15. Bioactivity of calcium phosphate bioceramic coating fabricated by laser cladding

    International Nuclear Information System (INIS)

    Zhu, Yizhi; Liu, Qibin; Xu, Peng; Li, Long; Jiang, Haibing; Bai, Yang

    2016-01-01

    There were always strong expectations for suitable biomaterials used for bone regeneration. In this study, to improve the biocompatiblity of titanium alloy, calcium phosphate bioceramic coating was obtained by laser cladding technology. The microstructure, phases, bioactivity, cell differentiation, morphology and resorption lacunae were investigated by optical microscope (OM), x-ray diffraction (XRD), methyl thiazolyl tetrazolium (MTT) assay, tartrate-resistant acid phosphatase (TRAP) staining and scanning electronic microscope (SEM), respectively. The results show that bioceramic coating consists of three layers, which are a substrate, an alloyed layer and a ceramic layer. Bioactive phases of β-tricalcium phosphate (β-TCP) and hydroxyapatite (HA) were found in ceramic coating. Osteoclast precursors have excellent proliferation on the bioceramic surface. The bioceramics coating could be digested by osteoclasts, which led to the resorption lacunae formed on its surface. It revealed that the gradient bioceramic coating has an excellent bioactivity. (letter)

  16. Fully-automated weld-cladding in boiler units

    Energy Technology Data Exchange (ETDEWEB)

    Heitz, S. [DH Schweisstechnologie und Service, Hohenthurm GmbH und Co. KG, Hohenthurm (Germany); Treder, M. [IMT Ingenieurbuero Martin Treder, Automatisierung und Industrieelektronik, Leisnig (Germany); Lorenz, G. [Siemens AG, Automation und Drives (A und D), Systems Engineering, Fuerth (Germany)

    2001-07-01

    526ld-up welding is a proven and cost-effective method of renovating heating surfaces in power plants and waste incineration plants, for example. However, personnel are subjected to dust, dirt, and heat and often have to work in confined conditions. A newly developed and patented equipment technology ensures better quality on a continuous basis and relieves the burden on personnel engaged in the weld-cladding of finned and diaphragm tube walls. (orig.) [German] Zur Sanierung von Heizflaechen in Kraftwerks- oder Muellverbrennungsanlagen ist das Auftragschweissen eine bewaehrte und wirtschaftliche Methode. Das Personal ist dabei jedoch durch Staub, Schmutz und Hitze belastet und muss nicht selten unter beengten Bedingungen arbeiten. Eine neu entwickelte und patentierte Geraetetechnik sorgt bei der Herstellung von Schweissplattierungen an Flossen- bzw. Membranrohrwaenden fuer bessere und kontinuierliche Qualitaet und entlastet das Bedienpersonal. (orig.)

  17. Pellet cladding interaction: mechanical and chemical aproach to modelling

    International Nuclear Information System (INIS)

    Atabek, R.; Chantant, M.; Pineira, T.; Joseph, J.

    1980-09-01

    An important experimental irradiation programme has been carried out for several years in order to determine the operating limits of PWR fuel elements, during power transients. In addition to the correlation giving the permissible power limit in terms of specific burn-up, the examinations after irradiation on the fuel rods provided results that made it possible to develop mechanical and chemical models that can explain the pellet-cladding interaction phenomena. The mechanical process is described by means of a code using the finite element method. This paper gives the description of the code and the comparison of the experiment-calculation results. The modelization of the chemical process is based on the analyses (qualitative and quantitative) of gamma spectrometry, carried out on sections of fuel rods having undergone a transient. The variations in radial concentration of the cesium and iodine have been particularly studied [fr

  18. Fuel- and clad-motion diagnostics: licensing needs

    International Nuclear Information System (INIS)

    Bari, R.A.; Meyer, J.F.

    1976-01-01

    The paper addresses the current state of uncertainty with respect to fuel and clad motion during a hypothetical core-disruptive accident in a liquid metal fast breeder reactor as it relates to licensing needs. It should be noted that the paper does not represent an official position of the U.S. Nuclear Regulatory Commission, but rather, represents, in part, opinions and conclusions of its contractors. Particular attention is given to the needs for an assessment of the course of events during a hypothetical core-disruptive accident in the Clinch River Breeder Reactor. However, some of the issues discussed are likely to be relevant to larger breeder reactors as well. The issues addressed are related to the needs associated with analyses of the loss-of-flow (LOF) accident without scram and the transient overpower (TOP) accident, without scram

  19. Iodine stress-corrosion cracking in irradiated Zircaloy cladding

    International Nuclear Information System (INIS)

    Mattas, R.F.; Yaggee, F.L.; Neimark, L.A.

    1979-01-01

    Irradiated Zircaloy cladding specimens, which had experienced fluences from 0.1 to 6 x 10 21 n/cm 2 (E>0.1 MeV), were gas-pressure tested in an iodine environment to investigate their stress-corrosion cracking (SCC) susceptibility. The test temperatures and hoop stresses ranged from 320 to 360 0 C and 150 to 500 MPa, respectively. The results indicate that irradiation, in general, increases the susceptibility of Zircaloy to iodine SCC. For specimens that experienced fluences >2 x 10 21 n/cm 2 (E>0.1 MeV), the 24-h failure stress was 177+-18 MPa, regardless of the preirradiation metallurgical condition. An analytical model for iodine SCC has been developed which agrees reasonably well with the test results

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

    International Nuclear Information System (INIS)

    Gamble, Kyle Allan Lawrence; Hales, Jason Dean; Barani, Tommaso; Pizzocri, Davide; Pastore, Giovanni

    2016-01-01

    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.

  1. 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.

  2. Laser Cladding of Ultra-Thin Nickel-Based Superalloy Sheets.

    Science.gov (United States)

    Gabriel, Tobias; Rommel, Daniel; Scherm, Florian; Gorywoda, Marek; Glatzel, Uwe

    2017-03-10

    Laser cladding is a well-established process to apply coatings on metals. However, on substrates considerably thinner than 1 mm it is only rarely described in the literature. In this work 200 µm thin sheets of nickel-based superalloy 718 are coated with a powder of a cobalt-based alloy, Co-28Cr-9W-1.5Si, by laser cladding. The process window is very narrow, therefore, a precisely controlled Yb fiber laser was used. To minimize the input of energy into the substrate, lines were deposited by setting single overlapping points. In a design of experiments (DoE) study, the process parameters of laser power, laser spot area, step size, exposure time, and solidification time were varied and optimized by examining the clad width, weld penetration, and alloying depth. The microstructure of the samples was investigated by optical microscope (OM) and scanning electron microscopy (SEM), combined with electron backscatter diffraction (EBSD) and energy dispersive X-ray spectroscopy (EDX). Similarly to laser cladding of thicker substrates, the laser power shows the highest influence on the resulting clad. With a higher laser power, the clad width and alloying depth increase, and with a larger laser spot area the weld penetration decreases. If the process parameters are controlled precisely, laser cladding of such thin sheets is manageable.

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

    International Nuclear Information System (INIS)

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

    1981-09-01

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

  4. Analysis of pellet cladding mechanical interaction margins in PWR fuel under power ramp condition

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jong Sung; Lee, Jin Seok; Kim, Hyeong Koo; Chung, Jing On [Korea Nuclear Fuel, Daejeon (Korea, Republic of); Mitchell, David; Aleshin, Yuriy [Westinghouse Electric Company, South Carolina (Colombia)

    2008-10-15

    Small flaws in PWR and BWR fuel such as a missing pellet surface (MPS), pellet fragmentation and cladding defects, play an important role during conditions when there is pellet-cladding mechanical interaction (PCMI) under power ramp conditions. In order to confirm the margin against PWR fuel failure by PCMI with the pellet and cladding imperfections, first the separating hoop stress for cladding failure is determined using the ANSYS FEA model for failed and intact fuel rods under various power ramp conditions. Second, the idealized rod power history is developed to achieve the maximum uniform cladding stress during the normal operational transients. Finally, the stress multiplication factor is calculated with the FEA model to deal with the effect of various shapes and sizes of MPS, pellet fragmentation and cladding defects on the PCMI behavior. Then the stress multiplication factor with pellet imperfections and cladding defects allowed by manufacturing tolerances is applied to the fuel performance analysis code results using the idealized power history to confirm margin to PCMI under power ramp condition.

  5. Deep surface rolling for fatigue life enhancement of laser clad aircraft aluminium alloy

    Science.gov (United States)

    Zhuang, W.; Liu, Q.; Djugum, R.; Sharp, P. K.; Paradowska, A.

    2014-11-01

    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.

  6. Management of NPP severe accident by prevention of clad-softening

    International Nuclear Information System (INIS)

    Saxena, Anil Kumar; Limaye, Sanjay Prabhakar; Bera, Subrata; Deo, Anuj Kumar

    2015-01-01

    Specified Emergency Core Cooling System (ECCS) flow rate is testimony of clad reaching to temperature lower than its softening temperature during loss of coolant accident (LOCA) in nuclear reactors. Coolant channel(s) of nuclear reactors with vertical fuel-assemblies e. g. LWRs gets voided in a short time as a result of double ended guillotine rupture in coolant pipeline. There is rapid and almost steep rise in clad temperature due to stored energy and decay heat if ECCS flow rate is less than a specified value. A computer program, based on moving mesh methodology, is developed to calculate rewetting velocity. The program is validated using experimental data. Numerical equations are solved by marching technique. The paper will bring out the fact that if coolant flow is less than a specified value the wet front will not reach the top of the clad. This will result some unrewetted clad portion. As the heating of this unrewetted clad is continued it may result softening of clad. If it happens in many channels the integrity of clad as a whole will be lost. There will be high probability that severe accident will take place. The paper presents a method to prevent softening and thus to ensure accident free operation of nuclear reactor. (author)

  7. Performance of IN-706 and PE-16 cladding in mixed-oxide fuel pins

    International Nuclear Information System (INIS)

    Makenas, B.J.; Lawrence, L.A.; Jensen, B.W.

    1982-05-01

    Iron-nickel base, precipitation-strengthened alloys, IN-706 and PE-16, advanced alloy cladding considered for breeder reactor applications, were irradiated in mixed-oxide fuel pins in the HEDL-P-60 subassembly in EBR-II. Initial selection of candidate advanced alloys was done using only nonfueled materials test results. However, to establish the performance characteristics of the candidate cladding alloys, i.e., dimensional stability and structural integrity under conditions of high neutron flux, elevated temperature, and applied stress, it was necessary to irradiate fuel pins under typical operating conditions. Fuel pins were clad with solution treated IN-706 and PE-16 and irradiated to peak fluences of 6.1 x 10 22 n/cm 2 (E > .1 MeV) and 8.8 x 10 22 n/cm 2 (E > .1 MeV) respectively. Fabrication and operating parameters for the fuel pins with the advanced cladding alloy candidates are summarized. Irradiation of HEDL-P-60 was interrupted with the breach of a pin with IN-706 cladding at 5.1 at % and the test was terminated with cladding breach in a pin with PE-16 cladding at 7.6 at %

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

    Directory of Open Access Journals (Sweden)

    Ki-Nam Jang

    2017-12-01

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

  9. Transparent conducting oxide clad limited area epitaxy semipolar III-nitride laser diodes

    KAUST Repository

    Myzaferi, A.

    2016-08-11

    The bottom cladding design of semipolar III-nitride laser diodes is limited by stress relaxation via misfit dislocations that form via the glide of pre-existing threading dislocations (TDs), whereas the top cladding is limited by the growth time and temperature of the p-type layers. These design limitations have individually been addressed by using limited area epitaxy (LAE) to block TD glide in n-type AlGaN bottom cladding layers and by using transparent conducting oxide (TCO) top cladding layers to reduce the growth time and temperature of the p-type layers. In addition, a TCO-based top cladding should have significantly lower resistivity than a conventional p-type (Al)GaN top cladding. In this work, LAE and indium-tin-oxide cladding layers are used simultaneously in a (202⎯⎯1) III-nitride laser structure. Lasing was achieved at 446 nm with a threshold current density of 8.5 kA/cm2 and a threshold voltage of 8.4 V.

  10. Laser Cladding of Ultra-Thin Nickel-Based Superalloy Sheets

    Directory of Open Access Journals (Sweden)

    Tobias Gabriel

    2017-03-01

    Full Text Available Laser cladding is a well-established process to apply coatings on metals. However, on substrates considerably thinner than 1 mm it is only rarely described in the literature. In this work 200 µm thin sheets of nickel-based superalloy 718 are coated with a powder of a cobalt-based alloy, Co–28Cr–9W–1.5Si, by laser cladding. The process window is very narrow, therefore, a precisely controlled Yb fiber laser was used. To minimize the input of energy into the substrate, lines were deposited by setting single overlapping points. In a design of experiments (DoE study, the process parameters of laser power, laser spot area, step size, exposure time, and solidification time were varied and optimized by examining the clad width, weld penetration, and alloying depth. The microstructure of the samples was investigated by optical microscope (OM and scanning electron microscopy (SEM, combined with electron backscatter diffraction (EBSD and energy dispersive X-ray spectroscopy (EDX. Similarly to laser cladding of thicker substrates, the laser power shows the highest influence on the resulting clad. With a higher laser power, the clad width and alloying depth increase, and with a larger laser spot area the weld penetration decreases. If the process parameters are controlled precisely, laser cladding of such thin sheets is manageable.

  11. Order of magnitude cost appraisal for selected aspects of clad waste management

    Energy Technology Data Exchange (ETDEWEB)

    Zima, G.E.

    1977-02-01

    A simple formula, incorporating the fixed charge rate principle, is applied to a clad waste management exercise involving densification, canning, transportation and salt disposal. For the purpose of comparison with the bulk of published nuclear waste management costs, cost and fixed charge rate data appropriate to roughly the period 1970 to 1973 are used. Within the context of this order of magnitude appraisal, densification displays some cost advantage, reflected principally in the transportation cost. Dependent on the degree of densification, above a certain clad waste generation rate the transportation savings may be expected to exceed reasonable densification costs. There is no explicit consideration of the decontamination step in this appraisal. The limited accessibility of surface effect decontamination to internal transuranic and activation product contamination suggests a quite small influence of decontamination on the transportation and disposal costs. Decontamination may, however, have a significant effect on the ease of establishing a practicable containment envelope of high reliability throughout the clad waste history. A brief comparison is made of clad waste management costs with the major costs of the nuclear power economy. This comparison implies a virtually unlimited technical latitude for clad waste treatment in accommodating the public safety without significant perturbation of nuclear power costs. It is submitted that clad waste management optimization will be under the primal constraint of maximizing thelong term public safety, with economic analysis useful only as a discriminator between waste handling alternatives of sensibly equivalent containment qualities. Some areas of clad waste treatment meriting increased attention are noted.

  12. Gas evolution behavior of aluminum in mortar

    International Nuclear Information System (INIS)

    Hashizume, Shuji; Matsumoto, Junko; Banba, Tsunetaka

    1996-10-01

    As a part of study of leaching behavior for solidified dry low level radioactive waste, gas evolution behavior of aluminum in mortar was investigated, and a plan of our research was proposed. The effect of pH on corrosion rate of aluminum, corrosion product, time dependency of corrosion rate of aluminum in mortar, change of corrosion mechanism, the effects of Na, Ca and Cl ions on corrosion rate of aluminum in mortar and corrosion behavior of aluminum when aluminum was used as sacrificed anode in reinforced concrete were previously clarified. Study of the effects of environmental factors such as pH, kind of ions and temperature on gas evolution behavior of aluminum and the effect of aluminum/carbon steel surface ratio no gas evolution behavior of aluminum were planed. (author). 75 refs

  13. Gas evolution behavior of aluminum in mortar

    Energy Technology Data Exchange (ETDEWEB)

    Hashizume, Shuji; Matsumoto, Junko; Banba, Tsunetaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-10-01

    As a part of study of leaching behavior for solidified dry low level radioactive waste, gas evolution behavior of aluminum in mortar was investigated, and a plan of our research was proposed. The effect of pH on corrosion rate of aluminum, corrosion product, time dependency of corrosion rate of aluminum in mortar, change of corrosion mechanism, the effects of Na, Ca and Cl ions on corrosion rate of aluminum in mortar and corrosion behavior of aluminum when aluminum was used as sacrificed anode in reinforced concrete were previously clarified. Study of the effects of environmental factors such as pH, kind of ions and temperature on gas evolution behavior of aluminum and the effect of aluminum/carbon steel surface ratio no gas evolution behavior of aluminum were planed. (author). 75 refs.

  14. The role of cladding material for performance of LWR control assemblies

    International Nuclear Information System (INIS)

    Dewes, P.; Roppelt, A.

    2000-01-01

    The lifetime of control assemblies in LWRs can be limited presently by mechanical failure of the absorber cladding. The major cause of failure is mechanical interaction of the absorber with the cladding due to irradiation induced dimensional changes such as absorber swelling and cladding creep, resulting in cracking of the clad. Such failures occurred in both BWRs and PWRs. Experience and in-reactor tests revealed that cracking can be avoided principally by two ways: First, if strain rates and hence, stresses in the cladding are kept low (well below the yield strength), significant strains can be tolerated. This is the case for the cladding of PWR control assemblies with slowly swelling Ag-In-Cd absorber. Recent examinations of highly exposed PWR control assemblies confirmed the design correlation up to the presently used strain limit. Second, in such cases where strongly swelling absorber material like boron carbide is still preferred, materials which are resistant against irradiation assisted stress corrosion cracking (IASCC) can be used. The influence of material composition and condition on IASCC was studied in-reactor using tubular samples of various stainless steels and Ni-base alloys stressed by swelling mandrels. In several programme steps high purity materials with special features had been identified as resistant to IASCC. Another process of cladding damage which may occur in PWRs is wear caused by friction of the control rods in the surrounding guide structure. For replacement control assemblies this problem is solved by coating of the cladding. There exists meanwhile excellent experience of up to 18 operation cycles with coated claddings. (author)

  15. Laser performance and modeling of RE3+:YAG double-clad crystalline fiber waveguides

    Science.gov (United States)

    Li, Da; Lee, Huai-Chuan; Meissner, Stephanie K.; Meissner, Helmuth E.

    2018-02-01

    We report on laser performance of ceramic Yb:YAG and single crystal Tm:YAG double-clad crystalline fiber waveguide (CFW) lasers towards the goal of demonstrating the design and manufacturing strategy of scaling to high output power. The laser component is a double-clad CFW, with RE3+:YAG (RE = Yb, Tm respectively) core, un-doped YAG inner cladding, and ceramic spinel or sapphire outer cladding. Laser performance of the CFW has been demonstrated with 53.6% slope efficiency and 27.5-W stable output power at 1030-nm for Yb:YAG CFW, and 31.6% slope efficiency and 46.7-W stable output power at 2019-nm for Tm:YAG CFW, respectively. Adhesive-Free Bond (AFB®) technology enables a designable refractive index difference between core and inner cladding, and designable core and inner cladding sizes, which are essential for single transverse mode CFW propagation. To guide further development of CFW designs, we present thermal modeling, power scaling and design of single transverse mode operation of double-clad CFWs and redefine the single-mode operation criterion for the double-clad structure design. The power scaling modeling of double-clad CFW shows that in order to achieve the maximum possible output power limited by the physical properties, including diode brightness, thermal lens effect, and simulated Brillion scattering, the length of waveguide is in the range of 0.5 2 meters. The length of an individual CFW is limited by single crystal growth and doping uniformity to about 100 to 200 mm lengths, and also by availability of starting crystals and manufacturing complexity. To overcome the limitation of CFW lengths, end-to-end proximity-coupling of CFWs is introduced.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar; Allen, Todd; Cole, James

    2013-02-27

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

  17. Study of pellet clad interaction defects in Dresden-3 fuel rods

    International Nuclear Information System (INIS)

    Pasupathi, V.; Perrin, J.S.

    1979-01-01

    During Cycle-3 operation of Dresden-3, fuel rod failures occurred following a transient power increase. Ten fuel rods from five of the leaking fuel assemblies were examined at Battelle's Columbus Laboratory and General Electric-Vallecitos Nuclear Center. Examinations consisted of nondestructive and destructive methods including metallography and scanning electron microscopy (SEM). Results showed the cause of fuel rod failure to be pellet clad interaction involving stress corrosion cracking. Results of SEM studies of the cladding crack surfaces and deposits on clad inner surfaces were in agreement with those reported by other investigators

  18. Swelling behavior of 20% CW 316 Stainless Steel cladding irradiated with and without adjacent fuel

    International Nuclear Information System (INIS)

    Makenas, B.J.; Bates, J.F.; Jost, J.W.

    1982-06-01

    Swelling behavior has been evaluated for irradiated 20% CW 316 Stainless Steel used as cladding material for mixed-oxide fuel pins in EBR-II. This behavior has been compared statistically with the behavior of a large number of specimens which were irradiated without adjacent fuel in the same reactor. In spite of the chemical environment and stresses experienced by fueled cladding, the fueled and nonfueled cladding appear to behave in a similar manner although some divergence was noted for one of the cases studied

  19. Effects of commercial cladding on the fracture behavior of pressure vessel steel plates

    International Nuclear Information System (INIS)

    Iskander, S.K.; Alexander, D.J.; Bolt, S.E.; Cook, K.V.; Corwin, W.R.; Oland, B.C.; Nanstad, R.K.; Robinson, G.C.

    1988-01-01

    The objective of this program is to determine the effect, if any, of stainless steel cladding upon the propagation of small surface cracks subjected to stress states similar to those produced by thermal shock conditions. Preliminary results from testing at temperature 10 deg. C and 60 deg. C below NDT have shown that (1) a tough surface layer (cladding and/or HAZ) has arrested running flaws under conditions where unclad plates have ruptured, and (2) the residual load-bearing capacity of clad plates with large subclad flaws significantly exceeded that of an unclad plate. (author)

  20. Effects of commercial cladding on the fracture behavior of pressure vessel steel plates

    International Nuclear Information System (INIS)

    Iskander, S.K.; Alexander, D.J.; Bolt, S.E.; Cook, K.V.; Corwin, W.R.; Oland, B.C.; Nanstad, R.K.; Robinson, G.C.

    1987-01-01

    The objective of this program is to determine the effect, if any, of stainless steel cladding upon the propagation of small surface cracks subjected to stress states similar to those produced by thermal shock conditions. Preliminary results from testing at temperatures 10 0 and 60 0 C below NDT have shown that (1) a tough surface layer (cladding and/or HAZ) has arrested running flaws under conditions where unclad plates have ruptured, and (2) the residual load-bearing capacity of clad plates with large subclad flaws significantly exceeded that of an unclad plate

  1. Observations of in-reactor endurance and rupture life for fueled and unfueled FTR cladding

    International Nuclear Information System (INIS)

    Lovell, A.J.; Christensen, B.Y.; Chin, B.A.

    1979-01-01

    Reactor component endurance limits are important to nuclear experimenters and operators. This paper investigates endurance limits of 316 CW fuel pin cladding. The objective of this paper is to compare and analyze two different sets of FTR fuel pin cladding data. The first data set is from unfueled pressurized cladding irradiated in the Experimental Breeder Reactor No. II (EBR-II). This data set was generated in an assembly in which the temperature was monitored and controlled. The second data set contains observations of breached and unbreached EBR-II test fuel pins covering a large range of temperature, power and burnup conditions

  2. An integrated approach to selecting materials for fuel cladding in advanced high-temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rangacharyulu, C., E-mail: chary.r@usask.ca [Univ. of Saskatchewan, Saskatoon, SK (Canada); Guzonas, D.A.; Pencer, J.; Nava-Dominguez, A.; Leung, L.K.H. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    An integrated approach has been developed for selection of fuel cladding materials for advanced high-temperature reactors. Reactor physics, thermalhydraulic and material analyses are being integrated in a systematic study comparing various candidate fuel-cladding alloys. The analyses established the axial and radial neutron fluxes, power distributions, axial and radial temperature distributions, rates of defect formation and helium production using AECL analytical toolsets and experimentally measured corrosion rates to optimize the material composition for fuel cladding. The project has just been initiated at University of Saskatchewan. Some preliminary results of the analyses are presented together with the path forward for the project. (author)

  3. Evolution of the microstructure and tribological performance of Ti–6Al–4V cladding with TiN powder

    International Nuclear Information System (INIS)

    Lin, Yu-Chi; Lin, Yuan-Ching; Chen, Yong-Chwang

    2012-01-01

    Highlights: ► Titanium nitrides (TiNs) powder was successfully clad on Ti–6Al–4V by a gas tungsten arc welding (GTAW) cladding process. ► Solidification process of the TiN cladding layer was investigated and discussed. ► Microhardness distribution and wear mechanism of the TiN clad layer were discussed. ► Wear performance of the TiN clad layer is higher than that of the Ti–6Al–4V substrate tenfold. -- Abstract: Titanium nitrides (TiNs) powder was used as a material to resist wear; it was then clad onto a Ti–6Al–4V substrate by gas tungsten arc welding (GTAW). During the cladding process, the TiN x reinforcing phase was formed in situ within the clad layer. Since the TiN x reinforcing phase exists within the clad layer, the hardness of the clad layer is double that of the substrate. Wear test results reveal that the wear resistance of TiN clad layer is up to ten times more resistant than the Ti–6Al–4V substrate. From the worn surface analysis, the primary wear mechanism of the Ti–6Al–4V specimen exhibited oxidation wear combined with adhesive wear, and the TiN clad layer specimen exhibited abrasive wear. This investigation also discusses the mechanism for forming the clad layer microstructure. During solidification of the clad layer, the motion of the liquid–solid interface caused the oval TiN x phase to cluster, producing a dendritic appearance.

  4. PWR clad ballooning: The effect of circumferential clad temperature variations on the burst strain/burst temperature relationship

    International Nuclear Information System (INIS)

    Barlow, P.

    1983-01-01

    By experiment, it has been shown by other workers that there is a reduction in the creep ductility of Zircaloy 4 in the α+β phase transition region. Results from single rod burst tests also show a reduction in burst strain in the α+β phase region. In this report it is shown theoretically that for single rod burst tests in the presence of circumferential temperature gradients, the temperature dependence of the mean burst strain is not determined by temperature variations in creep ductility, but is governed by the temperature sensitivity of the creep strain rate, which is shown to be a maximum in the α+β phase transition region. To demonstrate this effect, the mean clad strain at burst was calculated for creep straining at different temperature levels in the α, α+β and β phase regions. Cross-pin temperature gradients were applied which produced strain variations around the clad which were greatest in the α+β phase region. The mean strain at burst was determined using a maximum local burst strain (i.e. a creep ductility) which is independent of temperature. By assuming cross-pin temperature gradients which are typical of those observed during burst tests, then the calculated mean burst strain/burst temperature relationship gave good agreement with experiment. The calculations also show that when circumferential temperature differences are present, the calculated mean strain at burst is not sensitive to variations in the magnitude of the assumed creep ductility. This reduces the importance of the assumed burst criterion in the calculations. Hence a temperature independent creep ductility (e.g. 100% local strain) is adequate as a burst criterion for calculations under PWR LOCA conditions. (author)

  5. Scaleable Clean Aluminum Melting Systems

    Energy Technology Data Exchange (ETDEWEB)

    Han, Q.; Das, S.K. (Secat, Inc.)

    2008-02-15

    The project entitled 'Scaleable Clean Aluminum Melting Systems' was a Cooperative Research and Development Agreements (CRADAs) between Oak Ridge National Laboratory (ORNL) and Secat Inc. The three-year project was initially funded for the first year and was then canceled due to funding cuts at the DOE headquarters. The limited funds allowed the research team to visit industrial sites and investigate the status of using immersion heaters for aluminum melting applications. Primary concepts were proposed on the design of furnaces using immersion heaters for melting. The proposed project can continue if the funding agency resumes the funds to this research. The objective of this project was to develop and demonstrate integrated, retrofitable technologies for clean melting systems for aluminum in both the Metal Casting and integrated aluminum processing industries. The scope focused on immersion heating coupled with metal circulation systems that provide significant opportunity for energy savings as well as reduction of melt loss in the form of dross. The project aimed at the development and integration of technologies that would enable significant reduction in the energy consumption and environmental impacts of melting aluminum through substitution of immersion heating for the conventional radiant burner methods used in reverberatory furnaces. Specifically, the program would couple heater improvements with furnace modeling that would enable cost-effective retrofits to a range of existing furnace sizes, reducing the economic barrier to application.

  6. Chemical synthesis of aluminum nanoparticles

    Energy Technology Data Exchange (ETDEWEB)

    Ghanta, Sekher Reddy; Muralidharan, Krishnamurthi, E-mail: kmsc@uohyd.ernet.in [Advanced Center of Research in High Energy Materials (ACRHEM), University of Hyderabad (India)

    2013-06-15

    An alternate synthetic route has been described for the production of aluminum nanoparticles (Al-NPs). These Al-NPs were obtained through a reduction of aluminum acetylacetonate [Al(acac){sub 3}] by lithium aluminum hydride (LiAlH{sub 4}) in mestitylene at 165 Degree-Sign C. The side products were removed by repeated washing with dry, ice cold methanol and the reaction mixture was filtered to obtain gray-colored Al-NPs. The synthesized nanoparticles were characterized by Powder X-ray diffraction pattern and {sup 27}Al-MAS-NMR spectrum. The X-ray diffraction pattern confirmed the formation of face-centered cubic (fcc) form of aluminum. The size and morphology were investigated by scanning electron microscope and transmission electron microscope which showed particle of varying shapes with size ranging from 50 to 250 nm. The weight loss from the nanoparticles was studied by thermo gravimetric analysis which indicated that the nanoparticles were tightly bound with an unknown amorphous organic residue which cannot be removed by simple washing. The carbonaceous residue might be outcome of the decomposition of acac ligand which was responsible in stabilizing aluminum nanoparticles.

  7. Thermophysical Properties of Liquid Aluminum

    Science.gov (United States)

    Leitner, Matthias; Leitner, Thomas; Schmon, Alexander; Aziz, Kirmanj; Pottlacher, Gernot

    2017-06-01

    Ohmic pulse-heating with sub-microsecond time resolution is used to obtain thermophysical properties for aluminum in the liquid phase. Measurement of current through the sample, voltage drop across the sample, surface radiation, and volume expansion allow the calculation of specific heat capacity and the temperature dependencies of electrical resistivity, enthalpy, and density of the sample at melting and in the liquid phase. Thermal conductivity and thermal diffusivity as a function of temperature are estimated from resistivity data using the Wiedemann-Franz law. Data for liquid aluminum obtained by pulse-heating are quite rare because of the low melting temperature of aluminum with 933.47 K (660.32 °C), as the fast operating pyrometers used for the pulse-heating technique with rise times of about 100 ns generally might not be able to resolve the melting plateau of aluminum because they are not sensitive enough for such low temperature ranges. To overcome this obstacle, we constructed a new, fast pyrometer sensitive in this temperature region. Electromagnetic levitation, as the second experimental approach used, delivers data for surface tension (this quantity is not available by means of the pulse-heating technique) and for density of aluminum as a function of temperature. Data obtained will be extensively compared to existing literature data.

  8. 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

  9. Demonstration of fuel resistant to pellet-cladding interaction. Phase I. Final report

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1979-03-01

    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 final report for PHASE 1 of this program. Support tests have shown that the barrier fuel resists PCI far better than does the conventional Zircaloy-clad fuel. Power ramp tests thus far have shown good PCI resistance for Cu-barrier fuel at burnup > 12 MWd/kg-U and for Zr-liner fuel > 16 MWd/kg-U. The program calls for continued testing to still higher burnup levels in PHASE 2

  10. Microstructure and properties of laser clad coatings studied by orientation imaging microscopy

    NARCIS (Netherlands)

    Ocelik, V.; Furar, I.; De Hosson, J. Th. M.

    2010-01-01

    In this work orientation imaging microscopy (OIM), based on electron backscatter diffraction in scanning electron microscopy, was employed to examine in detail the relationship between laser cladding processing parameters and he properties and the microstructure of single and overlapping laser

  11. Theory of the frictional interaction between nuclear fuel cladding and a cracked ceramic pellet

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1976-02-01

    A summary is presented of the outcome of theoretical work detailed in five publications, reproduced as appendices, which is concerned with the tendency for the cladding tube of nuclear fuel elements to fracture as the result of power cycling or after a sudden upward power excursion. The relationship is shown between the properties of the clad, those of UO 2 pellets, and the tendency of the clad to fail during upward power excursions. The role of interfacial friction is explored and the benefit to be obtained by reducing it is calculated for cases where a soft metal interlayer is present. It is shown that the experimentally-confirmed magnitude of the strain-concentration in the arc of cladding over a radial pellet crack could not arise if there were interfaceons present. Accordingly, these defects, although they do occur in some sliding situations, are thought to be absent from the pellet clas interface in fuel pins. (author)

  12. The Development of Finite Element Model for Pellet-Cladding Mechanical Interaction

    International Nuclear Information System (INIS)

    Lee, Jin-Seok; Yoo, Jong-Sung; Kim, Hyeong-Koo; Kim, Yong-Hwan; Lee, Chong-Chul; Mitchell, D.; Aleshin, Y.

    2007-01-01

    This paper studies a FEA(Finite Element Analysis) model to describe PCMI under a power ramp and pressure loading conditions. From this PCMI model, the stress fields of pellet and cladding are evaluated. In order to compute the stress and strain fields in the cladding which were failed under the power ramp, this model might be extended to include the missing chip shape which occurred in the process of assembling or manufacturing of fuel rod. This missing chip of pellet has an affects on the temperature and the stress of the cladding in the vicinity of the chip edge so that the possibility of fuel failure is increased during the power ramp. The pellet fragmentation can be also accounted because the fuel cracking occurs immediately after reactor start-up and plays an important role in relaxing stresses in pellet but escalating the stress in cladding

  13. The Effect of Rare Earth on the Structure and Performance of Laser Clad Coatings

    Science.gov (United States)

    Bao, Ruiliang; Yu, Huijun; Chen, Chuanzhong; Dong, Qing

    Laser cladding is one kind of advanced surface modification technology and has the abroad prospect in making the wear-resistant coating on metal substrates. However, the application of laser cladding technology does not achieve the people's expectation in the practical production because of many defects such as cracks, pores and so on. The addiction of rare earth can effectively reduce the number of cracks in the clad coating and enhance the coating wear-resistance. In the paper, the effects of rare earth on metallurgical quality, microstructure, phase structure and wear-resistance are analyzed in turns. The preliminary discussion is also carried out on the effect mechanism of rare earth. At last, the development tendency of rare earth in the laser cladding has been briefly elaborated.

  14. Fabrication of versatile cladding light strippers and fiber end-caps with CO2 laser radiation

    Science.gov (United States)

    Steinke, M.; Theeg, T.; Wysmolek, M.; Ottenhues, C.; Pulzer, T.; Neumann, J.; Kracht, D.

    2018-02-01

    We report on novel fabrication schemes of versatile cladding light strippers and end-caps via CO2 laser radiation. We integrated cladding light strippers in SMA-like connectors for reliable and stable fiber-coupling of high-power laser diodes. Moreover, the application of cladding light strippers in typical fiber geometries for high-power fiber lasers was evaluated. In addition, we also developed processes to fuse end-caps to fiber end faces via CO2 laser radiation and inscribe the fibers with cladding light strippers near the end-cap. Corresponding results indicate the great potential of such devices as a monolithic and low-cost alternative to SMA connectors.

  15. Effect of reactor chemistry and operating variables on fuel cladding corrosion in PWRs

    International Nuclear Information System (INIS)

    Park, Moon Ghu; Lee, Sang Hee

    1997-01-01

    As the nuclear industry extends the fuel cycle length, waterside corrosion of zircaloy cladding has become a limiting factor in PWR fuel design. Many plant chemistry factors such as, higher lithium/boron concentration in the primary coolant can influence the corrosion behavior of zircaloy cladding. The chemistry effect can be amplified in higher duty fuel, particularlywhen surface boiling occurs. Local boiling can result in increased crud deposition on fuel cladding which may induce axial power offset anomalies (AOA), recently reported in several PWR units. In this study, the effect of reactor chemistry and operating variables on Zircaloy cladding corrosion is investigated and simulation studies are performed to evaluate the optimal primary chemistry condition for extended cycle operation. (author). 8 refs., 3 tabs., 16 figs

  16. The influence of ventilation on moisture conditions in facades with wooden cladding

    DEFF Research Database (Denmark)

    Hansen, Ernst Jan de Place; Brandt, Erik

    2009-01-01

    A ventilated cavity behind the cladding of timber frame walls is often considered good building practice that facilitates the removal of moisture from the construction. However, moisture will only be removed from the construction by ventilating it with dry air, whereas ventilating with humid air...... might add moisture to the construction. Full-size wall elements with wooden cladding placed in a test building were exposed to natural climate on the outside and to a humid indoor climate on the inside. Temperature and moisture conditions inside the wall elements and climate parameters were monitored....... Test parameters included cavity/unventilated cavity/no cavity, cavity size, vent geometry, type of cladding and type of wind barrier. The potential durability of the wooden façade claddings was evaluated by coupling the measured time series of moisture content, temperature and time by means of a model...

  17. Low-Stress Silicon Cladding for Surface Finishing Large UVOIR Mirrors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this Phase I research, ZeCoat Corporation will develop an affordable, low-stress silicon cladding process which is super-polishable for large UVOIR mirrors. The...

  18. The state-of-the-art laser bio-cladding technology

    Science.gov (United States)

    Liu, Jichang; Fuh, J. Y. H.; Lü, L.

    2010-11-01

    The current state and future trend of laser bio-cladding technology are discussed. Laser bio-cladding is used in implants including fabrication of metal scaffolds and bio-coating on the scaffolds. Scaffolds have been fabricated from stainless steel, Co-based alloy or Ti alloy using laser cladding, and new laser-deposited Ti alloys have been developed. Calcium phosphate bioceramic coatings have been deposited on scaffolds with laser to improve the wear resistence and corrosion resistence of implants and to induce bone regeneration. The types of biomaterial devices currently available in the market include replacement heart valve prosthesis, dental implants, hip/knee implants, catheters, pacemakers, oxygenators and vascular grafts. Laser bio-cladding process is attracting more and more attentions of people.

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

    International Nuclear Information System (INIS)

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

    1978-06-01

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

  20. Numerical study to represent non-isothermal melt-crystallization kinetics at laser-powder cladding

    CSIR Research Space (South Africa)

    Niziev, VG

    2013-04-01

    Full Text Available The study of laser-powder cladding process subject to heat transfer, melting and crystallization kinetics has been carried out numerically and experimentally. The Kolmogorov-Avrami equation was applied to describe the kinetics of the phase...

  1. Development and study the performance of PBA cladding modified fiber optic intrinsic biosensor for urea detection

    Energy Technology Data Exchange (ETDEWEB)

    Botewad, S. N.; Pahurkar, V. G.; Muley, G. G., E-mail: gajananggm@yahoo.co.in [Department of Physics, Sant Gadge Baba Amravati University, Amravati, Maharashtra, India-444602 (India)

    2016-05-06

    The fabrication and study of a cladding modified fiber optic intrinsic urea biosensor based on evanescent wave absorbance has been presented. The sensor was prepared using cladding modification technique by removing a small portion of cladding of an optical fiber and modifying with an active cladding of porous polyaniline-boric acid (PBA) matrix to immobilize enzyme-urease through cross-linking via glutaraldehyde. The nature of as-synthesized and deposited PBA film on fiber optic sensing element was studied by ultraviolet-visible (UV-vis) spectroscopy and X-ray diffraction (XRD) analysis. The performance of the developed sensor was studied for different urea concentrations in solutions prepared in phosphate buffer.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-12

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

  3. Circumferential nonuniformity of cladding radiation swelling of fast reactor peripheral fuel elements

    International Nuclear Information System (INIS)

    Reutov, V.F.; Farkhutdinov, K.G.

    1977-01-01

    The results are presented of the investigation into the perimeter radiation swelling of Kh18N10T stainless steel cladding in different cross sections of a peripheral fuel element of the BR-5 reactor. The fluence on the cladding is 1.8-2.9 x 10 22 fast neutr/cm 2 , the operating temperatures in different parts of the fuel element being 430 deg to 585 deg C. There has been observed circumferential non-uniformity of the distribution, concentration, and of the total volume of radiation cavities, which is due to temperature non-uniformity along the cladding perimeter. It is shown that such non-uniformity of radiation swelling of the cladding material may result in bending of the peripheral fuel element with regard to the fuel assembly sheath walls

  4. Aluminum nitride insulating films for MOSFET devices

    Science.gov (United States)

    Lewicki, G. W.; Maserjian, J.

    1972-01-01

    Application of aluminum nitrides as electrical insulator for electric capacitors is discussed. Electrical properties of aluminum nitrides are analyzed and specific use with field effect transistors is defined. Operational limits of field effect transistors are developed.

  5. Aluminum--Industry of the Future

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-01-23

    This 8-page brochure describes the Office of Industrial Technologies' Aluminum Industry of the Future; a partnership between the Department of Energy and the aluminum industry established to increase industrial energy and cost efficiency.

  6. Aluminum-carbon composite electrode

    Science.gov (United States)

    Farahmandi, C.J.; Dispennette, J.M.

    1998-07-07

    A high performance double layer capacitor having an electric double layer formed in the interface between activated carbon and an electrolyte is disclosed. The high performance double layer capacitor includes a pair of aluminum impregnated carbon composite electrodes having an evenly distributed and continuous path of aluminum impregnated within an activated carbon fiber preform saturated with a high performance electrolytic solution. The high performance double layer capacitor is capable of delivering at least 5 Wh/kg of useful energy at power ratings of at least 600 W/kg. 3 figs.

  7. Aluminum-air battery crystallizer

    Science.gov (United States)

    Maimoni, A.

    1987-01-01

    A prototype crystallizer system for the aluminum-air battery operated reliably through simulated startup and shutdown cycles and met its design objectives. The crystallizer system allows for crystallization and removal of the aluminium hydroxide reaction product; it is required to allow steady-state and long-term operation of the aluminum-air battery. The system has to minimize volume and maintain low turbulence and shear to minimize secondary nucleation and energy consumption while enhancing agglomeration. A lamella crystallizer satisfies system constraints.

  8. Aluminum-carbon composite electrode

    Science.gov (United States)

    Farahmandi, C. Joseph; Dispennette, John M.

    1998-07-07

    A high performance double layer capacitor having an electric double layer formed in the interface between activated carbon and an electrolyte is disclosed. The high performance double layer capacitor includes a pair of aluminum impregnated carbon composite electrodes having an evenly distributed and continuous path of aluminum impregnated within an activated carbon fiber preform saturated with a high performance electrolytic solution. The high performance double layer capacitor is capable of delivering at least 5 Wh/kg of useful energy at power ratings of at least 600 W/kg.

  9. Microstructure and wear-resistance of laser clad TiC particle-reinforced coating

    NARCIS (Netherlands)

    Lei, T.C.; Ouyang, J.H.; Pei, Y.T.; Zhou, Y.

    A TiC-Ni alloy composite coating was clad to 1045 steel substrate using a 2kW CO2 laser. The microstructural constituents of the clad layer are found to be gamma-Ni and TiCp in the dendrites, and a fine eutectic of gamma-Ni plus (Fe, Cr)(23)C-6 in the interdendritic areas. Partial dissolution and

  10. Steam oxidation of Zr 1% Nb clads of VVER fuels in high temperature

    International Nuclear Information System (INIS)

    Solyanyj, V.I.; Bibilashvili, Yu.K.; Dranenko, V.V.; Levin, A.Ya.; Izrajlevskij, L.B.; Morozov, A.M.

    1984-01-01

    In a wide range of accident conditions processes of clad corrosion effected by steam are rather intensive and in many respects influence the safety of NPP and the after-accident dismantling of a reactor core. This paper discusses the results of comprehensive studies into corrosion behaviour of Zr 1%Nb clads of VVER-type fuels at high temperatures. These studies are a continuation of previous work and the base for the design modelling of corrosion processes

  11. Experience in quality assurance of alloy D9 clad tubes for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Kapoor, K.; Prahlad, B.

    2012-01-01

    Stainless Steel Alloy D9 is the material for cladding in various sub-assemblies of Prototype Fast Breeder Reactor (PFBR). The fabrication, inspection, testing and supply of the clad tubes for the first core of PFBR is nearly completed. The paper also compares the specification requirements and the achieved results for some of the critical aspects which is arrived after completing supply against the first core requirement

  12. Thermal stress intensity factor for an axial crack in a clad cylinder

    International Nuclear Information System (INIS)

    Kuo, An Yu; Deardorf, A.F.; Riccardella, P.C.

    1993-01-01

    Many clad pressure vessels have been found to have cracks running through the inside surface cladding and into the base material. Although Young's moduli and Poisson's ratios of the clad and base materials are about the same for most of the industrial applications, coefficients of thermal expansion of the two dissimilar materials, clad and base materials, are usually quite different. For example, low alloy ferritic steel is a common base material for reactor pressure vessels (RPV) and the vessels are usually clad with austenitic stainless steel. Young's moduli for the low alloy steel and stainless steel at 350 F are 29,000 ksi and 28,000 ksi, respectively, while their coefficients of thermal expansion are 7.47x10 -6 in/in and 9.50x10 -6 in/in-degree F, respectively. The mismatch in coefficients of thermal expansion will cause high residual thermal stress even when the entire vessel is at a uniform temperature. This residual stress is one of the primary reasons why so many cracks have been found in the cladded components. In performing reactor pressure vessel integrity evaluation, such as computing probability of brittle fracture of the RPV, it is necessary to calculate stress intensity factors for cracks, which initiate from the clad material and run into the base metal. This paper presents a convenient method of calculating stress intensity factor for an axial crack emanating from the inside surface of a cladded cylinder under thermal loading. A J-integral like line integral was derived and used to calculate the stress intensity factors from finite element stress solutions of the problem

  13. Investigation of likely causes of white patch formation on irradiated WWER fuel rod claddings

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Velioukhanov, V.P.; Ioltoukhovski, A.Y.; Pogodin, V.P.

    1999-01-01

    The information concerning white patches observed on fuel cladding surfaces has been analytically treated. The analysis shows at least three kinds of the white patch appearance: bright white spots which appear to be loose corrosion product deposits disclosing corrosion pits upon spalling; indistinct streaks with separate pronounced spots 1-2 in dia. The spots seem to be thin superficial deposits; light-coloured dense uniform crud distributed over the surface of fuel claddings and fuel assembly jackets. (author)

  14. Pellet-clad interaction observations in boiling water reactor fuel elements

    International Nuclear Information System (INIS)

    Sahoo, K.C.; Bahl, J.K.; Sivaramakrishnan, K.S.; Roy, P.R.

    1981-01-01

    Under a programme to assess the performance of fuel elements of Tarapur Atomic Power Station, post-irradiation examination has been carried out on 18 fuel elements in the first phase. Pellet-clad mechanical interaction behaviour in 14 elements with varying burnup and irradiation history has been studied using eddy current testing technique. The data has been analysed to evaluate the role of pellet-clad mechanical interaction in PCI/SCC failure in power reactor operating conditions. (author)

  15. Fuel-cladding chemical interaction correlation for mixed-oxide fuel pins

    International Nuclear Information System (INIS)

    Lawrence, L.A.

    1986-10-01

    A revised wastage correlation was developed for FCCI with fabrication and operating parameters. The expansion of the data base to 305 data sets provided sufficient data to employ normal statistical techniques for calculation of confidence levels without unduly penalizing predictions. The correlation based on 316 SS cladding also adequately accounts for limited measured depths of interaction for fuel pins with D9 and HTq cladding

  16. Analysis of mechanical tensile properties of irradiated and annealed RPV weld overlay cladding

    International Nuclear Information System (INIS)

    Novak, J.

    1993-01-01

    Mechanical tensile properties of irradiated and annealed outer layer of reactor pressure vessel weld overlay cladding, composed of Cr19Ni10Nb alloy, have been experimentally determined by conventional tensile testing and indentation testing. The constitutive properties of weld overlay cladding are then modelled with two homogenization models of the constitutive properties of elastic-plastic matrix-inclusion composites; numerical and experimental results are then compared. 10 refs., 4 figs., 4 tabs

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

    International Nuclear Information System (INIS)

    Kaminsky, M.

    1980-01-01

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

  18. The effect of zinc on the aluminum anode of the aluminum-air battery

    Science.gov (United States)

    Tang, Yougen; Lu, Lingbin; Roesky, Herbert W.; Wang, Laiwen; Huang, Baiyun

    Aluminum is an ideal material for batteries, due to its excellent electrochemical performance. Herein, the effect of zinc on the aluminum anode of the aluminum-air battery, as an additive for aluminum alloy and electrolytes, has been studied. The results show that zinc can decrease the anodic polarization, restrain the hydrogen evolution and increase the anodic utilization rate.

  19. 75 FR 70689 - Kaiser Aluminum Fabricated Products, LLC; Kaiser Aluminum-Greenwood Forge Division; Currently...

    Science.gov (United States)

    2010-11-18

    ... DEPARTMENT OF LABOR Employment and Training Administration [TA-W-70,376] Kaiser Aluminum Fabricated Products, LLC; Kaiser Aluminum- Greenwood Forge Division; Currently Known As Contech Forgings, LLC..., applicable to workers of Kaiser Aluminum Fabricated Products, LLC, Kaiser Aluminum-Greenwood Forge Division...

  20. Synthesis and Processing of Nanocrystalline Aluminum Nitride

    OpenAIRE

    Duarte, Matthew Albert

    2016-01-01

    Synthesis, processing and characterization of nanocrystalline aluminum nitride has been systematically studied. Non-carbon based gas nitridation was used to reduce nanocrystalline γ-alumina, having a grain size of ~80 nm. Single phase aluminum nitride powder was obtained at firing temperatures of 1200°C. Further processing of AlN powders was performed by CAPAD (Current Activated Pressure Assisted Densification) to obtain dense single phase aluminum nitride. Dense bulk aluminum nitride was ob...

  1. Mineral resource of the month: aluminum

    Science.gov (United States)

    Bray, E. Lee

    2012-01-01

    The article offers information on aluminum, a mineral resource which is described as the third-most abundant element in Earth's crust. According to the article, aluminum is the second-most used metal. Hans Christian Oersted, a Danish chemist, was the first to isolate aluminum in the laboratory. Aluminum is described as lightweight, corrosion-resistant and an excellent conductor of electricity and heat.

  2. Deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-09-01

    The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-02-15

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

  5. 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.

  6. A thermodynamic model for the attack behaviour in stainless steel clad oxide fuel pins

    International Nuclear Information System (INIS)

    Goetzmann, O.

    1979-01-01

    So far, post irradiation examination of burnt fuel pins has not revealed a clear cut picture of the cladding attack situation. For seemingly same conditions sometimes attack occurs, sometimes not. This model tries to depict the reaction possibilities along the inner cladding wall on the basis of thermodynamic facts in the fuel pin. It shows how the thermodynamic driving force for attack changes along the fuel column, and with different initial and operational conditions. Two criteria for attack are postulated: attack as a result of the direct reaction of reactive elements with cladding components; and attack as a result of the action of a special agent (CsOH). In defining a reaction potenial the oxygen potential, the temperature conditions (cladding temperature and fuel surface temperature), and the fission products are involved. For the determination of the oxygen potential at the cladding, three models for the redistribution of oxygen across the fuel/clad gap are offered. The effect of various parameters, like rod power, gap conductance, oxygen potential, inner wall temperature, on the thermodynamic potential for attack is analysed. (Auth.)

  7. Laser cladding of copper with molybdenum for wear resistance enhancement in electrical contacts

    International Nuclear Information System (INIS)

    Ng, K.W.; Man, H.C.; Cheng, F.T.; Yue, T.M.

    2007-01-01

    Laser cladding of Mo on Cu has been attempted with the aim of enhancing the wear resistance and hence increasing the service life of electrical contacts made of Cu. In order to overcome the difficulties arising from the large difference in thermal properties and the low mutual solubility between Cu and Mo, Ni was introduced as an intermediate layer between Mo and Cu. The Ni and Mo layers were laser clad one after the other to form a sandwich layer of Mo/Ni/Cu. Excellent bonding between the clad layer and the Cu substrate was ensured by strong metallurgical bonding. The hardness of the surface of the clad layer is seven times higher than that of the Cu substrate. Pin-on-disc wear tests consistently showed that the abrasive wear resistance of the clad layer was also improved by a factor of seven as compared with untreated Cu substrate. The specific electrical contact resistance of the clad surface was about 5.6 x 10 -7 Ω cm 2

  8. High temperature steam oxidation of zircaloy-2 cladding of PHWR fuel element

    International Nuclear Information System (INIS)

    Sethumadhavan, V.; Sathe, S.M.; Sunil Kumar; Khan, K.B.; Sah, D.N.

    1997-07-01

    In the event of a postulated loss of coolant accident (LOCA) the cladding surface of PHWR fuel element will be exposed to high temperature steam environment which may result in extensive oxidation and embrittlement of the cladding tube. High temperature steam oxidation has been studied on 40 mm long tubular samples of as fabricated zircaloy-2 cladding tubes of PHWR fuel. These studies were carried out in the temperature range 923K - 1073K for time durations ranging from 30 minutes to 4 hours and in temperature range 1323K -1423K for exposure time up to 30 minutes. The weight gain and the thickness of zirconium oxide and alpha layers have been measured in the exposed samples. Microstructure developed in the tubes due to high temperature exposure has been examined. Correlations have been derived from the experimental data for calculating oxygen uptake by the cladding during high temperature steam exposures. A diffusion based model has been developed to calculate the growth of oxide and oxygen stabilised alpha zirconium layers and distribution of oxygen in the cladding. Theoretical model has been used to predict the multilayer growth and oxygen distribution in the cladding at any given temperature for specified time duration. The results indicate that at a temperature of 923K zircaloy-2 oxidation follows a cubic rate law but at temperatures of 973K - 1073K and at 1323K - 1423K the oxidation is governed by parabolic rate law. Model calculations are in good agreement with experimental measurements

  9. Characterization of hard coatings produced by laser cladding using laser-induced breakdown spectroscopy technique

    International Nuclear Information System (INIS)

    Varela, J.A.; Amado, J.M.; Tobar, M.J.; Mateo, M.P.; Yañez, A.; Nicolas, G.

    2015-01-01

    Highlights: • Chemical mapping and profiling by laser-induced breakdown spectroscopy (LIBS) of coatings produced by laser cladding. • Production of laser clads using tungsten carbide (WC) and nickel based matrix (NiCrBSi) powders. • Calibration by LIBS of hardfacing alloys with different WC concentrations. - Abstract: Protective coatings with a high abrasive wear resistance can be obtained from powders by laser cladding technique, in order to extend the service life of some industrial components. In this work, laser clad layers of self-fluxing NiCrBSi alloy powder mixed with WC powder have been produced on stainless steel substrates of austenitic type (AISI 304) in a first step and then chemically characterized by laser-induced breakdown spectroscopy (LIBS) technique. With the suitable laser processing parameters (mainly output power, beam scan speed and flow rate) and powders mixture proportions between WC ceramics and NiCrBSi alloys, dense pore free layers have been obtained on single tracks and on large areas with overlapped tracks. The results achieved by LIBS technique and applied for the first time to the analysis of laser clads provided the chemical composition of the tungsten carbides in metal alloy matrix. Different measurement modes (multiple point analyses, depth profiles and chemical maps) have been employed, demonstrating the usefulness of LIBS technique for the characterization of laser clads based on hardfacing alloys. The behavior of hardness can be explained by LIBS maps which evidenced the partial dilution of some WC spheres in the coating

  10. Review and perspective: Sapphire optical fiber cladding development for harsh environment sensing

    Science.gov (United States)

    Chen, Hui; Buric, Michael; Ohodnicki, Paul R.; Nakano, Jinichiro; Liu, Bo; Chorpening, Benjamin T.

    2018-03-01

    The potential to use single-crystal sapphire optical fiber as an alternative to silica optical fibers for sensing in high-temperature, high-pressure, and chemically aggressive harsh environments has been recognized for several decades. A key technological barrier to the widespread deployment of harsh environment sensors constructed with sapphire optical fibers has been the lack of an optical cladding that is durable under these conditions. However, researchers have not yet succeeded in incorporating a high-temperature cladding process into the typical fabrication process for single-crystal sapphire fibers, which generally involves seed-initiated fiber growth from the molten oxide state. While a number of advances in fabrication of a cladding after fiber-growth have been made over the last four decades, none have successfully transitioned to a commercial manufacturing process. This paper reviews the various strategies and techniques for fabricating an optically clad sapphire fiber which have been proposed and explored in published research. The limitations of current approaches and future prospects for sapphire fiber cladding are discussed, including fabrication methods and materials. The aim is to provide an understanding of the past research into optical cladding of sapphire fibers and to assess possible material systems for future research on this challenging problem for harsh environment sensors.

  11. The deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

    International Nuclear Information System (INIS)

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

    1986-09-01

    The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail. (author)

  12. Thermal stress in the edge cladding of Nova glass laser disks

    International Nuclear Information System (INIS)

    Pitts, J.H.; Kong, M.K.; Gerhard, M.A.

    1987-01-01

    We calculated thermal stresses in Nova glass laser disks having light-absorbing edge cladding glass attached to the periphery with an epoxy adhesive. Our closed-form solutions indicated that, because the epoxy adhesive is only 25 μm across, it does not significantly affect the thermal stress in the disk or cladding glass. Our numerical results showed a peak tensile stress in the cladding glass of 24 MPa when the cladding glass had a uniform absorption coefficient of 7.5 cm -1 . This peak value is reduced to 19 MPa if surface parasitic oscillation heating is eliminated by tilting the disk edges. The peak tensile stresses exceed the typical 7 to 14-MPa working stress for glass; however, we have not observed any disk or cladding glass failures at peak Nova fluences of 20 J/cm 2 . We have observed delamination of the epoxy adhesive bond at fluences several times that which would occur on Nova. Replacement laser disks will incorporate cladding with a reduced absorption coefficient of 4.5 cm -1 . Recent experiments show that this reduced absorption coefficient is satisfactory

  13. Progress and Challenges of Ultrasonic Testing for Stress in Remanufacturing Laser Cladding Coating.

    Science.gov (United States)

    Yan, Xiao-Ling; Dong, Shi-Yun; Xu, Bin-Shi; Cao, Yong

    2018-02-13

    Stress in laser cladding coating is an important factor affecting the safe operation of remanufacturing components. Ultrasonic testing has become a popular approach in the nondestructive evaluation of stress, because it has the advantages of safety, nondestructiveness, and online detection. This paper provides a review of ultrasonic testing for stress in remanufacturing laser cladding coating. It summarizes the recent research outcomes on ultrasonic testing for stress, and analyzes the mechanism of ultrasonic testing for stress. Remanufacturing laser cladding coating shows typical anisotropic behaviors. The ultrasonic testing signal in laser cladding coating is influenced by many complex factors, such as microstructure, defect, temperature, and surface roughness, among others. At present, ultrasonic testing for stress in laser cladding coating can only be done roughly. This paper discusses the active mechanism of micro/macro factors in the reliability of stress measurement, as well as the impact of stress measurement on the quality and safety of remanufacturing components. Based on the discussion, this paper proposes strategies to nondestructively, rapidly, and accurately measure stress in remanufacturing laser cladding coating.

  14. Recycling of Aluminum from Fibre Metal Laminates

    NARCIS (Netherlands)

    Zhu, G.; Xiao, Y.; Yang, Y.; Wang, J.; Sun, B.; Boom, R.

    2012-01-01

    Recycling of aluminum alloy scrap obtained from delaminated fibre metal laminates (FMLs) was studied through high temperature refining in the presence of a salt flux. The aluminum alloy scrap contains approximately mass fraction w(Cu) = 4.4%, w(Mg) = 1.1% and w(Mn) = 0.6% (2024 aluminum alloy). The

  15. 21 CFR 182.1125 - Aluminum sulfate.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 3 2010-04-01 2009-04-01 true Aluminum sulfate. 182.1125 Section 182.1125 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) FOOD FOR... Substances § 182.1125 Aluminum sulfate. (a) Product. Aluminum sulfate. (b) Conditions of use. This substance...

  16. 75 FR 80527 - Aluminum Extrusions From China

    Science.gov (United States)

    2010-12-22

    ...)] Aluminum Extrusions From China AGENCY: United States International Trade Commission. ACTION: Scheduling of... of subsidized and less-than-fair-value imports from China of aluminum extrusions, primarily provided... contained in Aluminum Extrusions From the People's Republic of China: Notice of Preliminary Determination of...

  17. 21 CFR 73.2645 - Aluminum powder.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 1 2010-04-01 2010-04-01 false Aluminum powder. 73.2645 Section 73.2645 Food and... ADDITIVES EXEMPT FROM CERTIFICATION Cosmetics § 73.2645 Aluminum powder. (a) Identity and specifications. The color additive aluminum powder shall conform in identity and specifications to the requirements of...

  18. 21 CFR 582.1125 - Aluminum sulfate.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Aluminum sulfate. 582.1125 Section 582.1125 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL... Additives § 582.1125 Aluminum sulfate. (a) Product. Aluminum sulfate. (b) Conditions of use. This substance...

  19. 21 CFR 172.310 - Aluminum nicotinate.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 3 2010-04-01 2009-04-01 true Aluminum nicotinate. 172.310 Section 172.310 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) FOOD FOR... Special Dietary and Nutritional Additives § 172.310 Aluminum nicotinate. Aluminum nicotinate may be safely...

  20. Study of diffusion bond development in 6061 aluminum and its relationship to future high density fuels fabrication.

    Energy Technology Data Exchange (ETDEWEB)

    Prokofiev, I.; Wiencek, T.; McGann, D.

    1997-10-07

    Powder metallurgy dispersions of uranium alloys and silicides in an aluminum matrix have been developed by the RERTR program as a new generation of proliferation-resistant fuels. Testing is done with miniplate-type fuel plates to simulate standard fuel with cladding and matrix in plate-type configurations. In order to seal the dispersion fuel plates, a diffusion bond must exist between the aluminum coverplates surrounding the fuel meat. Four different variations in the standard method for roll-bonding 6061 aluminum were studied. They included mechanical cleaning, addition of a getter material, modifications to the standard chemical etching, and welding methods. Aluminum test pieces were subjected to a bend test after each rolling pass. Results, based on 400 samples, indicate that at least a 70% reduction in thickness is required to produce a diffusion bond using the standard rollbonding method versus a 60% reduction using the Type II method in which the assembly was welded 100% and contained open 9mm holes at frame corners.