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Sample records for neutralized cladding removal

  1. 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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  3. Measurement and removal of cladding light in high power fiber systems

    Science.gov (United States)

    Walbaum, Till; Liem, Andreas; Schreiber, Thomas; Eberhardt, Ramona; Tünnermann, Andreas

    2018-02-01

    The amount of cladding light is important to ensure longevity of high power fiber components. However, it is usually measured either by adding a cladding light stripper (and thus permanently modifying the fiber) or by using a pinhole to only transmit the core light (ignoring that there may be cladding mode content in the core area). We present a novel noninvasive method to measure the cladding light content in double-clad fibers based on extrapolation from a cladding region of constant average intensity. The method can be extended to general multi-layer radially symmetric fibers, e.g. to evaluate light content in refractive index pedestal structures. To effectively remove cladding light in high power systems, cladding light strippers are used. We show that the stripping efficiency can be significantly improved by bending the fiber in such a device and present respective experimental data. Measurements were performed with respect to the numerical aperture as well, showing the dependency of the CLS efficiency on the NA of the cladding light and implying that efficiency data cannot reliably be given for a certain fiber in general without regard to the properties of the guided light.

  4. Method for decontaminating stainless cladding tubes

    International Nuclear Information System (INIS)

    Komatsu, Fumiaki.

    1986-01-01

    Purpose: To form an oxide film over the surface of stainless cladding tubes and to efficiently remove radioactive materials from the steel surface together with the oxide layer by the use of an acid water solution. Method: After the removal of water from cladding tubes that have passed through the re-processing process, an oxide film is formed on the surface of the cladding tubes by heating over 400 deg C in an oxidizing atmosphere and thereafter washed again in an acid water solution. When the cladding tubes are thus oxidized once, the stainless base metal itself is oxidized, an oxide layer of several 10 μm or more being formed thereon. In consequence, since the oxide layer is far inferior in corrosion resistance to stainless metals, a pickling liquid easily penetrates into the stainless metal through the oxide layer, thereby remarkably promoting the peeling of the layer from the base metal surface and also improving the residual radioactive material removing efficiency together. (Takahashi, M.)

  5. 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.

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

    Science.gov (United States)

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

    1999-03-16

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

  7. Method of processing spent fuel cladding tubes

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi; Ouchi, Atsuhiro; Imahashi, Hiromichi.

    1986-01-01

    Purpose: To decrease the residual activity of spent fuel cladding tubes in a short period of time and enable safety storage with simple storage equipments. Constitution: Spent fuel cladding tubes made of zirconium alloys discharged from a nuclear fuel reprocessing step are exposed to a grain boundary embrittling atmosphere to cause grain boundary destruction. This causes grain boundary fractures to the zirconium crystal grains as the matrix of nuclear fuels and then precipitation products precipitated to the grain boundary fractures are removed. The zirconium constituting the nuclear fuel cladding tube and other ingredient elements contained in the precipitation products are separated in this removing step and they are separately stored respectively. As a result, zirconium constituting most part of the composition of the spent nuclear fuel cladding tubes can be stored safely at a low activity level. (Takahashi, M.)

  8. Method for automatic filling of nuclear fuel rod cladding tubes

    International Nuclear Information System (INIS)

    Bezold, H.

    1979-01-01

    Prior to welding the zirconium alloy cladding tubes with end caps, they are automatically filled with nuclear fuel tablets and ceramic insulating tablets. The tablets are introduced into magazine drums and led through a drying oven to a discharging station. The empty cladding tubes are removed from this discharging station and filled with tablets. A filling stamp pushes out the columns of tablets in the magazine tubes of the magazine drum into the cladding tube. Weight and measurement of length determine the filled state of the cladding tube. The cladding tubes are then led to the welding station via a conveyor belt. (DG) [de

  9. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    International Nuclear Information System (INIS)

    Rudisill, T; John Mickalonis, J

    2006-01-01

    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 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 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 4 F)/ammonium nitrate (NH 4 NO 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 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 4 ) 2 ZrF 6 ) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of

  10. Penetrate-leach dissolution of zirconium-clad uranium and uranium dioxide fuels

    International Nuclear Information System (INIS)

    Harmon, H.D.

    1975-01-01

    A new decladding-dissolution process was developed for zirconium-clad uranium metal and UO 2 fuels. The proposed penetrate-leach process consists of penetrating the zirconium cladding with Alniflex solution (2M HF--1M HNO 3 --1M Al(NO 3 ) 3 --0.1M K 2 Cr 2 O 7 ) and of leaching the exposed core with 10M HNO 3 . Undissolved cladding pieces are discarded as solid waste. Periodic HF and HNO 3 additions, efficient agitation, and in-line zirconium analyses are required for successful control of ZrF 4 and/or AlF 3 precipitation during the cladding-penetration step. Preliminary solvent extraction studies indicated complete recovery of uranium with 30 vol. percent tributyl phosphate (TBP) from both Alniflex solution and blended Alniflex-HNO 3 leach solutions. With 7.5 vol. percent TBP, high extractant/feed flow ratios and low scrub flows are required for satisfactory uranium recovery from Alniflex solution. Modified waste-handling procedures may be required for Alniflex waste, because it cannot be evaporated before neutralization and large quantities of solids are generated on neutralization. The effect of unstable UZr 3 (epsilon phase of uranium-zirconium system) on the safety of penetrate-leach dissolution was investigated

  11. Treatment of stainless steels and zircaloy cladding hulls

    International Nuclear Information System (INIS)

    Jenkins, I.L.; Taylor, R.F.

    1978-01-01

    Results are reported on the fissile material content and the distribution of alpha and beta-gamma emitters in both types of cladding. Apart from very small amounts of residual fuel, fissile material is present as a deposit formed during the dissolution of fuel and also as material driven into the cladding by fission recoil. Alpha-emitters penetrate to depths of 1-2 μm into both S.S. and Zircaloy claddings. The surface deposits on individual hulls can be effectively removed by refluxing with nitric acid or by cleaning with nitric acid in an ultrasonic bath. The physical structural and handling behavior of hull assemblies are examined as being of key importance to the establishment of an efficient cleaning process. The reference leaching target is to extract residual fuel fragments and to remove surface deposits. Preferred routes for compaction, drumming, and encapsulation are briefly reviewed with regard to achieving a final package volume half that of the original hulls with associated hardware

  12. Underwater cladding with laser beam and plasma arc welding

    International Nuclear Information System (INIS)

    White, R.A.; Fusaro, R.; Jones, M.G.; Solomon, H.D.; Milian-Rodriguez, R.R.

    1997-01-01

    Two welding processes, plasma arc (transferred arc) (PTA) and laser beam, were investigated to apply cladding to austenitic stainless steels and Inconel 600. These processes have long been used to apply cladding layers , but the novel feature being reported here is that these cladding layers were applied underwater, with a water pressure equivalent to 24 m (80 ft). Being able to apply the cladding underwater is very important for many applications, including the construction of off-shore oil platforms and the repair of nuclear reactors. In the latter case, being able to weld underwater eliminates the need for draining the reactor and removing the fuel. Welding underwater in reactors presents numerous challenges, but the ability to weld without having to drain the reactor and remove the fuel provides a huge cost savings. Welding underwater in reactors must be done remotely, but because of the radioactive corrosion products and neutron activation of the steels, remote welding would also be required even if the reactor is drained and the fuel removed. In fact, without the shielding of the water, the remote welding required if the reactor is drained might be even more difficult than that required with underwater welds. Furthermore, as shall be shown, the underwater welds that the authors have made were of high quality and exhibit compressive rather than tensile residual stresses

  13. Fuel removing method for high burnup fuel and device therefor

    International Nuclear Information System (INIS)

    Terakado, Shogo; Owada, Isao; Kanno, Yoshio; Aizawa, Sakue; Yamahara, Takeshi.

    1993-01-01

    A through hole is perforated at the center of a fuel rod in a cladding tube by a diamond drill in a water vessel. Further, the through hole is enlarged by the diamond drill. A pellet removing tool is attached to a drill chuck instead of the diamond drill. Then, the thin cylindrical fuel pellet remaining on the inner surface of the cladding tube is removed by using a pellet removing tool while applying vibrations. Subsequently, a wire brush having a slightly larger diameter than that of the inner diameter of the cladding tube is attached to the drill chuck and rotated to finish the inner surface, so that a small amount of pellets remained on the inner surface of the cladding tube is removed. Pellet powders in the water vessel are collected and recovered to the water container. This can remove high burnup fuels which are firmly sticked to the cladding tube, without giving thermal or mechanical influences on the cladding tube. (I.N.)

  14. Removal of boron from aqueous solution by using neutralized red mud

    International Nuclear Information System (INIS)

    Cengeloglu, Yunus; Tor, Ali; Arslan, Gulsin; Ersoz, Mustafa; Gezgin, Sait

    2007-01-01

    The adsorptive removal of boron from aqueous solution by using the neutralized red mud was studied in batch equilibration technique. The effects of pH, adsorbent dosage, initial boron concentration and contact time on the adsorption were investigated. The experiments demonstrated that boron removal was of a little fluctuation in pH range of 2-7 and it takes 20 min to attain equilibrium. The adsorption data was analyzed using the Langmuir and the Freundlich isotherm models and it was found that the Freundlich isotherm model represented the measured sorption data well

  15. Critical cladding radius for hybrid cladding modes

    Science.gov (United States)

    Guyard, Romain; Leduc, Dominique; Lupi, Cyril; Lecieux, Yann

    2018-05-01

    In this article we explore some properties of the cladding modes guided by a step-index optical fiber. We show that the hybrid modes can be grouped by pairs and that it exists a critical cladding radius for which the modes of a pair share the same electromagnetic structure. We propose a robust method to determine the critical cladding radius and use it to perform a statistical study on the influence of the characteristics of the fiber on the critical cladding radius. Finally we show the importance of the critical cladding radius with respect to the coupling coefficient between the core mode and the cladding modes inside a long period grating.

  16. Residual stresses in weld-clad reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Bertram, W.

    1975-01-01

    Cladding of low alloy nuclear reactor pressure vessel steel with austenitic stainless steel introduces in heavy section components high residual stresses which may cause microcrack formation in stress relief heat treatment. In this investigation an attempt is made to contribute to the solution of the stress relief cracking problem by determining quantitatively the magnitude and distribution of the residual stresses after cladding and after subsequent stress relief heat treatment. The distribution of residual stresses was determined on the basis of a combined experimental-mathematical procedure. Heavy section plate specimens of low alloy steel as base material were given an austenitic monolayer-cladding using the techniques of strip electrode and plasma hot wire cladding, respectively. A number of plates was stress relief heat treated. Starting from the cladded surface the thickness of the plates was reduced by subsequent removal of layers of material. The elastic strain reaction to the removal of each layer was measured by strain gauges. From the data obtained the biaxial residual stress distribution was computed as a function of thickness using relations which are derived for this particular case. In summary, lower residual stresses are caused by reduced thickness of the components. As the heat input, is decreased at identical base material thickness, the residual stresses are lowered also. The height of the tensile residual stress peak, however, remains approximataly constant. In stress relief annealed condition the residual stresses in the cladding are in tension; in the base material the residual stresses are negligibly small

  17. Dynamics of a BWR with inclusion of boiling nonlinearity, clad temperature and void-dependent core power removal: Stability and bifurcation characteristics of advanced heavy water reactor (AHWR)

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Dinkar, E-mail: dinkar@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology Kanpur, Kanpur 208 016 (India); Kalra, Manjeet Singh, E-mail: drmanjeet.singh@dituniversity.edu.in [DIT University, Dehradun 248 009 (India); Wahi, Pankaj, E-mail: wahi@iitk.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Kanpur, Kanpur 208 016 (India)

    2016-11-15

    Highlights: • Simplified models with inclusion of the clad temperature are considered. • Boiling nonlinearity and core power removal have been modeled. • Method of multiple time scales has been used for nonlinear analysis to get the nature and amplitude of oscillations. • Incorporation of modeling complexities enhances the stability of system. • We find that reactors with higher nominal power are more desirable from the point of view of global stability. - Abstract: We study the effect of including boiling nonlinearity, clad temperature and void-dependent power removal from the primary loop in the mathematical modeling of a boiling water reactor (BWR) on its dynamic characteristics. The advanced heavy water reactor (AHWR) is taken as a case study. Towards this end, we have analyzed two different simplified models with different handling of the clad temperature. Each of these models has the necessary modifications pertaining to boiling nonlinearity and power removal from the primary loop. These simplified models incorporate the neutronics and thermal–hydraulic coupling. The effect of successive changes in the modeling assumptions on the linear stability of the reactor has been studied and we find that incorporation of each of these complexities in the model increases the stable operating region of the reactor. Further, the method of multiple time scales (MMTS) is exploited to carry out the nonlinear analysis with a view to predict the bifurcation characteristics of the reactor. Both subcritical and supercritical Hopf bifurcations are present in each model depending on the choice of operating parameters. These analytical observations from MMTS have been verified against numerical simulations. A parametric study on the effect of changing the nominal reactor power on the regions in the parametric space of void coefficient of reactivity and fuel temperature coefficient of reactivity with sub- and super-critical Hopf bifurcations has been performed for all

  18. A process to remove ammonia from PUREX plant effluents

    International Nuclear Information System (INIS)

    Moore, J.D.

    1990-01-01

    Zirconium-clad nuclear fuel from the Hanford N-Reactor is reprocessed in the PUREX (Plutonium Uranium Extraction) Plant operated by the Westinghouse Hanford Comapny. Before dissolution, cladding is chemically removed from the fuel elements with a solution of ammonium fluoride-ammonium nitrate (AFAN). a solution batch with an ammonia equivalent of about 1,100 kg is added to each fuel batch of 10 metric tons. This paper reports on this decladding process, know as the 'Zirflex' process which produces waste streams containing ammonia and ammonium slats. Waste stream treatment, includes ammonia scrubbing, scrub solution evaporation, residual solids dissolution, and chemical neutralization. These processes produce secondary liquid and gaseous waste streams containing varying concentrations of ammonia and low-level concentrations of radionuclides. Until legislative restrictions were imposed in 1987, these secondary streams were released to the soil in a liquid disposal 'crib' and to the atmosphere

  19. Second phase precipitation in irradiated Type 316 stainless steel cladding

    International Nuclear Information System (INIS)

    Hales, J.W.

    1978-05-01

    Differences in the phase composition of FFTF fuel cans following irradiation in the General Electric Test Reactor compared to HEDL fuel cans prompted laboratory studies to be conducted using cladding from the same lots of material used to fabricate the fuel pins and on cladding sections removed from the plenum area of the irradiated fuel pins to help establish the cause of the observed differences

  20. Structural cladding /clad structures

    DEFF Research Database (Denmark)

    Beim, Anne

    2012-01-01

    Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure in the pr......Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure...... to analyze, compare, and discuss how these various construction solutions point out strategies for development based on fundamentally different mindsets. The research questions address the following issues: How to learn from traditional construction principles: When do we see limitations of tectonic maneuver......, to ask for more restrictive building codes. As an example, in Denmark there are series of increasing demands in the current building legislations that are focused at enhancing the energy performance of buildings, which consequently foster rigid insulation standards and ask for improvement of air...

  1. Neutral Color Semitransparent Microstructured Perovskite Solar Cells

    KAUST Repository

    Eperon, Giles E.; Burlakov, Victor M.; Goriely, Alain; Snaith, Henry J.

    2014-01-01

    Neutral-colored semitransparent solar cells are commercially desired to integrate solar cells into the windows and cladding of buildings and automotive applications. Here, we report the use of morphological control of perovskite thin films to form

  2. Maintaining neutral sagittal cervical alignment after football helmet removal during emergency spine injury management.

    Science.gov (United States)

    Decoster, Laura C; Burns, Matthew F; Swartz, Erik E; Murthi, Dinakar S; Hernandez, Adam E; Vailas, James C; Isham, Linda L

    2012-04-15

    Descriptive laboratory study. To determine whether the placement of padding beneath the occiput after helmet removal is an effective intervention to maintain neutral sagittal cervical spine alignment in a position comparable with the helmeted condition. Current on-field recommendations for managing football athletes with suspected cervical spine injuries call for face mask removal, rather than helmet removal, because the combination of helmet and shoulder pads has been shown to maintain neutral cervical alignment. Therefore, in cases when helmet removal is required, recommendations also call for shoulder pad removal. Because removal of equipment causes motion, any technique that postpones the need to remove the shoulder pads would reduce prehospital motion. Four lateral radiographs of 20 male participants were obtained (age = 23.6 ± 2.7 years). Radiographs of participants wearing shoulder pads and helmet were first obtained. The helmet was removed and radiographs of participants with occipital padding were obtained immediately and 20 minutes later and finally without occipital padding. Cobb angle measurements for C2-C6 vertebral segments were determined by an orthopedic spine surgeon blinded to the study's purpose. Intraobserver reliability was determined using intraclass coefficient analysis. Measurements were analyzed using a 1×4 repeated-measures analysis of variance and post hoc pairwise comparisons with Bonferroni correction. Intraobserver analysis showed excellent reliability (intraclass correlation = 1.0; 95% confidence interval [CI], 0.999-1.0). Repeated-measures analysis of variance detected significant differences (F(3,17) = 13.34; P football helmet in the field, occipital padding (along with full body/head immobilization techniques) may be used to limit cervical lordosis, allowing safe delay of shoulder pad removal.

  3. 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...

  4. Estimation of penetration depth of fission products in cladding Hull

    International Nuclear Information System (INIS)

    Kim, Hee Moon; Jung, Yang Hong; Yoo, Byong Ok; Choo, Yong Sun; Hong, Kwon Pyo

    2005-01-01

    A disposal and a reprocessing for spent fuel rod with high burnup need de-cladding procedure. Pellet in this rod has been separated from a cladding hull to reduce a radioactivity of hull by chemical and mechanical methods. But fission products and actinides(U,Pu) still remain inside of cladding hull by chemical bonding and fission spike, which is called as 'contamination'. More specific removal of this contamination would have been considered. In this study, the sorts of fission products and penetration depth in hull were observed by EPMA test. To analyze this behavior, SRIM 2000 code was also used as energies of fission products and an oxide thickness of hull

  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. YAG laser cladding to heat exchanger flange in actual plant

    International Nuclear Information System (INIS)

    Toshio, Kojima

    2001-01-01

    This paper is a sequel to ''Development of YAG Laser Cladding Technology to Heat Exchanger Flange'' presented in ICONE-8. A YAG Laser cladding technology is a permanent repairing and preventive maintenance method for heat exchanger's flange (channel side) seating surface which is degraded by the corrosion in long term operation. The material of this flange is carbon steel, and that of cladding wire is type 316 stainless steel so as to have high corrosion resistance. In former paper above, the soundness of cladding layers were presented to be verified. This channel side flange is bolted with tube sheet (shell side) through metal gasket. As the tube sheet side is already cladded a corrosion resistant material, it needs to apply the repairing and preventive maintenance method to only channel side. In 2000 this technology had been performed to the actual heat exchanger (Residual Heat Removal Heat Exchanger; RHR Hx) flange in domestic nuclear power plant. This paper described the outline, special equipment, and our total evaluation for this actual laser cladding work. And also several technical subjects which we should solve and/or improve for the next project was presented. (author)

  7. Process for surface treatment of zirconium-containing cladding materials for fuel element or other components for nuclear reactors

    International Nuclear Information System (INIS)

    Videm, K.G.; Lunde, L.R.; Kooyman, H.H.

    1975-01-01

    A process for the surface treatment of zirconium-base cladding materials for fuel elements or other components for nuclear reactors is described. The treatment includes pickling the cladding material in a fluoride-containing bath, and then applying a protective coating through oxidation to the pickled cladding material. The fluoride-containing contaminants which remain on the surface of the cladding material during pickling are removed or rendered harmless by anodic oxidation

  8. Two-Channel SPR Sensor Combined Application of Polymer- and Vitreous-Clad Optic Fibers

    Directory of Open Access Journals (Sweden)

    Yong Wei

    2017-12-01

    Full Text Available By combining a polymer-clad optic fiber and a vitreous-clad optic fiber, we proposed and fabricated a novel optic fiber surface plasmon resonance (SPR sensor to conduct two-channel sensing at the same detection area. The traditional optic fiber SPR sensor has many disadvantages; for example, removing the cladding requires corrosion, operating it is dangerous, adjusting the dynamic response range is hard, and producing different resonance wavelengths in the sensing area to realize a multi-channel measurement is difficult. Therefore, in this paper, we skillfully used bare fiber grinding technology and reverse symmetry welding technology to remove the cladding in a multi-mode fiber and expose the evanescent field. On the basis of investigating the effect of the grinding angle on the dynamic range change of the SPR resonance valley wavelength and sensitivity, we combined polymer-clad fiber and vitreous-clad fiber by a smart design structure to realize at a single point a two-channel measurement fiber SPR sensor. In this paper, we obtained a beautiful spectral curve from a multi-mode fiber two-channel SPR sensor. In the detection range of the refractive rate between 1.333 RIU and 1.385 RIU, the resonance valley wavelength of channel Ⅰ shifted from 622 nm to 724 nm with a mean average sensitivity of 1961 nm/RIU and the resonance valley wavelength of channel Ⅱ shifted from 741 nm to 976 nm with a mean average sensitivity of 4519 nm/RIU.

  9. Two-Channel SPR Sensor Combined Application of Polymer- and Vitreous-Clad Optic Fibers.

    Science.gov (United States)

    Wei, Yong; Su, Yudong; Liu, Chunlan; Nie, Xiangfei; Liu, Zhihai; Zhang, Yu; Zhang, Yonghui

    2017-12-09

    By combining a polymer-clad optic fiber and a vitreous-clad optic fiber, we proposed and fabricated a novel optic fiber surface plasmon resonance (SPR) sensor to conduct two-channel sensing at the same detection area. The traditional optic fiber SPR sensor has many disadvantages; for example, removing the cladding requires corrosion, operating it is dangerous, adjusting the dynamic response range is hard, and producing different resonance wavelengths in the sensing area to realize a multi-channel measurement is difficult. Therefore, in this paper, we skillfully used bare fiber grinding technology and reverse symmetry welding technology to remove the cladding in a multi-mode fiber and expose the evanescent field. On the basis of investigating the effect of the grinding angle on the dynamic range change of the SPR resonance valley wavelength and sensitivity, we combined polymer-clad fiber and vitreous-clad fiber by a smart design structure to realize at a single point a two-channel measurement fiber SPR sensor. In this paper, we obtained a beautiful spectral curve from a multi-mode fiber two-channel SPR sensor. In the detection range of the refractive rate between 1.333 RIU and 1.385 RIU, the resonance valley wavelength of channel Ⅰ shifted from 622 nm to 724 nm with a mean average sensitivity of 1961 nm/RIU and the resonance valley wavelength of channel Ⅱ shifted from 741 nm to 976 nm with a mean average sensitivity of 4519 nm/RIU.

  10. Improved Background Removal in Sounding Rocket Neutral Atom Imaging Data

    Science.gov (United States)

    Smith, M. R.; Rowland, D. E.

    2017-12-01

    The VISIONS sounding rocket, launched into a substorm on Feb 7, 2013 from Poker Flat, Alaska had a novel miniaturized energetic neutral atom (ENA) imager onboard. We present further analysis of the ENA data from this rocket flight, including improved removal of ultraviolet and electron contamination. In particular, the relative error source contributions due to geocoronal, auroral, and airglow UV, as well as energetic electrons from 10 eV to 3 keV were assessed. The resulting data provide a more clear understanding of the spatial and temporal variations of the ion populations that are energized to tens or hundreds of eV.

  11. Effect of laser power on clad metal in laser-TIG combined metal cladding

    Science.gov (United States)

    Utsumi, Akihiro; Hino, Takanori; Matsuda, Jun; Tasoda, Takashi; Yoneda, Masafumi; Katsumura, Munehide; Yano, Tetsuo; Araki, Takao

    2003-03-01

    TIG arc welding has been used to date as a method for clad welding of white metal as bearing material. We propose a new clad welding process that combines a CO2 laser and a TIG arc, as a method for cladding at high speed. We hypothesized that this method would permit appropriate control of the melted quantity of base metal by varying the laser power. We carried out cladding while varying the laser power, and investigated the structure near the boundary between the clad layer and the base metal. Using the laser-TIG combined cladding, we found we were able to control appropriately the degree of dilution with the base metal. By applying this result to subsequent cladding, we were able to obtain a clad layer of high quality, which was slightly diluted with the base metal.

  12. Effects of non-uniform core flow on peak cladding temperature: MOXY/SCORE sensitivity calculations

    Energy Technology Data Exchange (ETDEWEB)

    Chang, S.C.

    1979-08-15

    The MOXY/SCORE computer program is used to evaluate the potential effect on peak cladding temperature of selective cooling that may result from a nonuniform mass flux at the core boundaries during the blowdown phase of the LOFT L2-4 test. The results of this study indicate that the effect of the flow nonuniformity at the core boundaries will be neutralized by a strong radial flow redistribution in the neighborhood of core boundaries. The implication is that the flow nonuniformity at the core boundaries has no significant effect on the thermal-hydraulic behavior and cladding temperature at the hot plane.

  13. Effects of non-uniform core flow on peak cladding temperature: MOXY/SCORE sensitivity calculations

    International Nuclear Information System (INIS)

    Chang, S.C.

    1979-01-01

    The MOXY/SCORE computer program is used to evaluate the potential effect on peak cladding temperature of selective cooling that may result from a nonuniform mass flux at the core boundaries during the blowdown phase of the LOFT L2-4 test. The results of this study indicate that the effect of the flow nonuniformity at the core boundaries will be neutralized by a strong radial flow redistribution in the neighborhood of core boundaries. The implication is that the flow nonuniformity at the core boundaries has no significant effect on the thermal-hydraulic behavior and cladding temperature at the hot plane

  14. 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.

  15. 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

  16. Laser cladding of turbine blades

    International Nuclear Information System (INIS)

    Shepeleva, L.; Medres, B.; Kaplan, W.D.; Bamberger, M.

    2000-01-01

    A comparative study of two different techniques for the application of wear-resistant coatings for contact surfaces of shroud shelves of gas turbine engine blades (GTE) has been conducted. Wear-resistant coatings were applied on In713 by laser cladding with direct injection of the cladding powder into the melt pool. Laser cladding was conducted with a TRUMPF-2500, CW-CO 2 laser. The laser cladding was compared with commercially available plasma cladding with wire. Both plasma and laser cladded zones were characterized by optical and scanning electron microscopy. It was found that the laser cladded zone has a higher microhardness value (650-820 HV) compared with that of the plasma treated material (420-440 HV). This is a result of the significant reduction in grain size in the case of laser cladding. Unlike the plasma cladded zones, the laser treated material is free of micropores and microcracks. (orig.)

  17. 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.

  18. Removal method of radium in mine water by filter sand

    International Nuclear Information System (INIS)

    Taki, Tomihiro; Naganuma, Masaki

    2003-01-01

    Trace radium is contained in mine water from the old mine road in Ningyo-Toge Environmental Engineering Center, JNC. We observed that filter sand with hydrated manganese oxide adsorbed radium in the mine water safely for long time. The removal method of radium by filter sand cladding with hydrated manganese oxide was studied. The results showed that radium was removed continuously and last for a long time from mine water with sodium hypochlorite solution by passing through the filter sand cladding with hydrated manganese. Only sodium hypochlorite solution was used. When excess of it was added, residue chlorine was used as chlorine disinfection. Filter sand cladding with hydrated manganese on the market can remove radium in the mine water. The removal efficiency of radium is the same as the radium coprecipitation method added with barium chloride. The cost is much lower than the ordinary methods. Amount of waste decreased to about 1/20 of the coprecipitation method. (S.Y.)

  19. Widely tunable asymmetric long-period fiber grating with high sensitivity using optical polymer on laser-ablated cladding.

    Science.gov (United States)

    Chen, Nan-Kuang; Hsu, Der-Yi; Chi, Sien

    2007-08-01

    We demonstrate high-efficiency, wideband-tunable, laser-ablated long-period fiber gratings that use an optical polymer overlay. Portions of the fiber cladding are periodically removed by CO(2) laser pulses to induce periodic index changes for coupling the core mode into cladding modes. An optical polymer with a high thermo-optic coefficient with a dispersion distinct from that of silica is used on a deep-ablated cladding structure so that the effective indices of cladding modes become dispersive and the resonant wavelengths can be efficiently tuned. The tuning efficiency can be as high as 15.8 nm/ degrees C, and the tuning range can be wider than 105 nm (1545-1650 nm).

  20. Automatic welding and cladding in heavy fabrication

    International Nuclear Information System (INIS)

    Altamer, A. de

    1980-01-01

    A description is given of the automatic welding processes used by an Italian fabricator of pressure vessels for petrochemical and nuclear plant. The automatic submerged arc welding, submerged arc strip cladding, pulsed TIG, hot wire TIG and MIG welding processes have proved satisfactory in terms of process reliability, metal deposition rate, and cost effectiveness for low alloy and carbon steels. An example shows sequences required during automatic butt welding, including heat treatments. Factors which govern satisfactory automatic welding include automatic anti-drift rotator device, electrode guidance and bead programming system, the capability of single and dual head operation, flux recovery and slag removal systems, operator environment and controls, maintaining continuity of welding and automatic reverse side grinding. Automatic welding is used for: joining vessel sections; joining tubes to tubeplate; cladding of vessel rings and tubes, dished ends and extruded nozzles; nozzle to shell and butt welds, including narrow gap welding. (author)

  1. Comparison of corrosion behavior between fusion cladded and explosive cladded Inconel 625/plain carbon steel bimetal plates

    International Nuclear Information System (INIS)

    Zareie Rajani, H.R.; Akbari Mousavi, S.A.A.; Madani Sani, F.

    2013-01-01

    Highlights: ► Both explosive and fusion cladding aggravate the corrosion resistance of Inconel 625. ► Fusion cladding is more detrimental to nonuniform corrosion resistance. ► Single-layered fusion coat does not show any repassivation ability. ► Adding more layers enhance the corrosion resistance of fusion cladding Inconel 625. ► High impact energy spoils the corrosion resistance of explosive cladding Inconel 625. -- Abstract: One of the main concerns in cladding Inconel 625 superalloy on desired substrates is deterioration of corrosion resistance due to cladding process. The present study aims to compare the effect of fusion cladding and explosive cladding procedures on corrosion behavior of Inconel 625 cladding on plain carbon steel as substrate. Also, an attempt has been made to investigate the role of load ratio and numbers of fusion layers in corrosion behavior of explosive and fusion cladding Inconel 625 respectively. In all cases, the cyclic polarization as an electrochemical method has been applied to assess the corrosion behavior. According to the obtained results, both cladding methods aggravate the corrosion resistance of Inconel 625. However, the fusion cladding process is more detrimental to nonuniform corrosion resistance, where the chemical nonuniformity of fusion cladding superalloy issuing from microsegregation, development of secondary phases and contamination of clad through dilution hinders formation of a stable passive layer. Moreover, it is observed that adding more fusion layers can enhance the nonuniform corrosion resistance of fusion cladding Inconel 625, though this resistance still remains weaker than explosive cladding superalloy. Also, the results indicate that raising the impact energy in explosive cladding procedure drops the corrosion resistance of Inconel 625.

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  3. Corrosion characteristics of K-claddings

    International Nuclear Information System (INIS)

    Park, J. Y.; Choi, B. K.; Jung, Y. H.; Jung, Y. H.

    2004-01-01

    The Improvement of the corrosion resistance of nuclear fuel claddings is the critical issue for the successful development of the high burn-up fuel. KAERI have developed the K-claddings having a superior corrosion resistance by controlling the alloying element addition and optimizing the manufacturing process. The comparative evaluation of the corrosion resistance for K-claddings and the foreign claddings was performed and the effect of the heat treatment on the corrosion behavior of K-claddings was also examined. Corrosion tests were carried out in the conditions of 360 .deg. C pure water, PWR-simulating loop and 400 .deg. C steam, From the results of the corrosion tests, it was found that the corrosion resistance of K-claddings is superior to those of Zry4 and A claddings and K6 showed a better corrosion resistance than K3. The corrosion behavior of K-cladding was strongly influenced by the final annealing rather than the intermediate annealing, and the corrosion resistance increased with decreasing the final annealing temperature

  4. CASTI handbook of cladding technology. 2. ed.

    International Nuclear Information System (INIS)

    Smith, L.; Celant, M.

    2000-01-01

    This updated (2000) CASTI handbook covers all aspects of clad products - the different means of manufacture, properties and applications in various industries. Topics include: an introduction to cladding technology, clad plate, clad pipes, bends, clad fittings, specification requirements of clad products, welding clad products, clad product application and case histories from around the world. Unique to this book is the documentation of case histories of major cladding projects from around the world and how the technology of that day has withstood the demands of time. Filled with over 100 photos and graphics illustrating the various cladding technology examples and products, this book truly documents the most recent technologies in the field of cladding technology used worldwide

  5. Development and fabrication of seamless Aluminium finned clad tubes for metallic uranium fuel rods for research reactor

    International Nuclear Information System (INIS)

    Singh, A.K.; Hussain, M.M.; Jayachandran, N.K.; Abdulla, K.K.

    2012-01-01

    Natural uranium metal or its alloy is used as fuel in nuclear reactors. Usually fuel is clad with compatible material to prevent its direct contact with coolant which prevents spread of activity. One of the methods of producing fuel for nuclear reactor is by co-drawing finished uranium rods with aluminum clad tube to develop intimate contact for effective heat removal during reactor operation. Presently seam welded Aluminium tubes are used as clad for Research Reactor fuel. The paper will highlight entire fabrication process followed for the fabrication of seamless Aluminium finned tubes along with relevant characterisation results

  6. Rheological evaluation of simulated neutralized current acid waste - transuranics

    International Nuclear Information System (INIS)

    Fow, C.L.; McCarthy, D.; Thornton, G.T.; Scott, P.A.; Bray, L.A.

    1986-09-01

    At the Hanford Plutonium and Uranium Extraction Plant (PUREX), in Richland, Washington, plutonium and uranium products are recovered from irradiated fuel by a solvent extraction process. A byproduct of this process is an aqueous waste stream that contains fission products. This waste stream, called current acid waste (CAW), is chemically neutralized and stored in double shell tanks (DSTs) on the Hanford Site. This neutralized current acid waste (NCAW) will be transported by pipe to B-Plant, a processing plant located nearby. In B-Plant, the transuranic (TRU) elements in NCAW are separated from the non-TRU elements. The majority of the TRU elements in NCAW are in the solids. Therefore, the primary processing operation is to separate the NCAW solids (NCAW-TRU) from the NCAW liquid. These two waste streams will be pumped to suitable holding tanks before being further processed for permanent disposal. To ensure that the retrieval and transportation of NCAW and NCAW-TRU are successful, researchers at Pacific Northwest Laboratory (PNL) evaluated the rheological and transport properties of the slurries. This evaluation had two phases. First, researchers conducted laboratory rheological evaluations of simulated NCAW and NCAW-TRU. The results of these evaluations were then correlated with classical rheological models and scaled up to predict the performance that is likely to occur in the full-scale system. This scale-up procedure has already been successfully used to predict the critical transport properties of a slurry (Neutralized Cladding Removal Waste) with rheological properties similar to those displayed by NCAW and NCAW-TRU

  7. Arisings of cladding wastes from nuclear fuel in the European Community

    International Nuclear Information System (INIS)

    Cottone, G.

    1978-01-01

    An inquiry has been made in the member states on composition, activation and amounts of cladding wastes arising in the European Community until 1990 from the following reactor types: BWR, PWR, SGHWR, AGR and FBR. The elaborated results of this inquiry are given in this report. On the basis of forecasted reprocessing capacities the cumulative amount of cladding waste in the Community was estimated to reach in 1985 and 1990, respectively, about 2,100 and 7,300 metric tons. This waste will mainly consist of zircaloy and of smaller amounts of stainless steel and nickel alloy. Assuming that 0.5% of the spent fuel remains with the cladding, the contamination has been estimated for cooling times varying from 1 to 1000 years. In the first centuries activation is prevailing, but contamination determines the long-term radioactivity; consequently better decontamination, removing the alpha-bearing compounds, would be beneficial in reducing the long term hazard

  8. About criteria of inadmissible embrittlement of zirconium fuel cladding during LOCA in the PWRs

    International Nuclear Information System (INIS)

    Osmachkin, V.S.

    1999-01-01

    According the licensing procedures the designers of the PWRs reactor have to prove the meeting of special safety requirements. One criteria on effectiveness of the Emergency Core Cooling System is not to exceeding some limited conditions of the fuel cladding during LOCA accidents (typical example T m ax o C, ECR<0,17 and oth.). The damage of fuel element in the core during LOCA is caused by the oxidation of the cladding, its embrittlement and thermal shock stresses after initiation of the heat removal by a cold water from emergency core cooling system. In the paper the conservatism in criteria to avoid brittle ruptures of the fuel elements is discussed. Taking into account the influence of fuel burnup on the property of the cladding and a potential presence of air in the steam, it is believed that criteria of survivability of the zircaloy fuel cladding during LOCA may not be enough conservative.(author)

  9. Cladding creepdown model for FRAPCON-2

    International Nuclear Information System (INIS)

    Shah, V.N.; Tolli, J.E.

    1985-02-01

    This report presents a cladding deformation model developed to analyze cladding creepdown during steady state operation in both a pressurized water reactor (PWR) and a boiling water reactor (BWR). This model accounts for variations in zircaloy cladding heat treatment; cold worked and stress relieved material, typically used in a PWR, and fully recrystallized material, typically used in a BWR. The model calculates cladding creepdown as a function of hoop stress, fast neutron flux, exposure time, and temperature. This report also presents a comparison between cladding creep calculations by this model and corresponding measurements from the KWU/CE program, ORNL HOBBIE experiments, and EPRI/Westinghouse Engineering cooperative project. The comparisons show that the model calculates cladding creep strains well. The analyses of non-fueled rods by FRAPCON-2 show that the cladding creepdown model was correctly incorporated. Also, analysis of a PWR rod test case shows that the FRAPCON-2 code can analyze pellet-cladding mechanical interaction caused by cladding creepdown and fuel swelling

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

    Science.gov (United States)

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

    2015-05-01

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

  11. Fuel-clad heat transfer coefficient of a defected fuel rod

    International Nuclear Information System (INIS)

    Bruet, M.; Stora, J.P.

    1976-01-01

    A special rod has been built with a stack of UO 2 pellets inside a thick zircaloy clad. The atmosphere inside the fuel rod can be changed and particularly the introduction of water is possible. The capsule was inserted in the Siloe pool reactor in a special device equipped with a neutron flux monitor. The fuel centerline temperature and the temperature at a certain radius of the clad were recorded by two thermocouples. The temperature profiles in the fuel and in the cladding have been calculated and then the heat transfer coefficient. In order to check the proper functioning of the device, two runs were successively achieved with a helium atmosphere. Then the helium atmosphere inside the fuel rod was removed and replaced by water. The heat transfer coefficients derived from the measurements at low power level are in agreement with the values given by the model based on thermal conductivity. However, for higher power levels, the heat transfer coefficients become higher than those based on the calculated gap

  12. Zirconium-barrier cladding attributes

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.; Rand, R.A.; Tucker, R.P.; Cheng, B.; Adamson, R.B.; Davies, J.H.; Armijo, J.S.; Wisner, S.B.

    1987-01-01

    This metallurgical study of Zr-barrier fuel cladding evaluates the importance of three salient attributes: (1) metallurgical bond between the zirconium liner and the Zircaloy substrate, (2) liner thickness (roughly 10% of the total cladding wall), and (3) softness (purity). The effect that each of these attributes has on the pellet-cladding interaction (PCI) resistance of the Zr-barrier fuel was studied by a combination of analytical model calculations and laboratory experiments using an expanding mandrel technique. Each of the attributes is shown to contribute to PCI resistance. The effect of the zirconium liner on fuel behavior during off-normal events in which steam comes in contact with the zirconium surface was studied experimentally. Simulations of loss-of-coolant accident (LOCA) showed that the behavior of Zr-barrier cladding is virtually indistinguishable from that of conventional Zircaloy cladding. If steam contacts the zirconium liner surface through a cladding perforation and the fuel rod is operated under normal power conditions, the zirconium liner is oxidized more rapidly than is Zircaloy, but the oxidation rate returns to the rate of Zircaloy oxidation when the oxide phase reaches the zirconium-Zircaloy metallurgical bond

  13. Development of high performance cladding materials

    International Nuclear Information System (INIS)

    Park, Jeong Yong; Jeong, Y. H.; Park, S. Y.

    2010-04-01

    The irradiation test for HANA claddings conducted and a series of evaluation for next-HANA claddings as well as their in-pile and out-of pile performances tests were also carried out at Halden research reactor. The 6th irradiation test have been completed successfully in Halden research reactor. As a result, HANA claddings showed high performance, such as corrosion resistance increased by 40% compared to Zircaloy-4. The high performance of HANA claddings in Halden test has enabled lead test rod program as the first step of the commercialization of HANA claddings. DB has been established for thermal and LOCA-related properties. It was confirmed from the thermal shock test that the integrity of HANA claddings was maintained in more expanded region than the criteria regulated by NRC. The manufacturing process of strips was established in order to apply HANA alloys, which were originally developed for the claddings, to the spacer grids. 250 kinds of model alloys for the next-generation claddings were designed and manufactured over 4 times and used to select the preliminary candidate alloys for the next-generation claddings. The selected candidate alloys showed 50% better corrosion resistance and 20% improved high temperature oxidation resistance compared to the foreign advanced claddings. We established the manufacturing condition controlling the performance of the dual-cooled claddings by changing the reduction rate in the cold working steps

  14. BWR fuel clad behaviour following LOCA

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

    1996-01-01

    Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

  15. 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.

  16. 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

  17. 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

  18. 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.

  19. Unirradiated cladding rip-propagation tests

    International Nuclear Information System (INIS)

    Hu, W.L.; Hunter, C.W.

    1981-04-01

    The size of cladding rips which develop when a fuel pin fails can affect the subassembly cooling and determine how rapidly fuel escapes from the pin. The object of the Cladding Rip Propagation Test (CRPT) was to quantify the failure development of cladding so that a more realistic fuel pin failure modeling may be performed. The test results for unirradiated 20% CS 316 stainless steel cladding show significantly different rip propagation behavior at different temperatures. At room temperature, the rip growth is stable as the rip extension increases monotonically with the applied deformation. At 500 0 C, the rip propagation becomes unstable after a short period of stable rip propagation. The rapid propagation rate is approximately 200 m/s, and the critical rip length is 9 mm. At test temperatures above 850 0 C, the cladding exhibits very high failure resistances, and failure occurs by multiple cracking at high cladding deformation. 13 figures

  20. 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

  1. Capturing reflected cladding modes from a fiber Bragg grating with a double-clad fiber coupler.

    Science.gov (United States)

    Baiad, Mohamad Diaa; Gagné, Mathieu; Lemire-Renaud, Simon; De Montigny, Etienne; Madore, Wendy-Julie; Godbout, Nicolas; Boudoux, Caroline; Kashyap, Raman

    2013-03-25

    We present a novel measurement scheme using a double-clad fiber coupler (DCFC) and a fiber Bragg grating (FBG) to resolve cladding modes. Direct measurement of the optical spectra and power in the cladding modes is obtained through the use of a specially designed DCFC spliced to a highly reflective FBG written into slightly etched standard photosensitive single mode fiber to match the inner cladding diameter of the DCFC. The DCFC is made by tapering and fusing two double-clad fibers (DCF) together. The device is capable of capturing backward propagating low and high order cladding modes simply and efficiently. Also, we demonstrate the capability of such a device to measure the surrounding refractive index (SRI) with an extremely high sensitivity of 69.769 ± 0.035 μW/RIU and a resolution of 1.433 × 10(-5) ± 8 × 10(-9) RIU between 1.37 and 1.45 RIU. The device provides a large SRI operating range from 1.30 to 1.45 RIU with sufficient discrimination for all individual captured cladding modes. The proposed scheme can be adapted to many different types of bend, temperature, refractive index and other evanescent wave based sensors.

  2. 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

  3. HF-based clad etching of fibre Bragg grating and its utilization in ...

    Indian Academy of Sciences (India)

    2014-02-09

    Feb 9, 2014 ... Abstract. This paper presents a fiber Bragg grating (FBG) based sensor to study the concentration of laser dye in dye–ethanol solution. The FBG used in this experiment is indigenously developed using 255 nm UV radiations from copper vapour laser. The cladding of the FBG was partially removed using ...

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

    Science.gov (United States)

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

    2014-04-07

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

  5. Diffusion in cladding materials

    International Nuclear Information System (INIS)

    Anand, M.S.; Pande, B.M.; Agarwala, R.P.

    1992-01-01

    Aluminium has been used as a cladding material in most research reactors because its low neutron absorption cross section and ease of fabrication. However, it is not suitable for cladding in power reactors and as such zircaloy-2 is normally used as a clad because it can withstand high temperature. It has low neutron absorption cross section, good oxidation, corrosion, creep properties and possesses good mechanical strength. With the passage of time, further development in this branch of science took place and designers started looking for better neutron economy and less hydrogen pickup in PHW reactors. The motion of fission products in the cladding material could pose a problem after long operation. In order to understand their behaviour under reactor environment, it is essential to study first the diffusion under normal conditions. These studies will throw light on the interaction of defects with impurities which would in turn help in understanding the mechanism of diffusion. In this article, it is intended to discuss the diffusion behaviour of impurities in cladding materials.(i.e. aluminium, zircaloy-2, zirconium-niobium alloy etc.). (author). 94 refs., 4 figs., 3 tabs

  6. Cladding creepdown under compression

    International Nuclear Information System (INIS)

    Hobson, D.O.

    1977-01-01

    Light-water power reactors use Zircaloy tubing as cladding to contain the UO 2 fuel pellets. In-service operating conditions impose an external hydrostatic force on the cladding, causing it to creep down into eventual contact with the fuel. Knowledge of the rate of such creepdown is of great importance to modelers of fuel element performance. An experimental system was devised for studying creepdown that meets several severe requirements by providing (1) correct stress state, (2) multiple positions for measuring radial displacement of the cladding surface, (3) high-precision data, and (4) an experimental configuration compact enough to fit in-reactor. A microcomputer-controlled, eddy-current monitoring system was developed for this study and has proven highly successful in measuring cladding deformation with time at temperatures of 371 0 C (700 0 F) and higher, and at pressures as high as 21 MPa

  7. 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.)

  8. Review and evaluation of cladding attack of LMFBR fuel

    International Nuclear Information System (INIS)

    Koizumi, M.; Nagai, S.; Furuya, H.; Muto, T.

    1977-01-01

    The behavior of cladding inner wall corrosion during irradiation was evaluated in terms of fuel density, fuel form, O/M ratio, plutonium concentration, cladding composition, cladding pretreatment, cladding inner diameter, burnup and cladding inner wall temperature. Factors which influence the corrosion are O/M ratio (oxygen to metal ratio), burn up, cladding inner diameter and cladding inner wall temperature. Maximum cladding inner wall corrosion depth was formulated as a function of O/M ratio, burn up and cladding inner wall temperature

  9. 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

  10. Clad Treatment in KARMA Code and Library

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  11. Removal of Neutral Red from aqueous solution by adsorption on spent cottonseed hull substrate

    International Nuclear Information System (INIS)

    Zhou Qi; Gong Wenqi; Xie Chuanxin; Yang Dongjiang; Ling Xiaoqing; Yuan Xiao; Chen Shaohua; Liu Xiaofang

    2011-01-01

    Cottonseed hull, a low-cost widely available agricultural waste in China, after used as substrate for the white rot fungus Pleurotus ostreatus cultivation, was tested for the removal of Neutral Red (NR), a cationic dye, from aqueous solution. A batch adsorption study was carried out with varied solution pH, adsorbent dosage, reaction time and initial NR concentration. The results show that the kinetics of dye removal by the spent cottonseed hull substrate (SCHS) is prompt in the first 5 min and the adsorption equilibrium can be attained after 240 min. The biosorption kinetics and equilibrium follow typical pseudo-second-order and Langmuir adsorption models. Thermodynamic parameters of ΔG o , ΔH o and ΔS o show that the adsorption is a spontaneous and endothermic process. Fourier transform infrared (FTIR) spectroscopy was used for the characterization of possible dye-biosorbent interaction. This study provides a facile method to produce low-cost biosorbent for the purification of dye contaminated water.

  12. Design of Matched Cladding Fiber with UV-sensitive Cladding for Minimization of Claddingmode Losses in Fiber Bragg Gratings

    DEFF Research Database (Denmark)

    Nielsen, Mads Lønstrup; Berendt, Martin Ole; Bjarklev, Anders Overgaard

    2000-01-01

    The effect on the Bragg-grating-induced cladding-mode coupling of varying the extent of the photosensitive region in a step-index fiber is analyzed. We introduce a figure of merit for the suppression of cladding-mode loss and compare different matched cladding fiber designs. It is found to be adv......The effect on the Bragg-grating-induced cladding-mode coupling of varying the extent of the photosensitive region in a step-index fiber is analyzed. We introduce a figure of merit for the suppression of cladding-mode loss and compare different matched cladding fiber designs. It is found...... to be advantageous to increase the extent of the photosensitive region. However, no significant improvement is obtained by extending the photosensitive region more than approximately 10 mu m into the cladding. This result is not in agreement with a simple analysis that neglects UV absorption, which suggests...... that the radius of the photosensitive region should be close to twice as large. (C) 2000 Academic Press....

  13. Plug-welding of ODS cladding tube for BOR-60 irradiation. Welding condition setting. Device remodeling and welding

    International Nuclear Information System (INIS)

    Seki, Masayuki; Ishibashi, Fujio; Kono, Syusaku; Hirako, Kazuhito; Tsukada, Tatsuya

    2003-04-01

    Irradiation test in BOR-60 at RIAR to judge practical use prospect of ODS cladding tube at early stage is planned as Japan-Russia a joint research. RIAR does fuel design of fuel pin used for this joint research. JNC manufactures ODS cladding tube and bar materials (two steel kind of martensite and ferrite), upper endplug production. They are welded by pressurized resistance welding, and are inspected in JNC Tokai, transported to RIAR. And RIAR manufactures vibration packing fuel pin. On the upper endplug welding by pressurized resistance welding method, we worded on the problems such as decision of welding condition by changing the size and crystallization of cladding tube and the design of endplug, and the chucking device remodeling to correspond to the long scale cladding tube welding system (included handling) and of quality assurance method. Especially, use of long scale cladding tube caused problem that bending transformation occurred in cladding tube by welding pressure. However, we solved this problem by shortening the distance of cladding tube colette chuck and pressure receiving, and by putting the sleeve in an internal space of welding machine, losing the bending of cladding tube. Moreover, welding defects were occurred by the difference of an inside state, an inside defect and recrystallization of cladding tube. We solved the problem by inside grinding for the edge of tube, angle beam method by ultrasonic wave, and ultrasonic wave form confirmation. Manufacturing process with long scale cladding tube including heat-treatment to remove combustion return and remaining stress was established besides, Afterwards, welding of ODS cladding tube and upper endplug. As the quality assurance system, we constructed [Documented procedure (referred to JOYO)] based on [Document of the QA plan] by OEC. Welding and inspection were executed by the document procedure. It is thought that the quality assurance method become references for the irradiation test in JOYO in the

  14. 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

  15. Analysis of corrosion behavior of KOFA cladding

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Kim, Ki Hang; Seo, Keum Seok; Chung, Jin Gon

    1994-01-01

    The corrosion behavior of KOFA cladding was analyzed using the oxide measurement data of KOFA fuel irradiated up to the fuel rod burnup of 35,000 MWD/MTU for two cycles in Kori-2. Even though KOFA cladding is a standard Zircaloy-4 manufactured by Westinghouse according to the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 specification, it was expected that in-pile corrosion behavior of KOFA cladding would not be equivalent to that of Siemens/KWU's cladding due to the differences in such manufacturing processes as cold work and heat treatment. The analysis of measured KOFA cladding oxidation showed that oxidation of KOFA cladding is at least 19 % lower than the design analysis based upon Siemens/KWU's HCW standard Zircaloy-4 cladding. Lower corrosion of KOFA cladding seems to result from the differences in the manufacturing processes and chemical composition although the burnup and oxide layer thickness of the measured fuel rods is relatively low and the amount of the oxidation data base is small

  16. Development of Cr Electroplated Cladding Tube for preventing Fuel-Cladding Chemical Interaction (FCCI)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Woo, Je Woong; Kim, Sung Ho; Cheon, Jin Sik; Lee, Byung Oon; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Metal fuel has been selected as a candidate fuel in the SFR because of its superior thermal conductivity as well as enhanced proliferation resistance in connection with the pyroprocessing. However, metal fuel suffers eutectic reaction (Fuel Cladding Chemical Interaction, FCCI) with the fuel cladding made of stainless steel at reactor operating temperature so that cladding thickness gradually reduces to endanger reactor safety. In order to mitigate FCCI, barrier concept has been proposed between the fuel and the cladding in designing fuel rod. Regarding this, KAERI has initiated barrier cladding development to prevent interdiffusion process as well as enhance the SFR fuel performance. Previous study revealed that Cr electroplating has been selected as one of the most promising options because of its technical and economic viability. This paper describes the development status of the Cr electroplating technology for the usage of fuel rod in SFR. This paper summarizes the status of Cr electroplating technology to prevent FCCI in metal fuel rod. It has been selected for the ease of practical application at the tube inner surface. Technical scoping, performance evaluation and optimization have been carried out. Application to the tube inner surface and in-pile test were conducted which revealed as effective.

  17. Laser surface cladding:a literature survey

    OpenAIRE

    Gedda, Hans

    2000-01-01

    This work consists of a literature survey of a laser surface cladding in order to investigate techniques to improve the cladding rate for the process. The high local heat input caused by the high power density of the laser generates stresses and the process is consider as slow when large areas are processed. To avoid these disadvantages the laser cladding process velocity can be increased three or four times by use of preheated wire instead of the powder delivery system. If laser cladding is ...

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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)

  3. Solid-phase zirconium and fluoride species in alkaline zircaloy cladding waste at Hanford.

    Science.gov (United States)

    Reynolds, Jacob G; Huber, Heinz J; Cooke, Gary A; Pestovich, John A

    2014-08-15

    The United States Department of Energy Hanford Site, near Richland, Washington, USA, processed plutonium between 1944 and 1987. Fifty-six million gallons of waste of various origins remain, including waste from removing zircaloy fuel cladding using the so-called Zirflex process. The speciation of zirconium and fluoride in this waste is important because of the corrosivity and reactivity of fluoride as well as the (potentially) high density of Zr-phases. This study evaluates the solid-phase speciation of zirconium and fluoride using X-ray diffraction (XRD) and scanning electron microscopy with energy dispersive spectroscopy (SEM-EDS). Two waste samples were analyzed: one waste sample that is relatively pure zirconium cladding waste from tank 241-AW-105 and another that is a blend of zirconium cladding wastes and other high-level wastes from tank 241-C-104. Villiaumite (NaF) was found to be the dominant fluoride species in the cladding waste and natrophosphate (Na7F[PO4]2 · 19H2O) was the dominant species in the blended waste. Most zirconium was present as a sub-micron amorphous Na-Zr-O phase in the cladding waste and a Na-Al-Zr-O phase in the blended waste. Some zirconium was present in both tanks as either rounded or elongated crystalline needles of Na-bearing ZrO2 that are up to 200 μm in length. These results provide waste process planners the speciation data needed to develop disposal processes for this waste. Copyright © 2014 Elsevier B.V. All rights reserved.

  4. 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.

  5. 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

  6. Interaction between thorium and potential clad materials

    International Nuclear Information System (INIS)

    Kale, G.B.; Gawde, P.S.; Sengupta, Pranesh

    2005-01-01

    Thorium based fuels are being used for nuclear reactors. The structural stability of fuel-clad assemblies in reactor systems depend upon the nature of interdiffusion reaction between fuel-cladding materials. Interdiffusion reaction thorium and various cladding materials is presented in this paper. (author)

  7. Flaw density examinations of a clad boiling water reactor pressure vessel segment

    International Nuclear Information System (INIS)

    Cook, K.V.; McClung, R.W.

    1986-01-01

    Flaw density is the greatest uncertainty involved in probabilistic analyses of reactor pressure vessel failure. As part of the Heavy-Section Steel Technology (HSST) Program, studies have been conducted to determine flaw density in a section of reactor pressure vessel cut from the Hope Creek Unit 2 vessel [nominally 0.7 by 3 m (2 by 10 ft)]. This section (removed from the scrapped vessel that was never in service) was evaluated nondestructively to determine the as-fabricated status. We had four primary objectives: (1) evaluate longitudinal and girth welds for flaws with manual ultrasonics, (2) evaluate the zone under the nominal 6.3-mm (0.25-in.) clad for cracking (again with manual ultrasonics), (3) evaluate the cladding for cracks with a high-sensitivity fluorescent penetrant method, and (4) determine the source of indications detected

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

    International Nuclear Information System (INIS)

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

    1994-10-01

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

  10. 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...

  11. Cladding using a 15 kW CO2 laser

    International Nuclear Information System (INIS)

    Vesely, E.J.; Verma, S.K.

    1989-01-01

    Laser alloying or cladding differs little in principle from the traditional forms of weld overlays, but lasers as a heat source offer some distinct advantages. With the selective heating attainable using high power lasers, good metallurgical bond of the clad layer, minimal dilution and typically, a very fine homogeneous microstructure can be obtained in the clad layer. This is a review of work in laser cladding using the 15 kW CO 2 laser. The authors discuss the ability of the laser clad surface to increase the high temperature oxidation resistance of a low-alloy carbon steel (4140). Examples of clads subjected to high- temperature thermal cycling of nickel-20% aluminum and TaC + 4140 clad low-alloy steel and straight high-temperature oxidation of Stellite 6-304L cladding on a 4140 substrate are given

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

    International Nuclear Information System (INIS)

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

    1985-02-01

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

  13. Laser cladding technology to small diameter pipes

    International Nuclear Information System (INIS)

    Fujimagari, H.; Hagiwara, M.; Kojima, T.

    2000-01-01

    A laser cladding method which produces a highly corrosion-resistant material coating layers (cladding) on the austenitic stainless steel (type 304 SS) pipe inner surface was developed to prevent SCC (stress corrosion cracking) occurrence. This technology is applicable to a narrow and long distance area from operators, because of the good accessibility of the YAG (yttrium-aluminum-garnet) laser beam that can be transmitted through an optical fiber. In this method a mixed paste metallic powder and heating-resistive organic solvent are firstly placed on the inner surface of a small pipe, and then a YAG laser beam transmitted through an optical fiber irradiates to the pasted area. A mixed paste will be melted and form a cladding layer subsequently. A cladding layer shows as excellent corrosion resistance property. This laser cladding (LC) method had already applied to several domestic nuclear power plants and had obtained a good reputation. This report introduces the outline of laser cladding technology, the developed equipment for practical application in the field, and the circumstance in actual plant application. (orig.)

  14. Fuel assembly and fuel cladding tube

    International Nuclear Information System (INIS)

    Tsutsumi, Shinro; Ito, Ken-ichi; Inagaki, Masatoshi; Nakajima, Junjiro.

    1996-01-01

    A fuel cladding tube is a zirconium liner tube formed by lining a pure zirconium layer on the inner side of a zirconium alloy tube. The fuel cladding tube is formed by extrusion molding of a composite billet formed by inserting a pure zirconium billet into a zirconium alloy billet. Accordingly, the pure zirconium layer and the zirconium alloy tube are strongly joined by metal bond. The fuel cladding tube has an external oxide film on the outer surface of the zirconium alloy tube and an internal oxide film on the inner side of the pure zirconium layer. The external oxide film has a thickness preferably of about 1μm. The internal oxide film has a thickness of not more than 10μm, preferably, from 1 to 5μm. With such a constitution, flaws to be formed on both inner and outer surfaces of the cladding tube upon assembling a fuel assembly can be reduced thereby enabling to reduce the amount of hydrogen absorbed to the cladding tube. (I.N.)

  15. TMI-2 reactor-vessel head removal and damaged-core-removal planning

    International Nuclear Information System (INIS)

    Logan, J.A.; Hultman, C.W.; Lewis, T.J.

    1982-01-01

    A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

  16. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  18. 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.

  19. Protective claddings for high strength chromium alloys

    Science.gov (United States)

    Collins, J. F.

    1971-01-01

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

  20. Polarization characteristics of double-clad elliptical fibers.

    Science.gov (United States)

    Zhang, F; Lit, J W

    1990-12-20

    A scalar variational analysis based on a Gaussian approximation of the fundamental mode of a double-clad elliptical fiber with a depressed inner cladding is studied. The polarization properties and graphic results are presented; they are given in terms of three parameters: the ratio of the major axis to the minor axis of the core, the ratio of the inner cladding major axis to the core major axis, and the difference between the core index and the inner cladding index. The variations of both the spot size and the field intensity with core ellipticity are examined. It is shown that high birefringence and dispersion-free orthogonal polarization modes can be obtained within the single-mode region and that the field intensity distribution may be more confined to the fiber center than in a single-clad elliptical fiber.

  1. Neutral Color Semitransparent Microstructured Perovskite Solar Cells

    KAUST Repository

    Eperon, Giles E.

    2014-01-28

    Neutral-colored semitransparent solar cells are commercially desired to integrate solar cells into the windows and cladding of buildings and automotive applications. Here, we report the use of morphological control of perovskite thin films to form semitransparent planar heterojunction solar cells with neutral color and comparatively high efficiencies. We take advantage of spontaneous dewetting to create microstructured arrays of perovskite "islands", on a length-scale small enough to appear continuous to the eye yet large enough to enable unattenuated transmission of light between the islands. The islands are thick enough to absorb most visible light, and the combination of completely absorbing and completely transparent regions results in neutral transmission of light. Using these films, we fabricate thin-film solar cells with respectable power conversion efficiencies. Remarkably, we find that such discontinuous films still have good rectification behavior and relatively high open-circuit voltages due to the inherent rectification between the n- and p-type charge collection layers. Furthermore, we demonstrate the ease of "color-tinting" such microstructured perovksite solar cells with no reduction in performance, by incorporation of a dye within the hole transport medium. © 2013 American Chemical Society.

  2. 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.

  3. Passive neutralization of acid mine drainage using basic oxygen furnace slag as neutralization material: experimental and modelling.

    Science.gov (United States)

    Zvimba, John N; Siyakatshana, Njabulo; Mathye, Matlhodi

    2017-03-01

    This study investigated passive neutralization of acid mine drainage using basic oxygen furnace slag as neutralization material over 90 days, with monitoring of the parameters' quality and assessment of their removal kinetics. The quality was observed to significantly improve over time with most parameters removed from the influent during the first 10 days. In this regard, removal of acidity, Fe(II), Mn, Co, Ni and Zn was characterized by fast kinetics while removal kinetics for Mg and SO 4 2- were observed to proceed slowly. The fast removal kinetics of acidity was attributed to fast release of alkalinity from slag minerals under mildly acidic conditions of the influent water. The removal of acidity through generation of alkalinity from the passive treatment system was also observed to generally govern the removal of metallic parameters through hydroxide formation, with overall percentage removals of 88-100% achieved. The removal kinetics for SO 4 2- was modelled using two approaches, yielding rate constant values of 1.56 and 1.53 L/(day mol) respectively, thereby confirming authenticity of SO 4 2- removal kinetics experimental data. The study findings provide insights into better understanding of the potential use of slags and their limitations, particularly in mine closure, as part of addressing this challenge in South Africa.

  4. 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])

  5. Analysis of coaxial laser micro cladding processing conditions

    OpenAIRE

    Tarasova, Tatiana Vasilievna; Gvozdeva, Galina Olegovna; Nowotny, Steffen; Ableyeva, Riana R.; Dolzhikova, Evgenia Yu

    2018-01-01

    The laser build-up cladding is a well-known technique for repair, coatings and additive manufacturing tasks. Modern equipment for the laser cladding enables material to be deposited with the lateral resolution of about 100 μm and to manufacture miniature precise parts. However, the micro cladding regimes are unknown. Determination of these regimes is an expensive task as a well-known relation between laser cladding parameters and melt pool dimensions are changing by technology micro-miniaturi...

  6. 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)

  7. Impact of reactor water chemistry on cladding performance

    International Nuclear Information System (INIS)

    Cox, B.

    1997-01-01

    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)

  8. Some proposed mechanisms for internal cladding corrosion

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Whitlow, W.H.

    1977-01-01

    In spite of extensive research during recent years, a comprehensive model for internal cladding corrosion in fast reactor oxide fuel pins has not yet been established. In this paper, a model is proposed which accounts for many of the features normally associated with this type of corrosion. The model is composed of a number of parts which describe the chronological sequence of events at the fuel/cladding interface. The corrosion reaction is visualised as being primarily chemical in character, involving the cladding steel, the fuel and the more aggressive fission products, notably caesium in the presence of oxygen. The model attempts to explain how corrosion starts, how it depends on the oxygen potential, why it occurs non-uniformly; also covered are phase changes within the cladding steel and morphological features such as the intergranular form of attack and the distribution of corrosion products in the fuel/cladding gap. (author)

  9. Some proposed mechanisms for internal cladding corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, M H; Pickering, S; Whitlow, W H [EURATOM (United Kingdom)

    1977-04-01

    In spite of extensive research during recent years, a comprehensive model for internal cladding corrosion in fast reactor oxide fuel pins has not yet been established. In this paper, a model is proposed which accounts for many of the features normally associated with this type of corrosion. The model is composed of a number of parts which describe the chronological sequence of events at the fuel/cladding interface. The corrosion reaction is visualised as being primarily chemical in character, involving the cladding steel, the fuel and the more aggressive fission products, notably caesium in the presence of oxygen. The model attempts to explain how corrosion starts, how it depends on the oxygen potential, why it occurs non-uniformly; also covered are phase changes within the cladding steel and morphological features such as the intergranular form of attack and the distribution of corrosion products in the fuel/cladding gap. (author)

  10. Nuclear fuel cladding material

    International Nuclear Information System (INIS)

    Nakahigashi, Shigeo.

    1982-01-01

    Purpose: To largely improve the durability and the safety of fuel cladding material. Constitution: Diffusion preventive layers, e.g., aluminum or the like are covered on both sides of a zirconium alloy base layer of thin material, and corrosion resistant layers, e.g., copper or the like are covered thereon. This thin plate material is intimately wound in a circularly tubular shape in a plurality of layers to form a fuel cladding tube. With such construction, corrosion of the tube due to fuel and impurity can be prevented by the corrosion resistant layers, and the diffusion of the corrosion resistant material to the zirconium alloy can be prevented by the diffusion preventive layers. Since a plurality of layers are cladded, even if the corrosion resistant layers are damaged or cracked due to stress corrosion, only one layer is damaged or cracked, but the other layers are not affected. (Sekiya, K.)

  11. Deep-probe metal-clad waveguide biosensors

    DEFF Research Database (Denmark)

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

    2007-01-01

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

  12. Laser cladding of Zr on Mg for improved corrosion properties

    International Nuclear Information System (INIS)

    Subramanian, R.; Sircar, S.; Mazumder, J.

    1989-01-01

    This paper reports the results of laser cladding of Mg-2wt%Zr, and Mg-5wt%Zr powder mixture onto magnesium. The microstructure of the laser clad was studied. From the microstructural study, the epitaxial regrowth of the clad region on the underlying substrate was observed. Martensite plates of different size were observed in transmission electron microscope for MG-2wt%Zr and Mg-5wt%Zr laser clad. The corrosion properties of the laser clad were evaluated in sea water (3.5% NaCl). The position of the laser claddings in the galvanic series of metals in sea water, the anodic polarization characteristics of the laser claddings and the protective nature and the stability of the passivating film formed have been determined. The formation of pits on the surface of the laser clad subjected to corrosion is reported. The corrosion properties of the laser claddings are compared with that of the commercially used magnesium alloy AZ91B

  13. Oxidation properties of laser clad Nb-Al alloys

    International Nuclear Information System (INIS)

    Tewari, S.K.; Mazumder, J.

    1992-01-01

    This paper reports on laser cladding parameters for non-equilibrium synthesis for several ternary and complex Nb-Al base alloys containing Ti, Cr, Si, Ni, B and C that have been established. Phase transformations occurring below 1500 degrees C have been determined using differential thermal analysis. Ductility of the clads is qualitatively evaluated from the extent of cracking around the microhardness indentations. Oxidation resistance of the clads in flowing air is measured at 800 degrees C, 1200 degrees C and 1400 degrees C and parabolic rate constants are calculated. Microstructure of the clads is studied using optical and scanning electron microscopes. X-ray diffraction and EDX techniques are used for identification of the oxides formed and the phases formed in as clad material. Oxide morphology is studied using SEM. Effect of alloying additions on the ductility and oxidation resistance of the laser clad Nb-Al alloys is discussed. The results are compared with those reported in literature for similar alloys produced by conventional processing methods

  14. Mechanisms of fuel-cladding chemical interaction: US interpretation

    International Nuclear Information System (INIS)

    Adamson, M.G.

    1977-01-01

    Proposed mechanisms of fuel-cladding chemical interaction (FCCI) in LMFBR fuel pins are reviewed and examined in terms of in-pile and out-of-pile data. From this examination several factors are identified which may govern the occurrence of localized deep intergranular penetrations of Type-316SS cladding. Using a plausible mechanistic hypothesis for FCCI, first steps have been taken towards developing a quantitative, physically-meaningful, mathematical method of predicting cladding wastage in operating fuel pins. Both kinetic and thermodynamic aspects of FCCI are considered in the development of this prediction method, together with a fuel chemistry model that describes the evolution of thermochemical conditions at the fuel-cladding gap. On the basis of results from recent fuel pin and laboratory tests a thermal transport mechanism has been proposed to explain the thermal gradient-induced migration of Fe, Cr, and Ni from cladding into the fuel. This mechanism involves chemical transport of the metallic cladding components (as tellurides) in liquid Cs-Te. (author)

  15. Mechanisms of fuel-cladding chemical interaction: US interpretation

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States)

    1977-04-01

    Proposed mechanisms of fuel-cladding chemical interaction (FCCI) in LMFBR fuel pins are reviewed and examined in terms of in-pile and out-of-pile data. From this examination several factors are identified which may govern the occurrence of localized deep intergranular penetrations of Type-316SS cladding. Using a plausible mechanistic hypothesis for FCCI, first steps have been taken towards developing a quantitative, physically-meaningful, mathematical method of predicting cladding wastage in operating fuel pins. Both kinetic and thermodynamic aspects of FCCI are considered in the development of this prediction method, together with a fuel chemistry model that describes the evolution of thermochemical conditions at the fuel-cladding gap. On the basis of results from recent fuel pin and laboratory tests a thermal transport mechanism has been proposed to explain the thermal gradient-induced migration of Fe, Cr, and Ni from cladding into the fuel. This mechanism involves chemical transport of the metallic cladding components (as tellurides) in liquid Cs-Te. (author)

  16. Fundamentals and industrial applications of high power laser beam cladding

    International Nuclear Information System (INIS)

    Bruck, G.J.

    1988-01-01

    Laser beam cladding has been refined such that clad characteristics are precisely determined through routine process control. This paper reviews the state of the art of laser cladding optical equipment, as well as the fundamental process/clad relationships that have been developed for high power processing. Major categories of industrial laser cladding are described with examples chose to highlight particular process attributes

  17. Flaw behavior in mechanically loaded clad plates

    International Nuclear Information System (INIS)

    Iskander, S.K.; Robinson, G.C.; Oland, C.B.

    1989-01-01

    A small crack near the inner surface of clad nuclear reactor pressure vessels is an important consideration in the safety assessment of the structural integrity of the vessel. Four-point bend tests on large plate specimens, conforming to ASTM specification for pressure vessel plates, alloy steels, quenched and tempered, Mn-Mo and Mn-Mo-Ni (A533) grade B six clad and two unclad with stainless steels 308, 309 and 312 weld wires, were performed to determine the effect of cladding upon the propagation of small surface cracks subjected to stress states. Results indicated that the tough surface layer composed of cladding and/or heat-affected zone has enhanced the load-bearing capacity of plates under conditions where unclad plates have ruptured. The results are interpreted in terms of fracture mechanics. The behavior of flaws in clad reactor pressure vessels is examined in the light of the test results. 11 refs., 8 figs., 2 tabs

  18. 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)

  19. Pellet-clad interaction in water reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

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

  20. 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.)

  1. Determination of plastic anisotropy of zirconium alloys cladding

    International Nuclear Information System (INIS)

    Yamshchikov, N.V.; Prasolov, P.F.; Shestak, V.E.

    1991-01-01

    Method for determining plastic anisotropy of zurconium alloy cladding is described. It is based on consideration of material as a combination of transversal crystallites with known distribution over orientations. Such approach enables to describe cladding resistance to plastic deformation at arbitrary stressed state, using the results of texture investigations and uniaxial tests of samples, cut out of claddings along three directions. Plastic anisotropy of fuel element claddings 9.15 and 13.6 mm in diameter up to several percents of plastic deformation is shown

  2. Cladding modes of optical fibers: properties and applications

    International Nuclear Information System (INIS)

    Ivanov, Oleg V; Nikitov, Sergei A; Gulyaev, Yurii V

    2006-01-01

    One of the new methods of fiber optics uses cladding modes for controlling propagation of radiation in optical fibers. This paper reviews the results of studies on the propagation, excitation, and interaction of cladding modes in optical fibers. The resonance between core and cladding modes excited by means of fiber Bragg gratings, including tilted ones, is analyzed. Propagation of cladding modes in microstructured fibers is considered. The most frequently used method of exciting cladding modes is described, based on the application of long-period fiber gratings. Examples are presented of long-period gratings used as sensors and gain equalizers for fiber amplifiers, as well as devices for coupling light into and out of optical fibers. (instruments and methods of investigation)

  3. Neutron-induced helium implantation in GCFR cladding

    International Nuclear Information System (INIS)

    Yamada, H.; Poeppel, R.B.; Sevy, R.H.

    1980-10-01

    The neutron-induced implantation of helium atoms on the exterior surfaces of the cladding of a prototypic gas-cooled fast reactor (GCFR) has been investigated analytically. A flux of recoil helium particles as high as 4.2 x 10 10 He/cm 2 .s at the cladding surface has been calculated at the peak power location in the core of a 300-MWe GCFR. The calculated profile of the helium implantation rates indicates that although some helium is implanted as deep as 20 μm, more than 99% of helium particles are implanted in the first 2-μm-deep layer below the cladding surface. Therefore, the implanted helium particles should mainly affect surface properties of the GCFR cladding

  4. Fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    Gueneau, C.; Piron, J.P.; Dumas, J.C.; Bouineau, V.; Iglesias, F.C.; Lewis, B.J.

    2015-01-01

    The chemistry of the nuclear fuel is very complex. Its chemical composition changes with time due to the formation of fission products and depends on the temperature level history within the fuel pellet and the clad during operation. Firstly, in thermal reactors, zircaloy oxidation from reaction with UO 2 fuel under high-temperature conditions will be addressed. Then other fuel-cladding interaction phenomena occurring in fast reactors will be described. Large thermal gradients existing between the centre and the periphery of the pellet induce the radial redistribution of the fuel constituents. The fuel pellet can react with the clad by different corrosion processes which can involve actinide and/or fission product transport via gas, liquid or/and solid phases. All these phenomena are briefly described in the case of different kinds of fuels (oxide, carbide, nitride, metallic) to be used in fast reactors. The way these phenomena are taken into account in fuel performance codes is presented. (authors)

  5. 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

  6. Influence of texture on fracture toughness of zircaloy cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-06-01

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

  7. 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

  8. 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)

  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. Experimental approach for adhesion strength of ATF cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Donghyun; Kim, Hyochan; Yang, Yongsik; In, Wangkee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Haksung [Hanyang University, Seoul (Korea, Republic of)

    2016-10-15

    The quality of a coating depends on the quality of its adhesion bond strength between the coating and the underlying substrate. Therefore, it is essential to evaluate the adhesion properties of the coating. There are many available test methods for the evaluation of coatings adhesion bond strength. Considering these restrictions of the coated cladding, the scratch test is useful for evaluation of adhesion properties compared to other methods. The purpose of the present study is to analyze the possibility of adhesion bond strength evaluation of ATF coated cladding by scratch testing on coatings cross sections. Experimental approach for adhesion strength of ATF coated cladding was investigated in the present study. The scratch testing was chosen as a testing method. Uncoated zircaloy-4 tube was employed as a reference and plasma spray and arc ion coating were selected as a ATF coated claddings for comparison. As a result, adhesion strengths of specimens affect the measured normal and tangential forces. For the future, the test will be conducted for CrAl coated cladding by laser coating, which is the most promising ATF cladding. Computational analysis with finite element method will also be conducted to analyze a stress distribution in the cladding tube.

  11. 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.

  12. Assessment of the fracture behavior of weld material from a full-thickness clad RPV shell segment

    International Nuclear Information System (INIS)

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

    1995-01-01

    A testing program is described that utilizes full-thickness clad beam specimens to quantify fracture toughness for shallow cracks in material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPVs). The beam specimens are fabricated from a section of an RPV shell (removed from a canceled nuclear plant) that includes weld, plate, and clad material. A summary of the testing program includes a description of the specimen geometry, material properties, the testing procedure, and the experimental results from three specimens. The yield strength of the weld material was determined to be 36% higher than the base material. The high yield strength for prototypic weld material may be implications for RPV integrity assessments. Fracture toughness data from three clad beam specimens are compared with other shallow- and deep-crack beam cruciform data generated previously from A 533 Grade B plate material. Difficulties with interpreting lower-bound fracture toughness curves constructed from the shallow-crack data are essentially resolved by adopting a single normalizing temperature parameter, namely, the nil-ductility transition temperature (NDT)

  13. 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

  14. Duplex stainless steel surface bay laser cladding

    International Nuclear Information System (INIS)

    Amigo, V.; Pineda, Y.; Segovia, F.; Vicente, A.

    2004-01-01

    Laser cladding is one of the most promising techniques to restore damaged surfaces and achieve properties similar to those of the base metal. In this work, duplex stainless steels have been cladded by a nickel alloy under different processing conditions. The influence of the beam speed and defocusing variables ha been evaluated in the microstructure both of the cladding and heat affected zone, HAZ. These results have been correlated to mechanical properties by means of microhardness measurements from cladding area to base metal through the interface. This technique has shown to be very appropriate to obtain controlled mechanical properties as they are determined by the solidification microstructure, originated by the transfer of mass and heat in the system. (Author) 21 refs

  15. Corrosion behaviour of cladded nickel base alloys

    International Nuclear Information System (INIS)

    Brandl, W.; Ruczinski, D.; Nolde, M.; Blum, J.

    1995-01-01

    As a consequence of the high cost of nickel base alloys their use as surface layers is convenient. In this paper the properties of SA-as well as RES-cladded NiMo 16Cr16Ti and NiCr21Mo14W being produced in single and multi-layer technique are compared and discussed with respect to their corrosion behaviour. Decisive criteria describing the qualities of the claddings are the mass loss, the susceptibility against intergranular corrosion and the pitting corrosion resistance. The results prove that RES cladding is the most suitable technique to produce corrosion resistant nickel base coatings. The corrosion behaviour of a two-layer RES deposition shows a better resistance against pitting than a three layer SAW cladding. 7 refs

  16. Dispersion properties of plasma cladded annular optical fiber

    Science.gov (United States)

    KianiMajd, M.; Hasanbeigi, A.; Mehdian, H.; Hajisharifi, K.

    2018-05-01

    One of the considerable problems in a conventional image transferring fiber optic system is the two-fold coupling of propagating hybrid modes. In this paper, using a simple and practical analytical approach based on exact modal vectorial analysis together with Maxwell's equations, we show that applying plasma as a cladding medium of an annular optical fiber can remove this defect of conventional fiber optic automatically without any external instrument as the polarization beam splitter. Moreover, the analysis indicates that the presence of plasma in the proposed optical fiber could extend the possibilities for controlling the propagation property. The proposed structure presents itself as a promising route to advanced optical processing and opens new avenues in applied optics and photonics.

  17. Waste removal in pyrochemical fuel processing for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Ackerman, J.P.; Johnson, T.R.; Laidler, J.J.

    1994-01-01

    Electrorefining in a molten salt electrolyte is used in the Integral Fast Reactor fuel cycle to recover actinides from spent fuel. Processes that are being developed for removing the waste constituents from the electrorefiner and incorporating them into the waste forms are described in this paper. During processing, halogen, chalcogen, alkali, alkaline earth, and rare earth fission products build up in the molten salt as metal halides and anions, and fuel cladding hulls and noble metal fission products remain as metals of various particle sizes. Essentially all transuranic actinides are collected as metals on cathodes, and are converted to new metal fuel. After processing, fission products and other waste are removed to a metal and a mineral waste form. The metal waste form contains the cladding hulls, noble metal fission products, and (optionally) most rare earths in a copper or stainless steel matrix. The mineral waste form contains fission products that have been removed from the salt into a zeolite or zeolite-derived matrix

  18. 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...

  19. Electron beam cladding of titanium on stainless steel plate

    International Nuclear Information System (INIS)

    Tomie, Michio; Abe, Nobuyuki; Yamada, Masanori; Noguchi, Shuichi.

    1990-01-01

    Fundamental characteristics of electron beam cladding was investigated. Titanium foil of 0.2mm thickness was cladded on stainless steel plate of 3mm thickness by scanning electron beam. Surface roughness and cladded layer were analyzed by surface roughness tester, microscope, scanning electron microscope and electron probe micro analyzer. Electron beam conditions were discussed for these fundamental characteristics. It is found that the energy density of the electron beam is one of the most important factor for cladding. (author)

  20. Pin clad strains in Phenix

    International Nuclear Information System (INIS)

    Languille, A.

    1979-07-01

    The Phenix reactor has operated for 4 years in a satisfactory manner. The first 2 sub-assembly loadings contained pins clad in solution treated 316. The principal pin strains are: diametral strain (swelling and irradiation creep), ovality and spiral bending of the pin (interaction of wire and pin cluster and wrapper). A pin cluster irradiated to a dose of 80 dpa F reached a pin diameter strain of 5%. This strain is principally due to swelling (low fission gas pressure). The principal parameters governing the swelling are instantaneous dose, time and temperature for a given type of pin cladding. Other types of steel are or will be irradiated in Phenix. In particular, cold-worked titanium stabilised 316 steel should contribute towards a reduction in the pin clad strains and increase the target burn-up in this reactor. (author)

  1. Potential for cladding thermal failure in LWRs during high temperature transients

    International Nuclear Information System (INIS)

    El Genk, M.S.

    1979-01-01

    The temperature increase in the fuel and the cladding during a PCM accident produces film boiling at the cladding surface which may induce zircaloy cladding failure, due to embrittlement, and fuel melting at the centerline of the fuel pellets. Molten fuel may extrude through radial cracks in the fuel and relocate in the fuel-cladding gap. Contact of extruded molten fuel with the cladding, which is at high temperature during film boiling, may induce cladding thermal failure due to melting. An assessment of central fuel melting and molten fuel extrusion into the fuel-cladding gap during a PCM accident is presented. The potential for thermal failure of the zircaloy cladding upon being contacted by molten fuel during such an accident is also analyzed and compared with the applicable experimental evidence

  2. Explosion Clad for Upstream Oil and Gas Equipment

    Science.gov (United States)

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

    2011-01-01

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

  3. Explosion Clad for Upstream Oil and Gas Equipment

    International Nuclear Information System (INIS)

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

    2011-01-01

    Today's upstream oil and gas facilities frequently involve the combination of high pressures, high temperatures, and highly corrosive environments, requiring equipment that is thick wall, corrosion resistant, and cost effective. When significant concentrations of CO 2 and/or H 2 S and/or chlorides are present, corrosion resistant alloys (CRA) can become the material of choice for separator equipment, piping, related components, and line pipe. They can provide reliable resistance to both corrosion and hydrogen embrittlement. For these applications, the more commonly used CRA's are 316L, 317L and duplex stainless steels, alloy 825 and alloy 625, dependent upon the application and the severity of the environment. Titanium is also an exceptional choice from the technical perspective, but is less commonly used except for heat exchangers. Explosion clad offers significant savings by providing a relatively thin corrosion resistant alloy on the surface metallurgically bonded to a thick, lower cost, steel substrate for the pressure containment. Developed and industrialized in the 1960's the explosion cladding technology can be used for cladding the more commonly used nickel based and stainless steel CRA's as well as titanium. It has many years of proven experience as a reliable and highly robust clad manufacturing process. The unique cold welding characteristics of explosion cladding reduce problems of alloy sensitization and dissimilar metal incompatibility. Explosion clad materials have been used extensively in both upstream and downstream oil, gas and petrochemical facilities for well over 40 years. The explosion clad equipment has demonstrated excellent resistance to corrosion, embrittlement and disbonding. Factors critical to insure reliable clad manufacture and equipment design and fabrication are addressed.

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

    International Nuclear Information System (INIS)

    Hu Jianliang; Yi Youping; Xie Mantang

    2011-01-01

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

  5. Study on process of laser cladded nuclear valve parts

    International Nuclear Information System (INIS)

    Zhang Chunliang

    2000-01-01

    The microstructure and performances of the Co-base alloy coatings that are formed by laser cladding, plasma spurt welding and arc surfacing on the nuclear valve-sealing surface have been studied and compared. The combination costs of laser cladding, plasma spurt welding and arc, surfacing have been analyzed and compared. The results showed that the laser cladding processing has the advantages of high efficiency, low energy cost, a little machining allowance, high rate of finished products and low combination cost, compared with plasma spurt welding processing and arc surfacing processing. The laser cladding technology can improve the qualities of nuclear valve parts and increase their service life. Therefore, the laser cladding processing is a new technology with developing potential

  6. Oxide thickness measurement technique for duplex-layer Zircaloy-4 cladding

    International Nuclear Information System (INIS)

    McClelland, R.G.; O'Leary, P.M.

    1992-01-01

    Siemens Nuclear Power Corporation (SNP) is investigating the use of duplex-layer Zircaloy-4 tubing to improve the waterside corrosion resistance of cladding for high-burnup pressurized water reactor (PWR) fuel designs. Standard SNP PWR cladding is typically 0.762-mm (0.030-in.)-thick Zircaloy-4. The SNP duplex cladding is nominally 0.660-mm (0.026-in.)-thick Zircalloy-4 with an ∼0.102-mm (0.004-in.) outer layer of another, more corrosion-resistant, zirconium-based alloy. It is common industry practice to monitor the in-reactor corrosion behavior of Zircaloy cladding by using an eddy-current 'lift-off' technique to measure the oxide thickness on the outer surface of the fuel cladding. The test program evaluated three different cladding samples, all with the same outer diameter and wall thickness: Zircaloy-4 and duplex clad types D2 and D4

  7. Investigations on dry sliding of laser cladded aluminum bronze

    Directory of Open Access Journals (Sweden)

    Freiße Hannes

    2016-01-01

    Full Text Available The aim of this study was to investigate the tribological behaviour of laser cladded aluminum bronze tool surfaces for dry metal forming. In a first part of this work a process window for cladding aluminum bronze on steel substrate was investigated to ensure a low dilution. Therefore, the cladding speed, the powder feed rate, the laser power and the distance between the process head and the substrate were varied. The target of the second part was to investigate the influence of different process parameters on the tribological behaviour of the cladded tracks. The laser claddings were carried out on both aluminum bronze and cold work tool steel as substrate materials. Two different particle sizes of the cladding powder material were used. The cladding speed was varied and a post-processing laser remelting treatment was applied. It is shown that the tribological behaviour of the surface in a dry oscillating ball-on-plate test is highly dependent on the substrate material. In the third part a deep drawing tool was additively manufactured by direct laser deposition. Furthermore, the tool was applied to form circular cups with and without lubrication.

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

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L. E.

    2014-07-01

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

  9. Fuel cladding behavior under rapid loading conditions

    Science.gov (United States)

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

    2016-02-01

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

  10. 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.

  11. Clad buffer rod sensors for liquid metals

    International Nuclear Information System (INIS)

    Jen, C.-K.; Ihara, I.

    1999-01-01

    Clad buffer rods, consisting of a core and a cladding, have been developed for ultrasonic monitoring of liquid metal processing. The cores of these rods are made of low ultrasonic-loss materials and the claddings are fabricated by thermal spray techniques. The clad geometry ensures proper ultrasonic guidance. The lengths of these rods ranges from tens of centimeters to 1m. On-line ultrasonic level measurements in liquid metals such as magnesium at 700 deg C and aluminum at 960 deg C are presented to demonstrate their operation at high temperature and their high ultrasonic performance. A spherical concave lens is machined at the rod end for improving the spatial resolution. High quality ultrasonic images have been obtained in the liquid zinc at 600 deg C. High spatial resolution is needed for the detection of inclusions in liquid metals during processing. We also show that the elastic properties such as density, longitudinal and shear wave velocities of liquid metals can be measured using a transducer which generates and receives both longitudinal and shear waves and is mounted at the end of a clad buffer rod. (author)

  12. 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

  13. Study on modes of energy action in laser-induction hybrid cladding

    International Nuclear Information System (INIS)

    Huang Yongjun; Zeng Xiaoyan

    2009-01-01

    The shape and microstructure in laser-induction hybrid cladding were investigated, in which the cladding material was provided by means of three different methods including the powder feeding, cold pre-placed coating (CPPC) and thermal pre-placed coating (TPPC). Moreover, the modes of energy action in laser-induction hybrid cladding were also studied. The results indicate that the cladding material supplying method has an important influence on the shape and microstructure of coating. The influence is decided by the mode of energy action in laser-induction hybrid cladding. During the TPPC hybrid cladding of Ni-based alloy, the laser and induction heating are mainly performed on coating. During the CPPC hybrid cladding of Ni-based alloy, the laser and induction heating are mainly performed on coating and substrate surface, respectively. In powder feeding hybrid cladding, a part of laser is absorbed by the powder particles directly, while the other part of laser penetrating powder cloud radiates on the molten pool. Meanwhile, the induction heating is entirely performed on the substrate. In addition, the wetting property on the interface is improved and the metallurgical bond between the coating and substrate is much easier to form. Therefore, the powder feeding laser-induction hybrid cladding has the highest cladding efficiency and the best bond property among three hybrid cladding methods.

  14. 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.

  15. 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.

  16. 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

  17. 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

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

    Science.gov (United States)

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

    2017-11-01

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

  19. Rectangular-cladding silicon slot waveguide with improved nonlinear performance

    Science.gov (United States)

    Huang, Zengzhi; Huang, Qingzhong; Wang, Yi; Xia, Jinsong

    2018-04-01

    Silicon slot waveguides have great potential in hybrid silicon integration to realize nonlinear optical applications. We propose a rectangular-cladding hybrid silicon slot waveguide. Simulation result shows that, with a rectangular-cladding, the slot waveguide can be formed by narrower silicon strips, so the two-photon absorption (TPA) loss in silicon is decreased. When the cladding material is a nonlinear polymer, the calculated TPA figure of merit (FOMTPA) is 4.4, close to the value of bulk nonlinear polymer of 5.0. This value confirms the good nonlinear performance of rectangular-cladding silicon slot waveguides.

  20. 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)

  1. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  3. Application of Coating Technology for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  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. 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...

  6. Comparison of fiber lasers based on distributed side-coupled cladding-pumped fibers and double-cladding fibers.

    Science.gov (United States)

    Huang, Zhihe; Cao, Jianqiu; Guo, Shaofeng; Chen, Jinbao; Xu, Xiaojun

    2014-04-01

    We compare both analytically and numerically the distributed side-coupled cladding-pumped (DSCCP) fiber lasers and double cladding fiber (DCF) lasers. We show that, through optimization of the coupling and absorbing coefficients, the optical-to-optical efficiency of DSCCP fiber lasers can be made as high as that of DCF lasers. At the same time, DSCCP fiber lasers are better than the DCF lasers in terms of thermal management.

  7. Fuel cladding mechanical interaction during power ramps

    International Nuclear Information System (INIS)

    Guerin, Y.

    1985-01-01

    Mechanical interaction between fuel and cladding may occur as a consequence of two types of phenomenon: i) fuel swelling especially at levels of caesium accumulation, and ii) thermal differential expansion during power changes. Slow overpower ramps which may occur during incidental events are of course one of the circumstances responsible for this second type of fuel cladding mechanical interaction (FCMI). Experiments and analysis of this problem that have been done at C.E.A. allow to determine the main parameters which will fix the level of stress and the risk of damage induced by the fuel in the cladding during overpower transients

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

    Directory of Open Access Journals (Sweden)

    CHEN Ming-hui

    2017-01-01

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

  9. Electrolysis of plutonium in neutral and basic solutions

    International Nuclear Information System (INIS)

    1978-01-01

    Experiments were conducted on electrolysis of Pu in waste streams. Removal of Pu by this process is maximum at pH 11. Runs on an actual waste stream showed that: Pu can be electrolyzed from neutral or basic solutions down to 10 -10 g/l. Am can also be removed. The removal efficiency is pH dependent. The deposits can be removed by acid leaching

  10. DEVELOPMENT OF LASER CLADDING WEAR-RESISTANT COATING ON TITANIUM ALLOYS

    OpenAIRE

    RUILIANG BAO; HUIJUN YU; CHUANZHONG CHEN; BIAO QI; LIJIAN ZHANG

    2006-01-01

    Laser cladding is an advanced surface modification technology with broad prospect in making wear-resistant coating on titanium alloys. In this paper, the influences of laser cladding processing parameters on the quality of coating are generalized as well as the selection of cladding materials on titanium alloys. The microstructure characteristics and strengthening mechanism of coating are also analyzed. In addition, the problems and precaution measures in the laser cladding are pointed out.

  11. Development of laser surface cladding through energy transmission over optical fiber

    International Nuclear Information System (INIS)

    Hirano, Kenji; Morishige, Norio; Irisawa, Toshio

    1990-01-01

    Much attention has recently been paid to laser cladding techniques as an approach in controlling the composition and structure of the metal surface. If YAG laser is used as the cladding method, the flexibility of laser cladding process increases extremely because YAG laser beam is transmitted through an optical fiber, and enabling cladding on pipes installed in actual plants. So experiments on YAG laser cladding through energy transmission over an optical fiber were performed to prevent stress corrosion cracking in austenitic stainless steel pipes. In order to build a cladding layer, mixed metal powder were pre-placed on the inner surface of the pipe using organic binder and the pre-placed powder beds were melted with YAG laser beam transmitted using an optical fiber. This paper introduces the method of building a cladding layer on pipes in actual nuclear plants. (author)

  12. 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)

  13. Reduction of Bragg-grating-induced coupling to cladding modes

    DEFF Research Database (Denmark)

    Berendt, Martin Ole; Bjarklev, Anders Overgaard; Soccolich, C.E.

    1999-01-01

    gratings in a depressed-cladding fiber are compared with simulations. The model gives good agreement with the measured transmission spectrum and accounts for the pronounced coupling to asymmetrical cladding modes, even when the grating is written with the smallest possible blaze. The asymmetry causing...... this is accounted for by the unavoidable attenuation of the UV light. It is found for the considered fiber designs that a high numerical-aperture fiber increases the spectral separation between the Bragg resonance and the onset of cladding-mode losses. A depressed-cladding fiber reduces the coupling strength......We discuss fiber designs that have been suggested for the reduction of Bragg-grating induced coupling to cladding modes. The discussion is based on a theoretical approach that includes the effect of asymmetry in the UV-induced index grating, made by UV-side writing. Experimental results from...

  14. Revitalization of Lightweight Cladding of Buildings and Its Impact on Environment

    Science.gov (United States)

    Liška, Pavel; Nečasová, Barbora; Kovářová, Barbora; Novotný, Michal

    2017-12-01

    The presented study reveals that the revitalization of lightweight claddings installed before 1990 can have a positive impact on the environment and on the reduction of greenhouse gases in particular. The main focus is placed on the revitalization of a structural system known as OD-001, commonly called the ‘Boleticky panel’ system, which was frequently utilised all around the Czech Republic in the period before 1990. Only revitalization methods utilizing contemporary structural designs and current materials were verified during this study. The ‘Boleticky panel’ system was the type of façade cladding most frequently installed on administrative buildings in what was then Czechoslovakia. It is a panel system where load-bearing structure of the panel itself consists of closed profiles that are suspended from the building’s load-bearing structure. This type of system saw a great deal of use for more than 20 years. From today’s point of view, its thermal and technical properties are completely unsatisfactory and the gradual structural degradation of such systems, with a direct impact on their mechanical resistance, has been monitored over the last few years. However, these defects can be completely eliminated by the selection of a suitable type of revitalization. Cladding revitalization can be divided into three main categories. Each category represents a different level of impact on the structure of the above described cladding system. The first category only involves the replacement of windows, while the second consists in the replacement both of the windows and the existing panel sections. The third category of revitalization entails the complete removal of the existing cladding system and its replacement with a new one. The Life Cycle Assessment method (LCA) was used for environmental impact assessment. The aims and intentions of this method are not to search for the most economical or technically perfect product, service or technology, but to find the

  15. 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.

  16. 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.

  17. Cladding properties under simulated fuel pin transients

    International Nuclear Information System (INIS)

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

    1975-01-01

    A description is given of the HEDL fuel pin testing program utilizing a recently developed Fuel Cladding Transient Tester (FCTT) to generate the requisite mechanical property information on irradiated and unirradiated fast reactor fuel cladding under temperature ramp conditions. The test procedure is described, and data are presented

  18. Formation of anomalous eutectic in Ni-Sn alloy by laser cladding

    Science.gov (United States)

    Wang, Zhitai; Lin, Xin; Cao, Yongqing; Liu, Fencheng; Huang, Weidong

    2018-02-01

    Ni-Sn anomalous eutectic is obtained by single track laser cladding with the scanning velocity from 1 mm/s to 10 mm/s using the Ni-32.5 wt.%Sn eutectic powders. The microstructure of the cladding layer and the grain orientations of anomalous eutectic were investigated. It is found that the microstructure is transformed from primary α-Ni dendrites and the interdendritic (α-Ni + Ni3Sn) eutectic at the bottom of the cladding layer to α-Ni and β-Ni3Sn anomalous eutectic at the top of the cladding layer, whether for single layer or multilayer laser cladding. The EBSD maps and pole figures indicate that the spatially structure of α-Ni phase is discontinuous and the Ni3Sn phase is continuous in anomalous eutectic. The transformation from epitaxial growth columnar at bottom of cladding layer to free nucleation equiaxed at the top occurs, i.e., the columnar to equiaxed transition (CET) at the top of cladding layer during laser cladding processing leads to the generation of anomalous eutectic.

  19. Characteristics of Ni-based coating layer formed by laser and plasma cladding processes

    International Nuclear Information System (INIS)

    Xu Guojian; Kutsuna, Muneharu; Liu Zhongjie; Zhang Hong

    2006-01-01

    The clad layers of Ni-based alloy were deposited on the SUS316L stainless plates by CO 2 laser and plasma cladding processes. The smooth clad bead was obtained by CO 2 laser cladding process. The phases of clad layer were investigated by an optical microscope, scanning electron microscopy (SEM), X-ray diffractometer (XRD), electron probe microanalysis (EPMA) and energy-dispersive spectrometer (EDS). The microstructures of clad layers belonged to a hypereutectic structure. Primary phases consist of boride CrB and carbide Cr 7 C 3 . The eutectic structure consists of Ni + CrB or Ni + Cr 7 C 3 . Compared with the plasma cladding, the fine microstructures, low dilutions, high Vickers hardness and excellent wear resistance were obtained by CO 2 laser cladding. All that show the laser cladding process has a higher efficiency and good cladding quality

  20. Composite polymer: Glass edge cladding for laser disks

    Science.gov (United States)

    Powell, H.T.; Wolfe, C.A.; Campbell, J.H.; Murray, J.E.; Riley, M.O.; Lyon, R.E.; Jessop, E.S.

    1987-11-02

    Large neodymium glass laser disks for disk amplifiers such as those used in the Nova laser require an edge cladding which absorbs at 1 micrometer. This cladding prevents edge reflections from causing parasitic oscillations which would otherwise deplete the gain. Nova now utilizes volume-absorbing monolithic-glass claddings which are fused at high temperature to the disks. These perform quite well but are expensive to produce. Absorbing glass strips are adhesively bonded to the edges of polygonal disks using a bonding agent whose index of refraction matches that of both the laser and absorbing glass. Optical finishing occurs after the strips are attached. Laser disks constructed with such claddings have shown identical gain performance to the previous Nova disks and have been tested for hundreds of shots without significant degradation. 18 figs.

  1. Composite polymer-glass edge cladding for laser disks

    Science.gov (United States)

    Powell, Howard T.; Riley, Michael O.; Wolfe, Charles R.; Lyon, Richard E.; Campbell, John H.; Jessop, Edward S.; Murray, James E.

    1989-01-01

    Large neodymium glass laser disks for disk amplifiers such as those used in the Nova laser require an edge cladding which absorbs at 1 micrometer. This cladding prevents edge reflections from causing parasitic oscillations which would otherwise deplete the gain. Nova now utilizes volume-absorbing monolithic-glass claddings which are fused at high temperature to the disks. These perform quite well but are expensive to produce. Absorbing glass strips are adhesively bonded to the edges of polygonal disks using a bonding agent whose index of refraction matches that of both the laser and absorbing glass. Optical finishing occurs after the strips are attached. Laser disks constructed with such claddings have shown identical gain performance to the previous Nova disks and have been tested for hundreds of shots without significant degradation.

  2. Electrical discharge machining for vessel sample removal

    International Nuclear Information System (INIS)

    Litka, T.J.

    1993-01-01

    Due to aging-related problems or essential metallurgy information (plant-life extension or decommissioning) of nuclear plants, sample removal from vessels may be required as part of an examination. Vessel or cladding samples with cracks may be removed to determine the cause of cracking. Vessel weld samples may be removed to determine the weld metallurgy. In all cases, an engineering analysis must be done prior to sample removal to determine the vessel's integrity upon sample removal. Electrical discharge machining (EDM) is being used for in-vessel nuclear power plant vessel sampling. Machining operations in reactor coolant system (RCS) components must be accomplished while collecting machining chips that could cause damage if they become part of the flow stream. The debris from EDM is a fine talclike particulate (no chips), which can be collected by flushing and filtration

  3. 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....

  4. 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

  5. Recent trend of titanium-clad steel plate/sheet (NKK)

    International Nuclear Information System (INIS)

    Kimura, Hideto

    1997-01-01

    The roll-bonding process for titanium-clad steel production enabled the on-line manufacturing and quality control of the products which are usually applied for the production of steel plate and sheet by the steel producers. The recent trend of roll-bonded titanium-clad steel which has an excellent corrosion resistance together with the advantage in cost-saving are mainly described in this article as to the demand, production technique and new application aspects. Though the predominant usage of titanium-clad steel plate has been in power-generating plants, enlargeing utilization in the chemical plants such as terephthalic acid production plants is leading the growth in the market of titanium-clad steel plate. Also, the application of titanium-clad steel plates and sheets for the lining the marine structures is expected as one of the best solution to long-term surface protection for their outstanding corrosion resistance against sea water. (author)

  6. The prediction problems of VVER fuel element cladding failure theory

    International Nuclear Information System (INIS)

    Pelykh, S.N.; Maksimov, M.V.; Ryabchikov, S.D.

    2016-01-01

    Highlights: • Fuel cladding failure forecasting is based on the fuel load history and the damage distribution. • The limit damage parameter is exceeded, though limit stresses are not reached. • The damage parameter plays a significant role in predicting the cladding failure. • The proposed failure probability criterion can be used to control the cladding tightness. - Abstract: A method for forecasting of VVER fuel element (FE) cladding failure due to accumulation of deformation damage parameter, taking into account the fuel assembly (FA) loading history and the damage parameter distribution among FEs included in the FA, has been developed. Using the concept of conservative FE groups, it is shown that the safety limit for damage parameter is exceeded for some FA rearrangement, though the limits for circumferential and equivalent stresses are not reached. This new result contradicts the wide-spread idea that the damage parameter value plays a minor role when estimating the limiting state of cladding. The necessary condition of rearrangement algorithm admissibility and the criterion for minimization of the probability of cladding failure due to damage parameter accumulation have been derived, for using in automated systems controlling the cladding tightness.

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

    Energy Technology Data Exchange (ETDEWEB)

    Boltax, A [Westinghouse Electric Corporation, Advanced Reactor Division, Madison, PA (United States); Biancheria, A

    1977-04-01

    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)

  8. 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)

  9. 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)

  10. 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....

  11. 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.

  12. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    Science.gov (United States)

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

    2015-06-01

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

  13. Repository emplacement costs for Al-clad high enriched uranium spent fuel

    International Nuclear Information System (INIS)

    McDonell, W.R.; Parks, P.B.

    1994-01-01

    A range of strategies for treatment and packaging of Al-clad high-enriched uranium (HEU) spent fuels to prevent or delay the onset of criticality in a geologic repository was evaluated in terms of the number of canisters produced and associated repository costs incurred. The results indicated that strategies in which neutron poisons were added to consolidated forms of the U-Al alloy fuel generally produced the lowest number of canisters and associated repository costs. Chemical processing whereby the HEU was removed from the waste form was also a low cost option. The repository costs generally increased for isotopic dilution strategies, because of the substantial depleted uranium added. Chemical dissolution strategies without HEU removal were also penalized because of the inert constituents in the final waste glass form. Avoiding repository criticality by limiting the fissile mass content of each canister incurred the highest repository costs

  14. 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.

  15. Composite polymer/glass edge claddings for new Nova laser disks

    International Nuclear Information System (INIS)

    Powell, H.T.; Campbell, J.H.; Edwards, G.

    1987-01-01

    Large Nd:glass laser disks like those used in Nova require an edge cladding which absorbs at 1 μm. This cladding prevents Fresnel reflections from the edges from causing parasitic oscillations which would otherwise reduce the gain. The original Nova disks had a Cu/sup 2+/-doped phosphate glass cladding which was cast at high temperature around the circumference of the disk. Although the performance of this cladding is excellent, it was expensive to produce. Consequently, in parallel with their efforts to develop Pt inclusion-free laser glass, the authors developed a composite polymer/glass edge cladding that can be applied at greatly reduced cost. Laser disks constructed with the new cladding design show identical performance to the previous Nova disks and have been tested for hundreds of shots without degradation. The new cladding consists of absorbing glass strips which are bonded to the edges of polygonal-rather that elliptical-shaped disks. The bond is made by an --25-μm thick clear epoxy adhesive whose index of refraction matches both the laser and absorbing glass. By blending aromatic and aliphatic epoxy constituents, they achieved an index-of-refraction match within approximately +-0.003 between the epoxy and glass. The epoxy was also chosen based on its damage resistance to flashlamp light and its adhesive strength to glass. The present cladding is a major improvement over a previous experimental cladding utilizing silicone rubber as a coupling agent. Early prototypes constructed without using the presented techniques exhibited failures from both mechanisms. Delamination failures occurred which clearly showed both surface and bulk-mode parasitic oscillation. Requirements on the polymer, disk size, and Nd doping to prevent these problems are presented

  16. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Feinroth, H.

    2000-01-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  17. Corrosion behavior of duplex and reference cladding in NPP Grohnde

    International Nuclear Information System (INIS)

    Besch, O.A.; Yagnik, S.K.; Eucken, C.M.; Bradley, E.R.

    1996-01-01

    The Nuclear Fuel Industry Research (NFIR) Group undertook a lead test assembly (LTA) program in NPP Grohnde PWR in Germany to assess the corrosion performance of duplex and reference cladding. Two identical 16 by 16 LTAs, each containing 32 peripheral test rods, completed four reactor cycles, reaching a peak rod burnup of 46 MWd/kgU. The results from poolside examinations performed at the end of each cycle, together with power histories and coolant chemistry, are reported. Five different cladding materials were characterized during fabrication. The corrosion performance of the cladding materials was tracked in long-term tests in high-pressure, high-temperature autoclaves. The relative ranking of corrosion behavior in such tests corresponded well with the in-reactor corrosion performance. The extent and distribution of hydriding in duplex and reference specimens during the autoclave testing has been characterized. The in-reactor corrosion data indicate that the low-tin Zircaloy-4 reference cladding, R2, had an improved corrosion resistance compared to high-tin Zircaloy-4 reference cladding, R1. Two types of duplex cladding, D1 (Zr-2.5% Nb) and D2 (Zr-0.4% Fe-0.5% Sn), showed an even further improvement in corrosion resistance compared to R2 cladding. The third duplex cladding, D3 (Zr-4 + 1.0% Nb), had significantly less corrosion resistance, which was inferior to R1. The in-reactor and out-reactor corrosion performances have been ranked

  18. Fuel cladding mechanical properties for transient analysis

    International Nuclear Information System (INIS)

    Johnson, G.D.; Hunter, C.W.; Hanson, J.E.

    1976-01-01

    Out-of-pile simulated transient tests have been conducted on irradiated fast-reactor fuel pin cladding specimens at heating rates of 10 0 F/s (5.6 0 K/s) and 200 0 F/s (111 0 K/s) to generate mechanical property information for use in describing cladding behavior during off-normal events. Mechanical property data were then analyzed, applying the Larson-Miller Parameter to the effects of heating rate and neutron fluence. Data from simulated transient tests on TREAT-tested fuel pins demonstrate that Plant Protective System termination of 3$/s transients prevents significant damage to cladding. The breach opening produced during simulated transient testing is shown to decrease in size with increasing neutron fluence

  19. Irradiation effects on mechanical properties of fuel element cladding from thermal reactors

    International Nuclear Information System (INIS)

    Chatterjee, S.

    2005-01-01

    During reactor operation, UO 2 expands more than the cladding tube (Zirconium alloys for thermal reactors), is hotter, cracks and swells. The fuel therefore will interact with the cladding, resulting in straining of the later. To minimize the possibility of rupture of the cladding, ideally it should have good ductility as well as high strength. However, the ductility reduces with increase in fuel element burn-up. Increased burn-up also increases swelling of the fuel, leading to increased contact pressure between the fuel and the cladding tube. This would cause strains to be concentrated over localized regions of the cladding. For fuel elements burnup exceeding 40 GWd/T, the contribution of embrittlement due to hydriding, and the increased possibility of embrittlement due to stress corrosion cracking, also need to be considered. In addition to the tensile properties, the other mechanical properties of interest to the performance of cladding tube in an operating fuel element are creep rate and fatigue endurance. Irradiation is reported to have insignificant effect on high cycle endurance limit, and fatigue from fuel element vibration is most unlikely, to be life limiting. Even though creep rates due to irradiation are reported to increase by an order of magnitude, the cladding creep ductility would be so high that creep type failures in fuel element would be most improbable. Thus, the most important limiting aspect of mechanical performance of fuel element cladding has been recognized as the tensile ductility resulting from the stress conditions experienced by the cladding. Some specific fission products of threshold amount (if) deposited on the cladding, and hydride morphology (e.g. hydride lenses). The presentation will brief about irradiation damage in cladding materials and its significance, background of search for better Zirconium alloys as cladding materials, and elaborate on the types of mechanical tests need to be conducted for the evaluation of claddings

  20. 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.

  1. Semipolar III-nitride laser diodes with zinc oxide cladding.

    Science.gov (United States)

    Myzaferi, Anisa; Reading, Arthur H; Farrell, Robert M; Cohen, Daniel A; Nakamura, Shuji; DenBaars, Steven P

    2017-07-24

    Incorporating transparent conducting oxide (TCO) top cladding layers into III-nitride laser diodes (LDs) improves device design by reducing the growth time and temperature of the p-type layers. We investigate using ZnO instead of ITO as the top cladding TCO of a semipolar (202¯1) III-nitride LD. Numerical modeling indicates that replacing ITO with ZnO reduces the internal loss in a TCO clad LD due to the lower optical absorption in ZnO. Lasing was achieved at 453 nm with a threshold current density of 8.6 kA/cm 2 and a threshold voltage of 10.3 V in a semipolar (202¯1) III-nitride LD with ZnO top cladding.

  2. 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)

  3. 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

  4. Cladding failure model III (CFM III). A simple model for iodine induced stress corrosion cracking of zirconium-lined barrier and standard zircaloy cladding

    International Nuclear Information System (INIS)

    Tasooji, A.; Miller, A.K.

    1984-01-01

    A previously developed unified model (SCCIG*) for predicting iodine induced SCC in standard Zircaloy cladding was modified recently into the ''SCCIG-B'' model which predicts the stress corrosion cracking behaviour of zirconium lined barrier cladding. Several published papers have presented the capability of these models for predicting various observed behaviours related to SCC. A closed form equation, called Cladding Failure Model III (CMFIII), has been derived from the SCCIG-B model. CFMIII takes the form of an explicit equation for the radial crack growth rate dc/dt as a function of hoop strain, crack depth, temperature, and surface iodine concentration in irradiated cladding (both barrier and standard Zircaloy). CMFIII has approximately the same predictive capabilities as the physically based SCCIG and/or SCCIG-B models but is computationally faster and more convenient and can be easily utilized in fuel performance codes for predicting the behaviour of barrier and standard claddings in reactor operations. (author)

  5. CO2 laser cladding of VERSAlloyTM on carbon steel with powder feeding

    International Nuclear Information System (INIS)

    Kim, Jae-Do; Kweon, Jin-Wook

    2007-01-01

    Laser cladding processing with metal powder feeding has been experimented on carbon steel with VERSAlloy TM . A special device for the metal powder feeding was designed and manufactured. By adopting proper cladding parameters, good clad layers and sound metallurgical bonding with the base metal were obtained. Analysis indicates that the micro hardness of clad layer and the heat-affected zone increased with increasing of cladding speed. The experimental results showed that VERSAlloy TM cladded well with carbon steel

  6. Fuel clad chemical interactions in fast reactor MOX fuels

    Energy Technology Data Exchange (ETDEWEB)

    Viswanathan, R., E-mail: rvis@igcar.gov.in

    2014-01-15

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

  7. 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)

  8. High performance fuel technology development : Development of high performance cladding materials

    International Nuclear Information System (INIS)

    Park, Jeongyong; Jeong, Y. H.; Park, S. Y.

    2012-04-01

    The superior in-pile performance of the HANA claddings have been verified by the successful irradiation test and in the Halden research reactor up to the high burn-up of 67GWD/MTU. The in-pile corrosion and creep resistances of HANA claddings were improved by 40% and 50%, respectively, over Zircaloy-4. HANA claddings have been also irradiated in the commercial reactor up to 2 reactor cycles, showing the corrosion resistance 40% better than that of ZIRLO in the same fuel assembly. Long-term out-of-pile performance tests for the candidates of the next generation cladding materials have produced the highly reliable test results. The final candidate alloys were selected and they showed the corrosion resistance 50% better than the foreign advanced claddings, which is beyond the original target. The LOCA-related properties were also improved by 20% over the foreign advanced claddings. In order to establish the optimal manufacturing process for the inner and outer claddings of the dual-cooled fuel, 18 different kinds of specimens were fabricated with various cold working and annealing conditions. Based on the performance tests and various out-of-pile test results obtained from the specimens, the optimal manufacturing process was established for the inner and outer cladding tubes of the dual-cooled fuel

  9. In-pile test results of HANA claddings in Halden research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Choi, Byoung Kwon; Jeong, Yong Hwan; Jung, Yun Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    It is a kind of facing tasks in the nuclear industry to develop advanced claddings for high burn-up fuel which is safer and more economical than the existing conventional ones. Since 1997, taking an initiative in KAERI, the Zr cladding development team has carried out the R and D activities for the development of the advanced claddings to be used in the high burn-up fuel (>70,000 MWD.MTU). The team had produced the advanced claddings (HANA, High-performance Alloy for Nuclear Application) from the patented composition and manufacturing process in the international collaboration with U.S. and Japan. Now, the HANA claddings have being demonstrated their good performances from the out-of-pile tests including the corrosion, creep, burst, tensile, microstructures LOCA, RIA, wear, and so on. In parallel to the out-of-pile performance tests, the HANA claddings are being undertaken to evaluate their in-pile properties in Halden research reactor. In this study, it is included the test overviews, conditions, and results of the HANA claddings in the Halden reactor.

  10. Secondary hydriding of defected zircaloy-clad fuel rods

    International Nuclear Information System (INIS)

    Olander, D.R.; Vaknin, S.

    1993-01-01

    The phenomenon of secondary hydriding in LWR fuel rods is critically reviewed. The current understanding of the process is summarized with emphasis on the sources of hydrogen in the rod provided by chemical reaction of water (steam) introduced via a primary defect in the cladding. As often noted in the literature, the role of hydrogen peroxide produced by steam radiolysis is to provide sources of hydrogen by cladding and fuel oxidation that are absent without fission-fragment irradiation of the gas. Quantitative description of the evolution of the chemical state inside the fuel rod is achieved by combining the chemical kinetics of the reactions between the gas and the fuel and cladding with the transport by diffusion of components of the gas in the gap. The chemistry-gas transport model provides the framework into which therate constants of the reactions between the gases in the gap and the fuel and cladding are incorporated. The output of the model calculation is the H 2 0/H 2 ratio in the gas and the degree of claddingand fuel oxidation as functions of distance from the primary defect. This output, when combined with a criterion for the onset of massive hydriding of the cladding, can provide a prediction of the time and location of a potential secondary hydriding failure. The chemistry-gas transport model is the starting point for mechanical and H-in-Zr migration analyses intended to determine the nature of the cladding failure caused by the development of the massive hydride on the inner wall

  11. Design optimization of multi-layer Silicon Carbide cladding for light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho, E-mail: euo@unm.edu [Department of Nuclear Engineering, University of New Mexico, MSC01 1120 1 University of New Mexico, Albuquerque, NM 87131 (United States); NO, Hee Cheon, E-mail: hcno@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2017-01-15

    Highlights: • SiC cladding designs are optimized with a multi-layer structural analysis code. • Layer radial thickness fraction that minimizes cladding fracture probability exists. • The demonstrated procedure is applicable for multi-layer SiC cladding design. • Duplex SiC with the inner composite fraction ∼0.4 is optimal in a reference case. • Increasing composite thermal conductivity markedly decreases SiC cladding stress. - Abstract: A parametric study that demonstrates a methodology for determining the optimum bilayer composition in a duplex SiC cladding is discussed. The structural performance of multi-layer SiC cladding design is significantly affected by radial thickness fraction of each layer. This study shows that there exists an optimal composite/monolith radial thickness fraction that minimizes failure probability for a duplex SiC cladding in steady-state operation. An exemplary reference case study shows that the duplex cladding with the inner composite fraction ∼0.4 and the outer CVD-SiC fraction ∼0.6 is found to be the optimal SiC cladding design for the current PWRs with the reference material choice for CVD-SiC and fiber reinforced composite. A marginal increase in the composite fraction from the presented optimal designs may lead to increase structural integrity by introducing some unquantified merits such as increasing damage tolerance. The major factors that affect the optimum cladding designs are temperature gradients and internal gas pressure. Clad wall thickness, thermal conductivity, and Weibull modulus are among the key design parameters/material properties.

  12. 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

  13. Probabilistic assessment of spent-fuel cladding breach

    International Nuclear Information System (INIS)

    Foadian, H.; Rashid, Y.R.; Seager, K.D.

    1991-01-01

    A methodology for determining the probability spent-fuel cladding breach due to normal and accident class B cask transport conditions is introduced. This technique uses deterministic stress analysis results as well as probabilistic cladding material properties, initial flaws, and breach criteria. Best estimates are presented for the probability distributions of irradiated Zircaloy properties such as ductility and fracture toughness, and for fuel rod initial conditions such as manufacturing flaws and PCI part-wall cracks. Example analyses are used to illustrate the implementation of this methodology for a BWR (GE 7 x 7) and a PWR (B ampersand W 15 x 15) assembly. The cladding breach probabilities for each assembly are tabulated for regulatory normal and accident transport conditions including fire

  14. Probabilistic assessment of spent-fuel cladding breach

    International Nuclear Information System (INIS)

    Foadian, H.; Rashid, Y.R.; Seager, K.D.

    1992-01-01

    In this paper a methodology for determining the probability of spent-fuel cladding breach due to normal and accident class B cask transport conditions is introduced. This technique uses deterministic stress analysis results as well as probabilistic cladding material properties, initial flaws, and breach criteria. Best estimates are presented for the probability distributions of irradiated Zircaloy properties such as ductility and fracture toughness, and for fuel rod initial conditions such as manufacturing flaws and PCI part-wall cracks. Example analyses are used to illustrate the implementation of this methodology for a BWR (GE 7 x 7) and a PWR (B and W 15 x 15) assembly. The cladding breach probabilities for each assembly are tabulated for regulatory normal and accident transport conditions including fire

  15. Stainless steel clad for light water reactor fuels. Final report

    International Nuclear Information System (INIS)

    Rivera, J.E.; Meyer, J.E.

    1980-07-01

    Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

  16. Biological removal of iron and manganese in rapid sand filters - Process understanding of iron and manganese removal

    DEFF Research Database (Denmark)

    Lin, Katie

    to precipitation and corrosion. Manganese and iron can either be removed physico-chemically or biologically or combined. The physico-chemical oxidation and precipitation of manganese can theoretically be achieved by aeration, but this process is slow unless pH is raised far above neutral, making the removal...... of manganese by simple aeration and precipitation under normal drinking water treatment conditions insignificant. Manganese may also be oxidized autocatalytically. Iron is usually easier to remove. First, iron is rapidly chemically oxidized by oxygen at neutral pH followed by precipitation and filtration......-filter, where iron is removed. Step 2: Filtration in an after-filter where e.g. ammonium and manganese is removed. The treatment relies on microbial processes and may present an alternative, greener and more sustainable approach for drinking water production spending less chemicals and energy than chemical (e...

  17. Cladding failure probability modeling for risk evaluations of fast reactors

    International Nuclear Information System (INIS)

    Mueller, C.J.; Kramer, J.M.

    1987-01-01

    This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current US innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery

  18. The characteristics of anodic coating of Al-alloy claddings

    International Nuclear Information System (INIS)

    Yang Yong; Zou Benhui; Guo Hong; Du Yanhua; Bai Zhiyong; Cai Zhenfang

    2014-01-01

    Aluminum alloy claddings of research reactor fuel elements should be corroded by sodium hydroxide solution and anodized in sulfuric acid solution, but there are often some uneven color phenomena on surfaces, and sometimes regions of 'black and white stripes' appear. In order to study the relationship of colorful stripes on coatings and the surface morphology of aluminum alloy claddings corroded by sodium hydroxide solution, surface microstructures and second phase particles of the aluminum alloy claddings, which were corroded by sodium hydroxide solution, are investigated metallographically and via SEM analysis; Meanwhile, thickness, microstructure, chemical composition and construction of anodic oxidation coatings on aluminum coatings are analyzed. It is shown that: 1) the darker the surface color of corroded aluminum alloy claddings is, the darker of anodic oxidation coating; 2) there are many micro-pores on anodized oxidation coatings, which is much similar to that of corroded aluminum alloy claddings according to the morphology and distribution. So, it can be deduced that the surface morphology of anodic coatings is inherited from the corroded surfaces. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-01

    To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at ORNL and is described fully in US Patent Application 20060070455, Expanded plug method for developing circumferential mechanical properties of tubular materials. This method is designed for testing fuel rod cladding ductility in a hot cell utilizing an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are the simplicity of measuring the test component assembly in the hot cell and the direct measurement of specimen strain. It was also found that cladding strength could be determined from the test results. The basic approach of this test method is to apply an axial compressive load to a cylindrical plug of polyurethane (or other materials) fitted inside a short ring of the test material to achieve radial expansion of the specimen. The diameter increase of the specimen is used to calculate the circumferential strain accrued during the test. The other two basic measurements are total applied load and amount of plug compression (extension). A simple procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves. However, several deficiencies exist in this expanded-plug loading ring test, which will impact accuracy of test results and introduce potential shear failure of the specimen due to inherited large axial compressive stress from the expansion plug test. First of all, the highly non-uniform stress and strain distribution resulted in the gage section of the clad. To ensure reliable testing and test repeatability, the potential for highly non-uniform stress distribution or displacement/strain deformation has to be eliminated at the gage section of the specimen. Second, significant

  20. 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)

  1. 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)

  2. Development Status of Accident Tolerant Fuel Cladding for LWRs

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  3. 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.

  4. Out-of-pile test of zirconium cladding simulating reactivity initiated accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. H.; Lee, M. H.; Choi, B. K.; Bang, J. K.; Jung, Y. H. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    Mechanical properties of zirconium cladding such as Zircaloy-4 and advanced cladding were evaluated by ring tension test to simulate Reactivity-Initiated Accident (RIA) as an out-pile test. Cladding was hydrided by means of charging hydrogen up to 1000ppm to simulate high-burnup situation, finally fabricated to circumferential tensile specimen. Ring tension test was carried out from 0.01 to 1/sec to keep pace with actual RIA event. The results showed that mechanical strength of zirconium cladding increased at the value of 7.8% but ductility decreased at the 34% as applied strain rate and absorbed hydrogen increased. Further activities regarding out-of-pile testing plans for simulated high-burnup cladding were discussed in this paper.

  5. State-of-the-technology review of fuel-cladding interaction

    International Nuclear Information System (INIS)

    Bailey, W.J.; Wilson, C.L.; MacGowan, L.J.; Pankaskie, P.J.

    1977-12-01

    A literature survey and a summarization of postulated fuel-cladding-interaction mechanisms and associated supportive data are reported. The results of that activity are described in the report and include comments on experience with power-ramped fuel, fuel-cladding mechanical interaction, stress-corrosion cracking and fission-product embrittlement, potential remedial actions, fuel-cladding-interaction mechanistic considerations, other ongoing programs, and related patents of interest. An assessment of the candidate fuel concepts to be evaluated as part of this program is provided

  6. Application of YAG laser cladding to the flange seating surface

    International Nuclear Information System (INIS)

    Nakanishi, Koki; Ninomiya, Kazuyuki; Nezaki, Koji

    1999-01-01

    Stainless cladding on carbon steel is usually conducted by shielded metal arc welding (SMAW) or gas tungsten arc welding (GTAW). YAG ( Yttrium-Aluminum-Garnet) laser welding is superior to these methods of welding in the following respects : (1) The heat affected zone (HAZ) is narrower and there is less distortion. (2) YAG laser cladding has the required chemical compositions, even with possibly fewer welding layers under controlled dilution. (3) Greater welding speed. YAG laser cladding application to vessel flange seating surfaces was examined in this study and the results are discussed. The following objectives were carried out : (1) Determination of welding conditions for satisfactory cladding layers and (2) whether cladding would be adequately possible at a cornered section of a stair-like plate, assuming actual flange shape. (3) Measurement of welding distortion and heat affected zone in carbon steel. The welding conditions for producing no-crack deposit with low dilution in carbon steel were clarified and welding by which cladding at cornered section would be possible was achieved. welding distortion by YAG laser was found less than with GTAW and HAZ made by first layer welding could be tempered appropriately by second layer welding. (author)

  7. Cladding failure probability modeling for risk evaluations of fast reactors

    International Nuclear Information System (INIS)

    Mueller, C.J.; Kramer, J.M.

    1987-01-01

    This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current U.S. innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery. (orig.)

  8. 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

  9. 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

  10. Observation of advanced particle removal rates in pump limiter simulation experiments

    International Nuclear Information System (INIS)

    Goebel, D.M.; Conn, R.W.

    1984-05-01

    The performance of particle removal schemes for density and impurity control in tokamaks and mirror machines depends strongly on the plasma parameters and local recycling near the plasma neutralizier plates and gas pumping ducts. The relationship between plasma density, electron temperature, ion energy and gas flow and particle removal rate through a pumping duct located near a plasma neutralizer plate has been experimentally investigated in the steady state plasma device PISCES. Results indicate that initially the particle removal by pumps at the end of the duct is proportional to the plasma flux to the plate. A nonlinear increase in the pumping rate occurs when the ionization mean free path for neutrals from the plate becomes less than the plasma radius. The transition from a transparent to an opaque plasma due to local ionization of the neutrals produced at the neutralizer plate greatly enhances the particle removal rate by recycling of the neutral gas as it flows away from the neutralizer plate or out of the pumping ducts. Parameters were varied to determine the importance of ballistic scattering of higher energy ions from the plate, but no effects were found in these experiments

  11. 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

  12. 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

  13. 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)

  14. Fuel compliance model for pellet-cladding mechanical interaction

    International Nuclear Information System (INIS)

    Shah, V.N.; Carlson, E.R.

    1985-01-01

    This paper describes two aspects of fuel pellet deformation that play significant roles in determining maximum cladding hoop strains during pellet-cladding mechanical interaction: compliance of fragmented fuel pellets and influence of the pellet end-face design on the transmission of axial compressive force in the fuel stack. The latter aspect affects cladding ridge formation and explains several related observations that cannot be explained by the hourglassing model. An empirical model, called the fuel compliance model and representing the above aspects of fuel deformation, has been developed using the results from two Halden experiments and incorporated into the FRAP-T6 fuel performance code

  15. WWER water chemistry related to fuel cladding behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, J; Zmitko, M [Nuclear Research Inst. plc., Rez (Czech Republic); Vrtilkova, V [Nuclear Fuel Inst., Prague (Czech Republic)

    1997-02-01

    Operational experience in WWER primary water chemistry and corrosion related to the fuel cladding is reviewed. Insignificant corrosion of fuel cladding was found which is caused by good corrosion resistance of Zr1Nb material and relatively low coolant temperature at WWER-440 reactor units. The differences in water chemistry control is outlined and an attention to the question of compatibility of Zircaloys with WWER water chemistry is given. Some results of research and development in field of zirconium alloy corrosion behaviour are discussed. Experimental facility for in-pile and out-of-pile cladding material corrosion testing is shown. (author). 14 refs, 5 figs, 3 tabs.

  16. The influence of residual stresses on small through-clad cracks in pressure vessels

    International Nuclear Information System (INIS)

    deLorenzi, H.G.; Schumacher, B.I.

    1984-01-01

    The influence of cladding residual stresses on the crack driving force for shallow cracks in the wall of a nuclear pressure vessel is investigated. Thermo-elastic-plastic analyses were carried out on long axial through-clad and sub-clad flaws on the inside of the vessel. The depth of the flaws were one and three times the cladding thickness, respectively. An analysis of a semielliptical axial through-clad flaw was also performed. It was assumed that the residual stresses arise due to the difference in the thermal expansion between the cladding and the base material during the cool down from stress relieving temperature to room temperature and due to the subsequent proof test before the vessel is put into service. The variation of the crack tip opening displacement during these loadings and during a subsequent thermal shock on the inside wall is described. The analyses for the long axial flaws suggest that the crack driving force is smaller for this type of flaw if the residual stresses in the cladding are taken into account than if one assumes that the cladding has no residual stresses. However, the analysis of the semielliptical flaw shows significantly different results. Here the crack driving force is higher than when the residual stresses are not taken into account and is maximum in the cladding at or near the clad/base material interface. This suggests that the crack would propagate along the clad/base material interface before it would penetrate deeper into the wall. The elastic-plastic behavior found in the analyses show that the cladding and the residual stresses in the cladding should be taken into acocunt when evaluating the severity of shallow surface cracks on the inside of a nuclear pressure vessel

  17. 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.

  18. 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

  19. Temperature measurements of the aluminium claddings of fuel elements in nuclear reactor

    International Nuclear Information System (INIS)

    Chen Daolong

    1986-01-01

    A method for embedding the sheathed thermocouples in the aluminium claddings of some fuel elements of experimental reactors by ultrasonic welding technique is described. The measurement results of the cladding temperature of fuel elements in reactors are given. By means of this method, the joint between the sheathed thermocouples and the cladding of fuel elements can be made very tight, there are no bulges on the cladding surfaces, and the sheathed thermocouples are embedded strongly and reliably. Therefore an essential means is provided for acquiring the stable and dynamic state data of the cladding temperature of in-core fuel elements

  20. Robust cladding light stripper for high-power fiber lasers using soft metals.

    Science.gov (United States)

    Babazadeh, Amin; Nasirabad, Reza Rezaei; Norouzey, Ahmad; Hejaz, Kamran; Poozesh, Reza; Heidariazar, Amir; Golshan, Ali Hamedani; Roohforouz, Ali; Jafari, S Naser Tabatabaei; Lafouti, Majid

    2014-04-20

    In this paper we present a novel method to reliably strip the unwanted cladding light in high-power fiber lasers. Soft metals are utilized to fabricate a high-power cladding light stripper (CLS). The capability of indium (In), aluminum (Al), tin (Sn), and gold (Au) in extracting unwanted cladding light is examined. The experiments show that these metals have the right features for stripping the unwanted light out of the cladding. We also find that the metal-cladding contact area is of great importance because it determines the attenuation and the thermal load on the CLS. These metals are examined in different forms to optimize the contact area to have the highest possible attenuation and avoid localized heating. The results show that sheets of indium are very effective in stripping unwanted cladding light.

  1. 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

  2. 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.

  3. Fuel cladding tube and fuel rod for BWR type reactor

    International Nuclear Information System (INIS)

    Urata, Megumu; Mitani, Shinji.

    1995-01-01

    A fuel cladding tube has grooves fabricated, on the surface thereof, with a predetermined difference between crest and bottom (depth of the groove) in the circumferential direction. The cross sectional shape thereof is sinusoidal. The distribution of the grain size of iron crud particles in coolants is within a range about from 2μm to 12μm. If the surface roughness of the fuel cladding tube (depth of the groove) is determined greater than 1.6μm and less than 12.5, iron cruds in coolants can be positively deposited on the surface of the fuel cladding tube. In addition, once deposited iron cruds can be prevented from peeling from the surface of the fuel cladding tube. With such procedures, iron cruds deposited and radioactivated on the fuel cladding tube can be prevented from peeling, to prevent and reduce the increase of radiation dose on the surface of the pipelines without providing any additional device. (I.N.)

  4. Experimental Setup with Transient Behavior of Fuel Cladding of SFR

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    Nowadays, in Korea, advanced cladding such as FC92 is developed and its transient behaviors are required for the safety analysis of SFR. Design and safety analyses of sodium-cooled fast reactor (SFR) require understanding fuel pin responses to a wide range of off-normal events. In a loss-of-flow (LOF) or transient over-power (TOP), the temperature of the cladding is rapidly increased above its steady-state service temperature. Transient tests have been performed in sections of fuel pin cladding and a large data base has been established for austenitic stainless steel such as 20% cold-worked 316 SS and ferritic/martensitic steels such as HT9. This paper summarizes the technical status of transient testing facilities and their results. Previous researches showed the transient behaviors of HT9 cladding. For the safety analyses in SFR in Korea, simulated transient tests with newly developed FC92 as well as HT9 cladding are being carried out.

  5. Characteristics of hydride precipitation and reorientation in spent-fuel cladding

    International Nuclear Information System (INIS)

    Chung, H. M.; Strain, R. V.; Billone, M. C.

    2000-01-01

    The morphology, number density, orientation, distribution, and crystallographic aspects of Zr hydrides in Zircaloy fuel cladding play important roles in fuel performance during all phases before and after discharge from the reactor, i.e., during normal operation, transient and accident situations in the reactor, temporary storage in a dry cask, and permanent storage in a waste repository. In the past, partly because of experimental difficulties, hydriding behavior in irradiated fuel cladding has been investigated mostly by optical microscopy (OM). In the present study, fundamental metallurgical and crystallographic characteristics of hydride precipitation and reorientation were investigated on the microscopic level by combined techniques of OM and transmission electron and scanning electron microscopy (TEM and SEM) of spent-fuel claddings discharged from several boiling and pressurized water reactors (BWRs and PWRs). Defueled sections of standard and Zr-lined Zircaloy-2 fuel claddings, irradiated to fluences of ∼3.3 x 10 21 n cm -2 and ∼9.2 x 10 21 n cm -2 (E > 1 MeV), respectively, were obtained from spent fuel rods discharged from two BWRs. Sections of standard and low-tin Zircaloy-4 claddings, irradiated to fluences of ∼4.4 x 10 21 n cm -2 , ∼5.9 x 10 21 n cm -2 , and ∼9.6 x 10 21 n cm -2 (E > 1 MeV) in three PWRs, were also obtained. Microstructural characteristics of hydrides were analyzed in as-irradiated condition and after gas-pressurization-burst or expanding-mandrel tests at 292-325 C in Ar for some of the spent-fuel claddings. Analyses were also conducted of hydride habit plane, morphology, and reorientation characteristics on unirradiated Zircaloy-4 cladding that contained dense radial hydrides. Reoriented hydrides in the slowly cooled unirradiated cladding were produced by expanding-mandrel loading

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  8. 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.)

  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. Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

    Science.gov (United States)

    Degueldre, Claude; Fahy, James; Kolosov, Oleg; Wilbraham, Richard J.; Döbeli, Max; Renevier, Nathalie; Ball, Jonathan; Ritter, Stefan

    2018-05-01

    The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (steel cladding is retained despite He2+ implantation.

  11. 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.

  12. 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

  13. Semi-empirical corrosion model for Zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Nadeem Elahi, Waseem; Atif Rana, Muhammad

    2015-01-01

    The Zircaloy-4 cladding tube in Pressurize Water Reactors (PWRs) bears corrosion due to fast neutron flux, coolant temperature, and water chemistry. The thickness of Zircaloy-4 cladding tube may be decreased due to the increase in corrosion penetration which may affect the integrity of the fuel rod. The tin content and inter-metallic particles sizes has been found significantly in the magnitude of oxide thickness. In present study we have developed a Semiempirical corrosion model by modifying the Arrhenius equation for corrosion as a function of acceleration factor for tin content and accumulative annealing. This developed model has been incorporated into fuel performance computer code. The cladding oxide thickness data obtained from the Semi-empirical corrosion model has been compared with the experimental results i.e., numerous cases of measured cladding oxide thickness from UO 2 fuel rods, irradiated in various PWRs. The results of the both studies lie within the error band of 20μm, which confirms the validity of the developed Semi-empirical corrosion model. Key words: Corrosion, Zircaloy-4, tin content, accumulative annealing factor, Semi-empirical, PWR. (author)

  14. 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

  15. Laser cladding of quasicrystalline alloys

    International Nuclear Information System (INIS)

    Audebert, F.; Sirkin, H.; Colaco, R.; Vilar, R.

    1998-01-01

    Quasicrystals are a new class of ordinated structures with metastable characteristics room temperature. Quasicrystalline phases can be obtained by rapid quenching from the melt of some alloys. In general, quasicrystals present properties which make these alloys promising for wear and corrosion resistant coatings applications. During the last years, the development of quasicrystalline coatings by means of thermal spray techniques has been impulsed. However, no references have been found of their application by means of laser techniques. In this work four claddings of quasicrystalline compositions formed over aluminium substrate, produced by a continuous CO 2 laser using simultaneous powders mixture injection are presented. The claddings were characterized by X ray diffraction, scanning electron microscopy and Vickers microhardness. (Author) 18 refs

  16. Development of composite polymer-glass edge claddings for Nova Laser Disks

    International Nuclear Information System (INIS)

    Campbell, J.H.; Edwards, G.; Frick, F.A.; Gemmell, D.S.; Gim, B.M.; Jancaitis, K.S.; Jessop, E.S.; Kong, M.K.; Lyon, R.E.; Murray, J.E.; Patton, H.G.; Pitts, J.H.; Powell, H.T.; Riley, M.O.; Wallerstein, E.P.; Wolfe, C.R.; Woods, B.W.

    1988-01-01

    Large Nd:glass laser disks for disk amplifiers require an edge cladding which absorbs at 1 μ m. This cladding prevents edge reflections from causing parasitic oscillations that would otherwise deplete the gain. The authors have developed a composite polymer-glass edge cladding that consists of absorbing glass strips bonded to the edges of laser glass disks using an epoxy adhesive. The edge cladding must survive a fluence of approximately 20 J/cm 2 in a 0.5-ms pulse. Failure can occur either by decomposition of the polymer or by mechanical failure from thermal stresses which leads to bond delamination. An epoxy has been developed that gives the required damage resistance, refractive index match and processing characteristics. A slight tilt of the disk edges greatly reduces the threat from parasitic oscillations and a glass surface treatment is used to promote bond adhesion. Laser disks fabricated with this new cladding show identical gain performance to disks using conventional fused-glass cladding and have been tested for over 2000 shots (equivalent to about a 4-year lifetime on Nova) with out degradation

  17. 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

  18. Cladding for transverse-pumped solid-state laser

    Science.gov (United States)

    Byer, Robert L. (Inventor); Fan, Tso Y. (Inventor)

    1989-01-01

    In a transverse pumped, solid state laser, a nonabsorptive cladding surrounds a gain medium. A single tranverse mode, namely the Transverse Electromagnetic (TEM) sub 00 mode, is provided. The TEM sub 00 model has a cross sectional diameter greater than a transverse dimension of the gain medium but less than a transverse dimension of the cladding. The required size of the gain medium is minimized while a threshold for laser output is lowered.

  19. ZIRCONIUM-CLADDING OF THORIUM

    Science.gov (United States)

    Beaver, R.J.

    1961-11-21

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

  20. 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

  1. A regression model for zircaloy cladding in-reactor creepdown: Database, development, and assessment

    International Nuclear Information System (INIS)

    Shah, V.N.; Tolli, J.E.; Lanning, D.

    1987-01-01

    The paper presents a cladding deformation model developed to analyze cladding creepdown during steady state operation in a PWR and a BWR. This model accounts for variation in the zircaloy cladding heat treatments - cold worked and stress relieved material typically used in a PWR and fully recrystallized material typically used in a BWR. This model calculates cladding creepdown as a function of hoop stress, fast neutron flux, exposure time, and temperature. The paper also presents a comparison between cladding creep calculations by the creepdown model and corresponding test results from the KWU/CE program. ORNL HOBBIE experiments, and EPRI/Westinghouse Engineering cooperative project. The comparisons show that the creepdown model calculates cladding creep strains reasonably well. (orig./HP)

  2. Study of cladding toughness in a pressure vessel steel water reactor

    International Nuclear Information System (INIS)

    Soulat, P.; Al Mundheri, M.

    1984-12-01

    Toughness of cladding and pressure vessel steel were determined at different temperatures in order to appreciate the participation of cladding resistance against crack propagation. The toughness of cladding is comparable with typical results on austenitic welds. The test on covered CT specimens shows the possibility of having a relatively good prevision of the behaviour of a coated structure

  3. 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

    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......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...

  4. Corrosion effect of fast reactor fuel claddings on their mechanical properties

    International Nuclear Information System (INIS)

    Davydov, E.F.; Krykov, F.N.; Shamardin, V.K.

    1985-01-01

    Fast reactor fuel cladding corrosion effect on its mechanical properties was investigated. UO 2 fuel elements were irradiated in the BOP-60 reactor at the linear heat rate of 42 kw/m. Fuel cladding is made of stainless steel OKh16N15M3BR. Calculated maximum cladding temperature is 920 K. Neutron fluence in the central part of fuel elements is 6.3x10 26 m+H- 2 . To investigate the strength changes temperature dependence of corrossion depth, cladding strength reduction factors was determined. Samples plasticity reduction with corrosion layer increase is considered to be a characteristic feature

  5. 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.

  6. Study on the improvement of nuclear fuel cladding reliability

    International Nuclear Information System (INIS)

    Rheem, Karp Soon; Han, Jung Ho; Jeong, Yong Hwan; Lee, Deok Hyun

    1987-12-01

    In order to improve the nuclear fuel cladding reliability for high burn-up fuels, the corrosion resistance of laser beam surface treated and β-quenched zircaloys and the mechanical characteristics including fatigue, burst, and out-of-pile PCMI characteristics of heat treated zircaloys were investigated. In addition, the inadiation characteristics of Ko-Ri reactor fuel claddings was examined. It was found that the wasteside corrosion resistance of commercial zircaloys was improved remarkably by laser beam surface treatment. The out-of-pile transient cladding failures were investigated in terms of hoop stress versus time-to-failures by means of mandrel loading units at 25 deg C and 325 deg C. Fatigue characteristics of the β-quenched and as-received zircaloy cladding were investigated by using an internal oil pressurization method which can simulate the load-following operation cycle. The results were in good agreement with the existing data obtained by conventional methods for commercial zircaloys. Burst tests were performed with commercial and the β-quenched zircaloys in high pressure argon gas atmosphere as a function of burst temperature. The burst stress decreased linearly in the α phase region up to 600 deg C and hereafter the decrement of the burst stress decreased gradually with temperature in the β-phase region. For the first time, the burst characteristic of the irradiated zircaloy-4 cladding tubes released from Ko-Ri nuclear power unit 1 was investigated, and attempts were made to trace the cause of cladding failures by examining the failed structure and fret marks by debris. (Author)

  7. 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)

  8. Kilowatt-level cladding light stripper for high-power fiber laser.

    Science.gov (United States)

    Yan, Ping; Sun, Junyi; Huang, Yusheng; Li, Dan; Wang, Xuejiao; Xiao, Qirong; Gong, Mali

    2017-03-01

    We designed and fabricated a high-power cladding light stripper (CLS) by combining a fiber-etched CLS with a cascaded polymer-recoated CLS. The etched fiber reorganizes the numerical aperture (NA) distribution of the cladding light, leading to an increase in the leakage power and a flatter distribution of the leakage proportion in the cascaded polymer-recoated fiber. The index distribution of the cascaded polymer-recoated fiber is carefully designed to ensure an even leakage of cladding light. More stages near the index of 1.451 are included to disperse the heat. The CLS is capable of working consistently under 1187 W of cladding light with an attenuation of 26.59 dB, and the highest local temperature is less than 35°C.

  9. 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

  10. Analysis of the stress raising action of flaws in laser clad deposits

    International Nuclear Information System (INIS)

    Alam, M.M.; Kaplan, A.F.H.; Tuominen, J.; Vuoristo, P.; Miettinen, J.; Poutala, J.; Näkki, J.; Junkala, J.; Peltola, T.; Barsoum, Z.

    2013-01-01

    Highlights: ► Laser clad defects are 0D-pores/inclusions, 1D-clad waviness or 2D-planar defects. ► Surface pore of laser clad bar initiates fatigue cracks. ► Side edge surface pores are more critical than in-clad surface pores. ► Smaller notch radius and angle of as-laser clad surface raises stress significantly. ► Planar inner defects grow faster towards surface. - Abstract: Fatigue cracking of laser clad cylindrical and square section bars depends upon a variety of factors. This paper presents Finite Element Analysis (FEA) of the different macro stress fields generated as well as stress raisers created by laser cladding defects for four different fatigue load conditions. As important as the defect types are their locations and orientations, categorized into zero-, one- and two-dimensional defects. Pores and inclusions become critical close to surfaces. The performance of as-clad surfaces can be governed by the sharpness of surface notches and planar defects like hot cracks or lack-of-fusion (LOF) are most critical if oriented vertically, transverse to the bar axis. The combination of the macro stress field with the defect type and its position and orientation determines whether it is the most critical stress raiser. Based on calculated cases, quantitative and qualitative charts were developed as guidelines to visualize the trends of different combinations

  11. Iodine induced stress corrosion cracking of zircaloy cladding tubes

    International Nuclear Information System (INIS)

    Brunisholz, L.; Lemaignan, C.

    1984-01-01

    Iodine is considered as one of the major fission products responsible for PCI failure of Zry cladding by stress corrosion cracking (SCC). Usual analysis of SCC involves both initiation and growth as sequential processes. In order to analyse initiation and growth independently and to be able to apply the procedures of fracture mechanics to the design of cladding, with respect to SCC, stress corrosion tests of Zry cladding tubes were undertaken with a small fatigue crack (approx. 200 μm) induced in the inner wall of each tube before pressurization. Details are given on the techniques used to induce the fatigue crack, the pressurization test procedure and the results obtained on stress releaved or recrystallized Zry 4 tubings. It is shown that the Ksub(ISCC) values obtained during these experiments are in good agreement with those obtained from large DCB fracture mechanics samples. Conclusions will be drawn on the applicability of linear elastic fracture mechanics (LEFM) to cladding design and related safety analysis. The work now underway is aimed at obtaining better understanding of the initiation step. It includes the irradiation of Zry samples with heavy ions to simulate the effect of recoil fragments implanted in the inner surface of the cladding, that could create a brittle layer of about 10 μm

  12. Mechanical performance of SiC three-layer cladding in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Angelici Avincola, Valentina, E-mail: valentina.avincola@kit.edu [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Guenoun, Pierre, E-mail: pguenoun@mit.edu [Massachusetts Institute of Technology, 77 Massachusetts Ave, Cambridge, MA 02139 (United States); Shirvan, Koroush, E-mail: kshirvan@mit.edu [Massachusetts Institute of Technology, 77 Massachusetts Ave, Cambridge, MA 02139 (United States)

    2016-12-15

    Highlights: • FEA calculations of the stress distribution in SiC three-layer cladding. • Simulation of SiC mechanical performance under operation and accident conditions. • Failure probability analysis of SiC in steady-state and accident conditions. - Abstract: The silicon carbide cladding concept is currently under investigation with regard to increasing the accident tolerance and economic performance of light-water reactor fuels. In this work, the stress fields in the multi-layered silicon carbide cladding for LWR fuels are calculated using the commercial finite element analysis software ADINA. The material properties under irradiation are implemented as a function of temperature. The cladding is studied under operating and accident conditions, specifically for the loss-of-coolant accident (LOCA). During the LOCA, the blowdown and the reflood phases are modeled, including the quench waterfront. The calculated stresses along the cladding thickness show a high sensitivity to the assumptions regarding material properties. The resulting stresses are compared with experimental data and the probability of failure is calculated considering a Weibull model.

  13. 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)

  14. Laser cladding of Colmonoy 6 powder on AISI316L austenitic stainless steel

    International Nuclear Information System (INIS)

    Zhang, H.; Shi, Y.; Kutsuna, M.; Xu, G.J.

    2010-01-01

    Stainless steels are widely used in nuclear power plant due to their good corrosion resistance, but their wear resistance is relatively low. Therefore, it is very important to improve this property by surface treatment. This paper investigates cladding Colmonoy 6 powder on AISI316L austenitic stainless steel by CO 2 laser. It is found that preheating is necessary for preventing cracking in the laser cladding procedure and 450 o C is the proper preheating temperature. The effects of laser power, traveling speed, defocusing distance, powder feed rate on the bead height, bead width, penetration depth and dilution are investigated. The friction and wear test results show that the friction coefficient of specimens with laser cladding is lower than that of specimens without laser cladding, and the wear resistance of specimens has been increased 53 times after laser cladding, which reveals that laser cladding layer plays roles on wear resistance. The microstructures of laser cladding layer are composed of Ni-rich austenitic, boride and carbide.

  15. Cladding of Advanced Al Alloys Employing Friction Stir Welding

    NARCIS (Netherlands)

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

    2013-01-01

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

  16. Characterization of SiC–SiC composites for accident tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Deck, C.P., E-mail: Christian.Deck@ga.com; Jacobsen, G.M.; Sheeder, J.; Gutierrez, O.; Zhang, J.; Stone, J.; Khalifa, H.E.; Back, C.A.

    2015-11-15

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

  17. Analysis of fuel cladding chemical interaction in mixed oxide fuel pins

    International Nuclear Information System (INIS)

    Weber, J.W.; Dutt, D.S.

    1976-01-01

    An analysis is presented of the observed interaction between mixed oxide 75 wt percent UO 2 --25 wt percent PuO 2 fuel and 316--20 percent CW stainless steel cladding in LMFBR type fuel pins irradiated in EBR-II. A description is given of the test pins and their operating conditions together with, metallographic observations and measurements of the fuel/cladding reaction, and a correlation equation is developed relating depth of cladding attack to temperature and burnup. Some recent data on cladding reaction in fuel pins with low initial O/M in the fuel are given and compared with the correlation equation curves

  18. 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

  19. Crack resistance curves determination of tube cladding material

    Energy Technology Data Exchange (ETDEWEB)

    Bertsch, J. [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland)]. E-mail: johannes.bertsch@psi.ch; Hoffelner, W. [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland)

    2006-06-30

    Zirconium based alloys have been in use as fuel cladding material in light water reactors since many years. As claddings change their mechanical properties during service, it is essential for the assessment of mechanical integrity to provide parameters for potential rupture behaviour. Usually, fracture mechanics parameters like the fracture toughness K {sub IC} or, for high plastic strains, the J-integral based elastic-plastic fracture toughness J {sub IC} are employed. In claddings with a very small wall thickness the determination of toughness needs the extension of the J-concept beyond limits of standards. In the paper a new method based on the traditional J approach is presented. Crack resistance curves (J-R curves) were created for unirradiated thin walled Zircaloy-4 and aluminium cladding tube pieces at room temperature using the single sample method. The procedure of creating sharp fatigue starter cracks with respect to optical recording was optimized. It is shown that the chosen test method is appropriate for the determination of complete J-R curves including the values J {sub 0.2} (J at 0.2 mm crack length), J {sub m} (J corresponding to the maximum load) and the slope of the curve.

  20. Optical Cladding Waveguides in Dielectric Crystals Produced by Femtosecond Laser Inscription

    Directory of Open Access Journals (Sweden)

    Chen Feng

    2013-11-01

    Full Text Available In this work, the recent progress of our research on optical cladding waveguides in dielectric crystals produced by femtosecond laser inscription has been overviewed. With different scales at cross sections, the cladding waveguides support guidance from single mode to highly multi-modes, and work for wavelength till mid-infrared regimes. Applications of the fabricated cladding structures as new integrated light sources are introduced.

  1. 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.

  2. Study on microstructure and high temperature wear resistance of laser cladded nuclear valve clack

    International Nuclear Information System (INIS)

    Zhang Chunliang; Chen Zichen

    2002-01-01

    Laser cladding of Co-base alloy on the nuclear valve-sealing surface are performed with a 5 kW CO 2 transverse flowing laser. The microstructure and the high temperature impact-slide wear resistance of the laser cladded coating and the plasma cladded coating are studied. The results show that the microstructure, the dilution rate and the high temperature impact-slide wear resistance of the laser cladded coating have obvious advantages over the spurt cladding processing

  3. Development of Silicide Coating on Molybdenum Alloy Cladding

    International Nuclear Information System (INIS)

    Lim, Woojin; Ryu, Ho Jin

    2015-01-01

    The molybdenum alloy is considered as one of the accident tolerant fuel (ATF) cladding materials due to its high temperature mechanical properties. However, molybdenum has a weak oxidation resistance at elevated temperatures. To modify the oxidation resistance of molybdenum cladding, silicide coating on the cladding is considered. Molybdenum silicide layers are oxidized to SiO 2 in an oxidation atmosphere. The SiO 2 protective layer isolates the substrate from the oxidizing atmosphere. Pack cementation deposition technique is widely adopted for silicide coating for molybdenum alloys due to its simple procedure, homogeneous coating quality and chemical compatibility. In this study, the pack cementation method was conducted to develop molybdenum silicide layers on molybdenum alloys. It was found that the Mo 3 Si layer was deposited on substrate instead of MoSi 2 because of short holding time. It means that through the extension of holding time, MoSi 2 layer can be formed on molybdenum substrate to enhance the oxidation resistance of molybdenum. The accident tolerant fuel (ATF) concept is to delay the process following an accident by reducing the oxidation rate at high temperatures and to delay swelling and rupture of fuel claddings. The current research for Atf can be categorized into three groups: First, modification of existing zirconium-based alloy cladding by improving the high temperature oxidation resistance and strength. Second, replacing Zirconium based alloys with alternative metallic materials such as refractory elements with high temperature oxidation resistance and strength. Third, designing alternative fuel structures using ceramic and composite systems

  4. 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.)

  5. Asymptotic Method for Cladding Stress Evaluation in PCMI

    International Nuclear Information System (INIS)

    Kim, Hyungkyu; Kim, Jaeyong; Yoon, Kyungho; Lee, Kanghee; Kang, Heungseok

    2014-01-01

    A PCMI (Pellet Cladding Mechanical Interaction) failure was first reported in the GETR (General Electric Test Reactor) at Vacellitos in 1963, and such failures are still occurring. Since the high stress values in the cladding tube has been of a crucial concern in PCMI studies, there have been many researches on the stress analysis of a cladding tube pressed by a pellet. Typical works can be found in some references. It has often been assumed, however, that the cracks in the pellet were equally spaced and the pellet was a rigid body. In addition, the friction coefficient was arbitrarily chosen so that a slipping between the pellets and cladding tube could not be logically defined. Moreover, the stress intensification due to the sharp edge of a pellet fragment has never been realistically considered. These problems above drove us to launch a framework of a PCMI study particularly on stress analysis technology to improve the present analysis method incorporating the actual PCMI conditions such as the stress intensification, arbitrary distribution of the pellet cracks, material properties (esp. pellet) and slipping behavior of the pellet/cladding interface. As a first step of this work, this paper introduces an asymptotic method that was originally developed for a stress analysis in the vicinity of a sharp notch of a homogeneous body. The intrinsic reason for applying this method is to simulate the stress singularity that is expected to take place at the sharp edge of a pellet fragment due to cracking during irradiation. As a first attempt of this work, an eigenvalue problem is formulated in the case of adhered contact, and the generalized stress intensity factors are defined and evaluated. Although some works obviously remain to be accomplished, for the present framework on the PCMI analysis (e. g., slipping behaviour, contact force etc.), it was addressed that the asymptotic method can produce the stress values that cause the cladding tube failure in PCMI more

  6. 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

  7. Clad-coolant chemical interaction

    International Nuclear Information System (INIS)

    Iglesias, F.C.; Lewis, B.J.; Desgranges, C.; Toffolon, C.

    2015-01-01

    This paper provides an overview of the kinetics for zircaloy clad oxidation behaviour in steam and air during reactor accident conditions. The generation of chemical heat from metal/water reaction is considered. Low-temperature oxidation of zircaloy due to water-side corrosion is further described. (authors)

  8. 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

  9. The effect of laser process parameters on microstructure and dilution rate of cladding coatings

    Science.gov (United States)

    Bin, Liu; Heping, Liu; Xingbin, Jing; Yuxin, Li; Peikang, Bai

    2018-02-01

    In order to broaden the range of application of Q235 steel, it is necessary to repair the surface of steel. High performance 316L stainless steel coating was successfully obtained on Q235 steel by laser cladding technology. The effect of laser cladding parameters on the geometrical size and appearance of single cladding layer was investigated. The experimental results show that laser current has an important influence on the surface morphology of single channel cladding. When the current is from 155A to 165A, the cladding coating becomes smooth. The laser current has an effect on the geometric cross section size and dilution rate of single cladding. The results revealed that with the rising of laser current, the width, height and depth of layer increase gradually. With the rising of laser current, the dilution rate of cladding layer is gradually increasing.

  10. Corrosion behaviour of zircaloy 4 fuel rod cladding in EDF power plants

    Energy Technology Data Exchange (ETDEWEB)

    Romary, H; Deydier, D [EDF, Direction de l` Equipment SEPTEN, Villeurbanne (France)

    1997-02-01

    Since the beginning of the French nuclear program, a surveillance of fuel has been carried out in order to evaluate the fuel behaviour under irradiation. Until now, nuclear fuels provided by suppliers have met EDF requirements concerning fuel behaviour and reliability. But, the need to minimize the costs and to increase the flexibility of the power plants led EDF to the definition of new targets: optimization of the core management and fuel cycle economy. The fuel behaviour experience shows that some of these new requirements cannot be fully fulfilled by the present standard fuel due to some technological limits. Particularly, burnup enhancement is limited by the oxidation and the hydriding of the Zircaloy 4 fuel rod cladding. Also, fuel suppliers and EDF need to have a better knowledge of the Zy-4 cladding behaviour in order to define the existing margins and the limiting factors. For this reason, in-reactor fuel characterization programs have been set up by fuel suppliers and EDF for a few years. This paper presents the main results and conclusions of EDF experience on Zy-4 in-reactor corrosion behaviour. Data obtained from oxide layer or zirconia thickness measurements show that corrosion performance of Zy-4 fuel rod cladding, as irradiated until now in EDF reactors, is satisfactory but not sufficient to meet the future needs. The fuel suppliers propose in order to improve the corrosion resistance of fuel rod cladding, low tin Zy-4 cladding and then optimized Zy-4 cladding. Irradiation of these claddings are ongoing. The available corrosion data show the better in-reactor corrosion resistance of optimized Zy-4 fuel rod cladding compared to the standard Zy-4 cladding. The scheduled fuel surveillance program will confirm if the optimized Zy-4 fuel rod cladding will meet the requirements for the future high burnup and high flexibility fuel. (author). 10 refs, 19 figs, 4 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  12. Oxidation Behavior of FeCrAl -coated Zirconium Cladding prepared by Laser Coating

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Il-Hyun; Kim, Hyun-Gil; Choi, Byung-Kwan; Park, Jeong-Yong; Koo, Yang-Hyun; Kim, Jin-Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    From the recent research trends, the ATF cladding concepts for enhanced accident tolerance are divided as follows: Mo-Zr cladding to increase the high temperature strength, cladding coating to increase the high temperature oxidation resistance, FeCrAl alloy and SiC/SiCf material to increase the oxidation resistance and strength at high temperature. To commercialize the ATF cladding concepts, various factors are considered, such as safety under normal and accident conditions, economy for the fuel cycle, and developing development challenges, and schedule. From the proposed concepts, it is known that the cladding coating, FeCrAl alloy, and Zr-Mo claddings are considered as a near/mid-term application, whereas the SiC material is considered as a long-term application. Among them, the benefit of cladding coating on Zr-based alloys is the fuel cycle economy regarding the manufacturing, neutron cross section, and high tritium permeation characteristics. However, the challenge of cladding coating on Zr-based alloys is the lower oxidation resistance and mechanical strength at high-temperature than other concepts. Another important point is the adhesion property between the Zr-based alloy and coating materials. A laser coating method supplied with FeCrAl powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a FeCrAl-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured.

  13. Finite Element Analysis of Laser Engineered Net Shape (LENS™) Tungsten Clad Squeeze Pins

    Science.gov (United States)

    Sakhuja, Amit; Brevick, Jerald R.

    2004-06-01

    In the aluminum high-pressure die-casting and indirect squeeze casting processes, local "squeeze" pins are often used to minimize internal solidification shrinkage in heavy casting sections. Squeeze pins frequently fail in service due to molten aluminum adhering to the H13 tool steel pins ("soldering"). A wide variety of coating materials and methods have been developed to minimize soldering on H13. However, these coatings are typically very thin, and experience has shown their performance on squeeze pins is highly variable. The LENS™ process was employed in this research to deposit a relatively thick tungsten cladding on squeeze pins. An advantage of this process was that the process parameters could be precisely controlled in order to produce a satisfactory cladding. Two fixtures were designed and constructed to enable the end and outer diameter (OD) of the squeeze pins to be clad. Analyses were performed on the clad pins to evaluate the microstructure and chemical composition of the tungsten cladding and the cladding-H13 substrate interface. A thermo-mechanical finite element analysis (FEA) was performed to assess the stress distribution as a function of cladding thickness on the pins during a typical casting thermal cycle. FEA results were validated via a physical test, where the clad squeeze pins were immersed into molten aluminum. Pins subjected to the test were evaluated for thermally induced cracking and resistance to soldering of the tungsten cladding.

  14. Finite element analysis of laser engineered net shape (LENSTM) tungsten clad squeeze pins

    International Nuclear Information System (INIS)

    Sakhuja, Amit; Brevick, Jerald R.

    2004-01-01

    In the aluminum high-pressure die-casting and indirect squeeze casting processes, local 'squeeze' pins are often used to minimize internal solidification shrinkage in heavy casting sections. Squeeze pins frequently fail in service due to molten aluminum adhering to the H13 tool steel pins ('soldering'). A wide variety of coating materials and methods have been developed to minimize soldering on H13. However, these coatings are typically very thin, and experience has shown their performance on squeeze pins is highly variable. The LENS TM process was employed in this research to deposit a relatively thick tungsten cladding on squeeze pins. An advantage of this process was that the process parameters could be precisely controlled in order to produce a satisfactory cladding. Two fixtures were designed and constructed to enable the end and outer diameter (OD) of the squeeze pins to be clad. Analyses were performed on the clad pins to evaluate the microstructure and chemical composition of the tungsten cladding and the cladding-H13 substrate interface. A thermo-mechanical finite element analysis (FEA) was performed to assess the stress distribution as a function of cladding thickness on the pins during a typical casting thermal cycle. FEA results were validated via a physical test, where the clad squeeze pins were immersed into molten aluminum. Pins subjected to the test were evaluated for thermally induced cracking and resistance to soldering of the tungsten cladding

  15. Deep surface rolling for fatigue life enhancement of laser clad aircraft aluminium alloy

    Energy Technology Data Exchange (ETDEWEB)

    Zhuang, W., E-mail: wyman.zhuang@dsto.defence.gov.au [Aerospace Division, Defence Science and Technology Organisation, 506 Lorimer Street, Fishermans Bend, Victoria 3207 (Australia); Liu, Q.; Djugum, R.; Sharp, P.K. [Aerospace Division, Defence Science and Technology Organisation, 506 Lorimer Street, Fishermans Bend, Victoria 3207 (Australia); Paradowska, A. [Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW 2232 (Australia)

    2014-11-30

    Highlights: • Deep surface rolling as a post-repair enhancement technology was applied to the laser cladded 7075-T651 aluminium alloy specimens that simulated corrosion damage blend-out repair. • The residual stresses induced by the deep surface rolling process were measured. • The deep surface rolling process can introduce deep and high magnitude compressive residual stresses beyond the laser clad and substrate interface. • Spectrum fatigue test showed the fatigue life was significantly increased by deep surface rolling. - Abstract: Deep surface rolling can introduce deep compressive residual stresses into the surface of aircraft metallic structure to extend its fatigue life. To develop cost-effective aircraft structural repair technologies such as laser cladding, deep surface rolling was considered as an advanced post-repair surface enhancement technology. In this study, aluminium alloy 7075-T651 specimens with a blend-out region were first repaired using laser cladding technology. The surface of the laser cladding region was then treated by deep surface rolling. Fatigue testing was subsequently conducted for the laser clad, deep surface rolled and post-heat treated laser clad specimens. It was found that deep surface rolling can significantly improve the fatigue life in comparison with the laser clad baseline repair. In addition, three dimensional residual stresses were measured using neutron diffraction techniques. The results demonstrate that beneficial compressive residual stresses induced by deep surface rolling can reach considerable depths (more than 1.0 mm) below the laser clad surface.

  16. Deep surface rolling for fatigue life enhancement of laser clad aircraft aluminium alloy

    International Nuclear Information System (INIS)

    Zhuang, W.; Liu, Q.; Djugum, R.; Sharp, P.K.; Paradowska, A.

    2014-01-01

    Highlights: • Deep surface rolling as a post-repair enhancement technology was applied to the laser cladded 7075-T651 aluminium alloy specimens that simulated corrosion damage blend-out repair. • The residual stresses induced by the deep surface rolling process were measured. • The deep surface rolling process can introduce deep and high magnitude compressive residual stresses beyond the laser clad and substrate interface. • Spectrum fatigue test showed the fatigue life was significantly increased by deep surface rolling. - Abstract: Deep surface rolling can introduce deep compressive residual stresses into the surface of aircraft metallic structure to extend its fatigue life. To develop cost-effective aircraft structural repair technologies such as laser cladding, deep surface rolling was considered as an advanced post-repair surface enhancement technology. In this study, aluminium alloy 7075-T651 specimens with a blend-out region were first repaired using laser cladding technology. The surface of the laser cladding region was then treated by deep surface rolling. Fatigue testing was subsequently conducted for the laser clad, deep surface rolled and post-heat treated laser clad specimens. It was found that deep surface rolling can significantly improve the fatigue life in comparison with the laser clad baseline repair. In addition, three dimensional residual stresses were measured using neutron diffraction techniques. The results demonstrate that beneficial compressive residual stresses induced by deep surface rolling can reach considerable depths (more than 1.0 mm) below the laser clad surface

  17. 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)

  18. 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)

  19. Hollow cylindrical plasma filament waveguide with discontinuous finite thickness cladding

    International Nuclear Information System (INIS)

    Alshershby, Mostafa; Hao Zuoqiang; Lin Jingquan

    2013-01-01

    We have explored here a hollow cylindrical laser plasma multifilament waveguide with discontinuous finite thickness cladding, in which the separation between individual filaments is in the range of several millimeters and the waveguide cladding thickness is in the order of the microwave penetration depth. Such parameters give a closer representation of a realistic laser filament waveguide sustained by a long stable propagation of femtosecond (fs) laser pulses. We report how the waveguide losses depend on structural parameters like normalized plasma filament spacing, filament to filament distance or pitch, normal spatial frequency, and radius of the plasma filament. We found that for typical plasma parameters, the proposed waveguide can support guided modes of microwaves in extremely high frequency even with a cladding consisting of only one ring of plasma filaments. The loss of the microwave radiation is mainly caused by tunneling through the discontinuous finite cladding, i.e., confinement loss, and is weakly dependent on the plasma absorption. In addition, the analysis indicates that the propagation loss is fairly large compared with the loss of a plasma waveguide with a continuous infinite thickness cladding, while they are comparable when using a cladding contains more than one ring. Compared to free space propagation, this waveguide still presents a superior microwave transmission to some distance in the order of the filamentation length; thus, the laser plasma filaments waveguide may be a potential channel for transporting pulsed-modulated microwaves if ensuring a long and stable propagation of fs laser pulses.

  20. 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

  1. 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.)

  2. Development and application of preventive maintenance technique for pipes using laser cladding method

    International Nuclear Information System (INIS)

    Hatakenaka, Hiroaki; Yamadera, Masao; Shiraiwa, Takanori.

    1995-01-01

    A laser cladding method which produces a highly corrosion-resisting coating (cladding) on the surface of the material was developed for the purpose of preventing stress corrosion cracking (SCC) in the austenitic stainless steel (Type 304). In this method, metallic powder paste is applied on the inner surface of pipes, and then a YAG laser beam is irradiated to the paste, which melts and forms a clad with excellent corrosion resistance. Recently, the laser cladding method was practically and successfully applied to the actual nuclear power plant in Japan. This report describes this laser cladding technique, the equipment, and actual works in the field. (author)

  3. 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

  4. 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)

  5. Fracture properties of hydrided Zircaloy-4 cladding in recrystallization and stress-relief anneal conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Hsiao-Hung, E-mail: hhhsu@iner.gov.tw [Institute of Nuclear Energy Research (INER), Lungtan Township, Taoyuan County 325, Taiwan (China); Institute of Materials Engineering, National Taiwan Ocean University, Keelung 202, Taiwan (China); Tsay, Leu-Wen [Institute of Materials Engineering, National Taiwan Ocean University, Keelung 202, Taiwan (China)

    2012-03-15

    In this work, the stress-relieved (SRA) and recrystallized (RXA) Zircaloy-4 cladding specimens were hydrogen-charged to the target concentration of 300 wppm and then manufactured into X-specimens for fracture toughness test. The hydrogen embrittlement susceptibility of Zircaloy-4 cladding specimens in both SRA and RXA conditions were investigated. At the hydrogen concentration level of 300 wppm, J-integral values for RXA cladding were higher than those for SRA cladding at both 25 Degree-Sign C and 300 Degree-Sign C. The formation of brittle zirconium hydrides had a significant impact on the fracture toughness of Zircaloy-4 cladding in both SRA and RXA states, especially at 25 Degree-Sign C. Among all the tests, SRA cladding tested at 25 Degree-Sign C exhibited a great loss of the fracture toughness. The micrographic and fractographic observations further demonstrated that the fracture toughness of Zircaloy-4 cladding would be improved by the coarse grains in RXA cladding, but degraded by zirconium hydrides precipitated along the grain boundary.

  6. Research on Microstructure and Property of TiC-Co Composite Material Made by Laser Cladding

    Science.gov (United States)

    Zhang, Wei

    The experiment of laser cladding on the surface of 2Cr13 steel was made. Titanium carbide (TiC) powder and Co-base alloy powder were used as cladding material. The microstructure and property of laser cladding layer were tested. The research showed that laser cladding layer had better properties such as minute crystals, deeper layer, higher hardness and good metallurgical bonding with base metal. The structure of cladding was supersaturated solid solution with dispersed titanium carbide. The average hardness of cladding zone was 660HV0.2. 2Cr13 steel was widely used in the field of turbine blades. Using laser cladding, the good wear layer would greatly increase the useful life of turbine blades.

  7. The fuel-cladding interfacial friction coefficient in water-cooled reactor fuel rods

    International Nuclear Information System (INIS)

    Smith, E.

    1979-01-01

    A central problem in the development of cladding failure criteria and of effective operational, design or material remedies is to know whether the cladding stress is enhanced significantly near cladding ridges, pellet chips or fuel pellet cracks; the latter may also be coincident with cladding ridges at pellet-pellet interfaces. As regards the fuel pellet crack source of cladding stress concentration, the magnitude of the uranium dioxide-Zircaloy interfacial friction coefficient μ governs the magnitude and distribution of the enhanced cladding stress. Considerable discussion, particularly at a Post-Conference Seminar associated with the SMIRT 4 Conference, has focussed on the value of μ, the author taking the view that it is unlikely to be large (< 0.5). The reasoning behind this view is as follows. A fuel pellet should fracture during a power ramp when the tensile hoop stress within the pellet exceeds the fuel's fracture stress. Since the preferred position for a fuel pellet crack to form is at the fuel-cladding interface midway between existing fuel cracks, where the interfacial shear stress changes sign, the pellet segment size after a power ramp provides a limit to the magnitude of the interfacial shear stresses and consequently to the value of μ. With this argument as a basis, the author's early work used the Gittus fuel rod model, in which there is a symmetric distribution of fuel pellet cracks and symmetric interfacial slippage, to show that μ < 0.5 if it is assumed that the average hoop stress within the cladding attains yield levels. It was therefore suggested that a high interfacial friction coefficient is unlikely to be operative during a power ramp; this result was used to support the view that interfacial friction effects do not play a dominant role in stress corrosion crack formation within the cladding. (orig.)

  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......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...

  9. Effects of the inner mould material on the aluminium–316L stainless steel explosive clad pipe

    International Nuclear Information System (INIS)

    Guo, Xunzhong; Tao, Jie; Wang, Wentao; Li, Huaguan; Wang, Chen

    2013-01-01

    Highlights: ► Different mould materials were adopted to evaluate the effect of the constraint on the clad quality. ► The interface characteristics of clad pipe were analyzed for the different clad pipe. ► The clad pipes possess excellent bonding quality. - Abstract: The clad pipe played an important part in the pipeline system of the nuclear power industry. To prepare the clad pipe with even macrosize and excellent bonding quality, in this work, different mould materials were adopted to evaluate the effect of the constraint on the clad quality of the bimetal pipe prepared by explosive cladding. The experiment results indicated that, the dimension uniformity and bonding interface of clad pipe were poor by using low melting point alloy as mould material; the local bulge or the cracking of the clad pipe existed when the SiC powder was utilized. When the steel mould was adopted, the outer diameter of the clad pipe was uniform from head to tail. In addition, the metallurgical bonding was formed. Furthermore, the results of shear test, bending test and flattening test showed that the bonding quality was excellent. Therefore, the Al–316L SS clad pipe could endure the second plastic forming

  10. Embedded cladding surface thermocouples on Zircaloy-sheathed heater rods

    International Nuclear Information System (INIS)

    Wilkins, S.C.

    1977-06-01

    Titanium-sheathed Type K thermocouples embedded in the cladding wall of zircaloy-sheathed heater rods are described. These thermocouples constitute part of a program intended to characterize the uncertainty of measurements made by surface-mounted cladding thermocouples on nuclear fuel rods. Fabrication and installation detail, and laboratory testing of sample thermocouple installations are included

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

    International Nuclear Information System (INIS)

    Sindelar, R.L.; Peacock, H.B. Jr.; Lam, P.S.; Iyer, N.C.; Louthan, M.R. Jr.; Murphy, J.R.

    1996-01-01

    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

  12. 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

  13. General considerations on the oxide fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    Pascard, R.

    1977-01-01

    Since the very first experimental irradiations in thermal reactors, performed in view of the future Rapsodie fuel general study, corrosion cladding anomalies were observed. After 10 years of Rapsodie and more than two years of Phenix, performance brought definite confirmation of the chemical reactions between the irradiated fuel and cladding. That is the reason for which the fuel designers express an urgent need for determining the corrosion rates. Semi-empirical laws and mechanisms describing corrosion processes are proposed. Erratic conditions for appearance of the oxide-cladding corrosion are stressed upon. Obviously such a problem can be fully appreciated only by a statistical approach based on a large number of observations on the true LMFBR fuel pins

  14. Laser cladding of Ti-6Al-4V with various carbide powders

    International Nuclear Information System (INIS)

    Folkes, J.A.; Shibata, K.

    1994-01-01

    Laser cladding Ti-6Al-4V can be achieved with various weight percentages of different carbide powders. The microstructure and morphology of the clad layer is determined by the cladding powder composition, for a given set of laser parameters, such that 10 and 20 wt% Cr 3 C 2 results in a β + TiC clad microstructure; 10 and 20 wt% WC results in an α + TiC clad microstructure (plus some original WC); and Mo 2 C gives an α + β + TiC or β + TiC structure, depending on the weight percentage of Mo 2 C. The morphology of the TiC in all cases is dendritic or feathery, depending on the carbide content. The microstructure observed in all cases agreed well with that theoretically predicted from the energetics of carbide formation and β-stabilizing properties of each element

  15. The influence of hydride on fracture toughness of recrystallized Zircaloy-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Hsiao-Hung, E-mail: 175877@mail.csc.com.tw [Institute of Nuclear Energy Research (INER), Lungtan Township, Taoyuan County 32546, Taiwan, ROC (China); China Steel Corporation, Hsiao Kang District, Kaohsiung 81233, Taiwan, ROC (China); Chiang, Ming-Feng [China Steel Corporation, Hsiao Kang District, Kaohsiung 81233, Taiwan, ROC (China); Chen, Yen-Chen [Institute of Nuclear Energy Research (INER), Lungtan Township, Taoyuan County 32546, Taiwan, ROC (China)

    2014-04-01

    In this work, RXA cladding tubes were hydrogen-charged to target hydrogen content levels between 150 and 800 wppm (part per million by weight). The strings of zirconium hydrides observed in the cross sections are mostly oriented in the circumferential direction. The fracture toughness of hydrided RXA Zircaloy-4 cladding was measured to evaluate its hydride embrittlement susceptibility. With increasing hydrogen content, the fracture toughness of hydrided RXA cladding decreases at both 25 °C and 300 °C. Moreover, highly localized hydrides (forming a hydride rim) aggravate the degradation of the fracture properties of RXA Zircaloy-4 cladding at both 25 °C and 300 °C. Brittle features in the form of quasi-cleavages and secondary cracks were observed on the fracture surface of the hydride rim, even for RXA cladding tested at 300 °C.

  16. 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

  17. 5.4W cladding-pumped Nd:YAG silica fiber laser

    OpenAIRE

    Yoo, S.; Webb, A.S.; Standish, R.J.; May-Smith, T.C.; Sahu, J.K.

    2012-01-01

    We report on the spectroscopy and laser characteristics of Nd-doped fiber, fabricated by rod-in-tube from Nd:YAG as a core material with silica cladding. A cladding-pumped CW laser operation at 1058nm with 52% slope-efficiency is demonstrated.

  18. 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 %

  19. Prediction of failure of highly irradiated Zircaloy clad tubes under reactivity initiated accidents

    International Nuclear Information System (INIS)

    Jernkvist, L.O.

    2003-01-01

    This paper deals with failure of irradiated Zircaloy tubes under the heat-up stage of a reactivity initiated accident (RIA). More precisely, by use of a model for plastic strain localization and necking failure, we theoretically analyse the effects of local surface defects on clad ductility and survivability under RIA. The results show that even very shallow surface defects, e.g. arising from a non-uniform or partially spilled oxide layer, have a strong limiting effect on clad ductility. Moreover, in presence of surface defects, the ability of the clad tube to expand radially without necking failure is found to be extremely sensitive to the stress biaxiality ratio σ zz /σ θθ , which is here assumed to be in the range from 0 to 1. The results of our analysis are compared with clad ductility data available in literature, and their consequences for clad failure prediction under RIA are discussed. In particular, the results raise serious concerns regarding the applicability of failure criteria, which are based on clad strain energy density. These criteria do not capture the observed sensitivity to stress biaxiality on clad failure propensity. (author)

  20. 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

  1. Single-mode fiber laser based on core-cladding mode conversion.

    Science.gov (United States)

    Suzuki, Shigeru; Schülzgen, Axel; Peyghambarian, N

    2008-02-15

    A single-mode fiber laser based on an intracavity core-cladding mode conversion is demonstrated. The fiber laser consists of an Er-doped active fiber and two fiber Bragg gratings. One Bragg grating is a core-cladding mode converter, and the other Bragg grating is a narrowband high reflector that selects the lasing wavelength. Coupling a single core mode and a single cladding mode by the grating mode converter, the laser operates as a hybrid single-mode laser. This approach for designing a laser cavity provides a much larger mode area than conventional large-mode-area step-index fibers.

  2. 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.)

  3. Magnetostrictive clad steel plates for high-performance vibration energy harvesting

    Science.gov (United States)

    Yang, Zhenjun; Nakajima, Kenya; Onodera, Ryuichi; Tayama, Tsuyoki; Chiba, Daiki; Narita, Fumio

    2018-02-01

    Energy harvesting technology is becoming increasingly important with the appearance of the Internet of things. In this study, a magnetostrictive clad steel plate for harvesting vibration energy was proposed. It comprises a cold-rolled FeCo alloy and cold-rolled steel joined together by thermal diffusion bonding. The performances of the magnetostrictive FeCo clad steel plate and conventional FeCo plate cantilevers were compared under bending vibration; the results indicated that the clad steel plate construct exhibits high voltage and power output compared to a single-plate construct. Finite element analysis of the cantilevers under bending provided insights into the magnetic features of a clad steel plate, which is crucial for its high performance. For comparison, the experimental results of a commercial piezoelectric bimorph cantilever were also reported. In addition, the cold-rolled FeCo and Ni alloys were joined by thermal diffusion bonding, which exhibited outstanding energy harvesting performance. The larger the plate volume, the more the energy generated. The results of this study indicated not only a promising application for the magnetostrictive FeCo clad steel plate as an efficient energy harvester, related to small vibrations, but also the notable feasibility for the formation of integrated units to support high-power trains, automobiles, and electric vehicles.

  4. Validation and evaluation of common large-area display set (CLADS) performance specification

    Science.gov (United States)

    Hermann, David J.; Gorenflo, Ronald L.

    1998-09-01

    Battelle is under contract with Warner Robins Air Logistics Center to design a Common Large Area Display Set (CLADS) for use in multiple Command, Control, Communications, Computers, and Intelligence (C4I) applications that currently use 19- inch Cathode Ray Tubes (CRTs). Battelle engineers have built and fully tested pre-production prototypes of the CLADS design for AWACS, and are completing pre-production prototype displays for three other platforms simultaneously. With the CLADS design, any display technology that can be packaged to meet the form, fit, and function requirements defined by the Common Large Area Display Head Assembly (CLADHA) performance specification is a candidate for CLADS applications. This technology independent feature reduced the risk of CLADS development, permits life long technology insertion upgrades without unnecessary redesign, and addresses many of the obsolescence problems associated with COTS technology-based acquisition. Performance and environmental testing were performed on the AWACS CLADS and continues on other platforms as a part of the performance specification validation process. A simulator assessment and flight assessment were successfully completed for the AWACS CLADS, and lessons learned from these assessments are being incorporated into the performance specifications. Draft CLADS specifications were released to potential display integrators and manufacturers for review in 1997, and the final version of the performance specifications are scheduled to be released to display integrators and manufacturers in May, 1998. Initial USAF applications include replacements for the E-3 AWACS color monitor assembly, E-8 Joint STARS graphics display unit, and ABCCC airborne color display. Initial U.S. Navy applications include the E-2C ACIS display. For these applications, reliability and maintainability are key objectives. The common design will reduce the cost of operation and maintenance by an estimated 3.3M per year on E-3 AWACS

  5. 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

  6. 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

  7. In-pile cladding tests at NRI Rez and PIE capabilities and experience

    International Nuclear Information System (INIS)

    Zmitko, M.

    2002-01-01

    In-pile cladding corrosion test facilities and relevant post-irradiation capabilities at NRI Rez plc are overviewed. Basic information about the research rector LVR-15 and in-pile water loops is given. An experience in the field of Zr-alloy cladding corrosion testing and investigation of cladding corrosion behaviour is demonstrated for two experimental programmes conducted at NRI Rez in the past period. The first example describes results obtained at studying of corrosion behaviour of advanced Zr-alloys under PWR conditions with a special concern to a high lithium content and subcooled surface boiling. The second example informs about completion of the experimental programme supported by the IAEA which is focused on investigation of Zircaloy-4 cladding behaviour under VVER water chemistry, thermal-hydraulic and irradiation conditions with the main to obtain experimental data for an assessment of the Zircaloy-4 cladding compatibility with VVER conditions. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Alemohammad, Hamidreza [University of Waterloo, Department of Mechanical and Mechatronics Engineering, 200 University Avenue West, Waterloo, ON N2L 3G1 (Canada)], E-mail: shalemoh@engmail.uwaterloo.ca; Esmaeili, Shahrzad [University of Waterloo, Department of Mechanical and Mechatronics Engineering, 200 University Avenue West, Waterloo, ON N2L 3G1 (Canada); Toyserkani, Ehsan [University of Waterloo, Department of Mechanical and Mechatronics Engineering, 200 University Avenue West, Waterloo, ON N2L 3G1 (Canada)

    2007-05-15

    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{sub 3} and {beta} (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{sub 3} phase.

  9. Modeling the geometric formation and powder deposition mass in laser induction hybrid cladding

    International Nuclear Information System (INIS)

    Huang, Yong Jun; Yuan, Sheng Fa

    2012-01-01

    A new laser induction hybrid cladding technique on cylinder work piece is presented. Based on a series of laser induction hybrid experiments by off axial powder feeding, the predicting models of individual clad geometric formation and powder catchment were developed in terms of powder feeding rate, laser special energy and induction energy density using multiple regression analysis. In addition, confirmation tests were performed to make a comparison between the predicting results and measured ones. Via the experiments and analysis, the conclusions can be lead to that the process parameters have crucial influence on the clad geometric formation and powder catchment, and that the predicting model reflects well the relationship between the clad geometric formation and process parameters in laser induction hybrid cladding

  10. 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.

  11. 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.

  12. Ultrasonic spot welding of Al/Mg/Al tri-layered clad sheets

    International Nuclear Information System (INIS)

    Macwan, A.; Patel, V.K.; Jiang, X.Q.; Li, C.; Bhole, S.D.; Chen, D.L.

    2014-01-01

    Highlights: • The optimal welding condition is achieved at 100 J and 0.1 s. • Failure load first increases and then decreases with increasing welding energy. • The highest failure load after welding is close to that of the clad sheets. • At low energy levels failure occurs in the mode of interfacial failure. • At high energy levels failure takes place at the edge of nugget region. - Abstract: Solid-state ultrasonic spot welding (USW) was used to join Al/Mg/Al tri-layered clad sheets, aiming at exploring weldability and identifying failure mode in relation to the welding energy. It was observed that the application of a low welding energy of 100 J was able to achieve the optimal welding condition during USW at a very short welding time of 0.1 s for the tri-layered clad sheets. The optimal lap shear failure load obtained was equivalent to that of the as-received Al/Mg/Al tri-layered clad sheets. With increasing welding energy, the lap shear failure load initially increased and then decreased after reaching a maximum value. At a welding energy of 25 J, failure occurred in the mode of interfacial failure along the center Al/Al weld interface due to insufficient bonding. At a welding energy of 50 J, 75 J and 100 J, failure was also characterized by the interfacial failure mode, but it occurred along the Al/Mg clad interface rather than the center Al/Al weld interface, suggesting stronger bonding of the Al/Al weld interface than that of the Al/Mg clad interface. The overall weld strength of the Al/Mg/Al tri-layered clad sheets was thus governed by the Al/Mg clad interface strength. At a welding energy of 125 J and 150 J, thinning of weld nugget and extensive deformation at the edge of welding tip caused failure at the edge of nugget region, leading to a lower lap shear failure load

  13. Advanced LWR Nuclear Fuel Cladding Development

    International Nuclear Information System (INIS)

    Bragg-Sitton, S.; Griffith, G.

    2012-01-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

  14. Cladding failure by local plastic instability

    International Nuclear Information System (INIS)

    Kramer, J.M.; Deitrich, L.W.

    1977-01-01

    Cladding failure is one of the major considerations in analysis of fast-reactor fuel pin behavior during hypothetical accident transients since time, location and nature of failure govern the early post-failure material motion and reactivity feedback. Out-of-Pile transient burst tests of both irradiated and unirradiated fast-reactor cladding show that local plastic instability, or bulging, often precedes rupture. To investigate the details of cladding bulging, a perturbation analysis of the equations governing the large deformation of a cylindrical shell has been developed. The overall deformation history is assumed to consist of a small perturbation epsilon of the radial displacement superimposed on large axisymmetric displacements. Computations have been carried out using high temperature properties of stainless steel in conjunction with various constitutive theories, including a generalization of the Endochronic Theory of Plasticity in which both time-independent and time-dependent plastic strains are modeled. Although the results of the calculations are all qualitatively similar, it appears that modeling of both time-independent and time-dependent plastic strains is necessary to interpret the transient burst test results. Sources for bulge formation that have been considered include initial geometric imperfections and thermal perturbations due to either eccentric fuel pellets or non-symmetric cooling. (Auth.)

  15. 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...

  16. Modelling of ultrasonic nondestructive testing of cracks in claddings

    International Nuclear Information System (INIS)

    Bostroem, Anders; Zagbai, Theo

    2006-05-01

    Nondestructive testing with ultrasound is a standard procedure in the nuclear power industry. To develop and qualify the methods extensive experimental work with test blocks is usually required. This can be very time-consuming and costly and it also requires a good physical intuition of the situation. A reliable mathematical model of the testing situation can, therefore, be very valuable and cost-effective as it can reduce experimental work significantly. A good mathematical model enhances the physical intuition and is very useful for parametric studies, as a pedagogical tool, and for the qualification of procedures and personnel. The present project has been concerned with the modelling of defects in claddings. A cladding is a layer of material that is put on for corrosion protection, in the nuclear power industry this layer is often an austenitic steel that is welded onto the surface. The cladding is usually anisotropic and to some degree it is most likely also inhomogeneous, particularly in that the direction of the anisotropy is varying. This degree of inhomogeneity is unknown but probably not very pronounced so for modelling purposes it may be a valid assumption to take the cladding to be homogeneous. However, another important complicating factor with claddings is that the interface between the cladding and the base material is often corrugated. This corrugation can have large effects on the transmission of ultrasound through the interface and can thus greatly affect the detectability of defects in the cladding. In the present project the only type of defect that is considered is a planar crack that is situated inside the cladding. The investigations are, furthermore, limited to two dimensions, and the crack is then only a straight line. The crack can be arbitrarily oriented and situated, but it must not intersect the interface to the base material. The crack can be surface-breaking, and this is often the case of most practical interest, but it should then be

  17. A New Material Constitutive Model for Predicting Cladding Failure

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Joe; Dunham, Robert [ANATECH Corp., San Diego, CA (United States); Rashid, Mark [University of California Davis, Davis, CA (United States); Machiels, Albert [EPRI, Palo Alto, CA (United States)

    2009-06-15

    An important issue in fuel performance and safety evaluations is the characterization of the effects of hydrides on cladding mechanical response and failure behavior. The hydride structure formed during power operation transforms the cladding into a complex multi-material composite, with through-thickness concentration profile that causes cladding ductility to vary by more than an order of magnitude between ID and OD. However, current practice of mechanical property testing treats the cladding as a homogeneous material characterized by a single stress-strain curve, regardless of its hydride morphology. Consequently, as irradiation conditions and hydrides evolution change, new material property testing is required, which results in a state of continuous need for valid material property data. A recently developed constitutive model, treats the cladding as a multi-material composite in which the metal and the hydride platelets are treated as separate material phases with their own elastic-plastic and fracture properties and interacting at their interfaces with appropriate constraint conditions between them to ensure strain and stress compatibility. An essential feature of the model is a multi-phase damage formulation that models the complex interaction between the hydride phases and the metal matrix and the coupled effect of radial and circumferential hydrides on cladding stress-strain response. This gives the model the capability of directly predicting cladding failure progression during the loading event and, as such, provides a unique tool for constructing failure criteria analytically where none could be developed by conventional material testing. Implementation of the model in a fuel behavior code provides the capability to predict in-reactor operational failures due to PCI or missing pellet surfaces (MPS) without having to rely on failure criteria. Even, a stronger motivation for use of the model is in the transportation accidents analysis of spent fuel

  18. Oxidation behavior of laser-clad NiAlCrHf alloys

    International Nuclear Information System (INIS)

    Ribaudo, C.R.

    1991-01-01

    Laser cladding is the process where a mechanical mixture of powders is rapidly melted and fused to a solid substrate using a CO 2 laser. The effects of laser cladding upon scale retention on NiAlCrHf alloys after cyclic and isothermal exposure to air were investigated. The stress developed in the scale during cooling after exposure was estimated using a thermoelastic model. Additions of up to ∼2 1/2 wt % Hf increasingly promote retention of scales grown at 1,200C. Laser-clad samples containing ∼2 1/2 wt % Hf retained almost-intact scales. The improvement in scale retention is due to improved toughness in scales containing hafnia-rich polycrystallites possibly via microcracking initiated by anisotropic thermal contraction of the hafnia. Laser cladding provides a large concentration of ∼1 μm Hf-rich particles that are precursors of the hafnia in the scale as well as a fine-dendrite spacing that reduced the mean free distance between particles

  19. 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.)

  20. Pellet Cladding Mechanical Interaction Modeling Using the Extended Finite Element Method

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin W.; Jiang, Wen; Dolbow, John E.; Peco, Christian

    2016-09-01

    As a brittle material, the ceramic UO2 used as light water reactor fuel experiences significant fracturing throughout its life, beginning with the first rise to power of fresh fuel. This has multiple effects on the thermal and mechanical response of the fuel/cladding system. One such effect that is particularly important is that when there is mechanical contact between the fuel and cladding, cracks that extending from the outer surface of the fuel into the volume of the fuel cause elevated stresses in the adjacent cladding, which can potentially lead to cladding failure. Modeling the thermal and mechanical response of the cladding in the vicinity of these surface-breaking cracks in the fuel can provide important insights into this behavior to help avoid operating conditions that could lead to cladding failure. Such modeling has traditionally been done in the context of finite-element-based fuel performance analysis by modifying the fuel mesh to introduce discrete cracks. While this approach is effective in capturing the important behavior at the fuel/cladding interface, there are multiple drawbacks to explicitly incorporating the cracks in the finite element mesh. Because the cracks are incorporated in the original mesh, the mesh must be modified for cracks of specified location and depth, so it is difficult to account for crack propagation and the formation of new cracks at other locations. The extended finite element method (XFEM) has emerged in recent years as a powerful method to represent arbitrary, evolving, discrete discontinuities within the context of the finite element method. Development work is underway by the authors to implement XFEM in the BISON fuel performance code, and this capability has previously been demonstrated in simulations of fracture propagation in ceramic nuclear fuel. These preliminary demonstrations have included only the fuel, and excluded the cladding for simplicity. This paper presents initial results of efforts to apply XFEM to

  1. Influence of pellet-clad-gap-size on LWR fuel rod performance

    International Nuclear Information System (INIS)

    Brzoska, B.; Fuchs, H.P.; Garzarolli, F.; Manzel, R.

    1979-01-01

    The as-fabricated pellet-clad-gap size varies due to fabricational tolerances of the cladding inner diameter and the pellet outer diameter. The consequences of these variations on the fuel rod behaviour are analyzed using the KWU fuel rod code CARO. The code predictions are compared with experimental results of special pathfinder test fuel rods irradiated in the OBRIGHEIM nuclear power plant. These test fuel rods include gap sizer in the range of 140 μm to 270 μm, prepressurization between 13 bar to 36 bar and Helium and Argon fill gases irradiated up to a local burnup of 35 MWd/kg(U). Post irradiation examination were performed at different burnups. CARC calculations have been performed with special emphasis in cladding creep down, fission gas release and pellet clad gap closure. (orig.)

  2. 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)

  3. Cladding tube manufacturing technology

    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

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

  4. Cladding tube manufacturing technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

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

    Directory of Open Access Journals (Sweden)

    Ki-Nam Jang

    2017-12-01

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

  6. Microstructure of laser cladded martensitic stainless steel

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2006-08-01

    Full Text Available and martensite with 10% ferrite for Material B. Table 7 - Proposed martensitic stainless steel alloys for laser cladding Material C* Cr Ni Mn Si Mo Co Ms (ºC)* Cr eq Ni eq Material A 0.4 13 - 1 0.5 2.5 5.5 120 16.5 12.5 Material B 0.2 15 2 1 0.7 2.5 5.5 117... dilution, low heat input, less distortion, increased mechanical and corrosion properties excellent repeatability and control of process parameters. Solidification of laser cladded martensitic stainless steel is primarily austenitic. Microstructures...

  7. Laser and Pressure Resistance Weld of Thin-Wall Cladding for LWR Accident-Tolerant Fuels

    Science.gov (United States)

    Gan, J.; Jerred, N.; Perez, E.; Haggard, D. C.

    2018-02-01

    FeCrAl alloy with typical composition of approximately Fe-15Cr-5Al is considered a primary candidate cladding material for light water reactor accident-tolerant fuel because of its superior resistance to oxidation in high-temperature steam compared with Zircaloy cladding. Thin-walled FeCrAl cladding at 350 μm wall thickness is required, and techniques for joining endplug to cladding need to be developed. Fusion-based laser weld and solid-state joining with pressure resistance weld were investigated in this study. The results of microstructural characterization, mechanical property evaluation by tensile testing, and hydraulic pressure burst testing of the welds for the cladding-endplug specimen are discussed.

  8. 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

  9. Technology readiness level (TRL) assessment of cladding alloys for advanced nuclear fuels

    International Nuclear Information System (INIS)

    Shepherd, Daniel

    2015-01-01

    Reliable fuel claddings are essential for the safe, sustainable and economic operation of nuclear stations. This paper presents a worldwide TRL assessment of advanced claddings for Gen III and IV reactors following an extensive literature review. Claddings include austenitic, ferritic/martensitic (F/M), reduced activation (RA) and oxide dispersion strengthened (ODS) steels as well as advanced iron-based alloys (Kanthal alloys). Also assessed are alloys of zirconium, nickel (including Hastelloy R ), titanium, chromium, vanadium and refractory metals (Nb, Mo, Ta and W). Comparison is made with Cf/C and SiCf/SiC composites, MAX phase ceramics, cermets and TRISO fuel particle coatings. The results show in general that the higher the maximum operating temperature of the cladding, the lower the TRL. Advanced claddings were found to have lower TRLs than the corresponding fuel materials, and therefore may be the limiting factor in the deployment of advanced fuels and even possibly the entire reactor in the case of Gen IV. (authors)

  10. 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

  11. Behavior of high burnup fuel rod cladding during long-term dry storage in CASTOR casks

    International Nuclear Information System (INIS)

    Schaberg, A.; Spilker, H.; Goll, W.

    2000-01-01

    Short-time creep and rupture tests were performed to assess the strain potential of cladding of high burnt rods under conditions of dry storage. The tests comprised optimized Zr y-4 cladding samples from fuel rods irradiated to burnups of up to 64 MWd/kg U and were carried out at temperatures of 573 and 643 K at cladding stresses of about 400 and 600 MPa. The stresses, much higher than those occurring in a fuel rod, were chosen to reach circumferential elongations of about 2% within an envisaged testing time of 3-4 days. The creep tests were followed by a low temperature test at 423 K and 100 MPa to assess the long-term behavior of the cladding ductility especially with regard to the effect of a higher hydrogen content in the cladding due to the high burnup. The creep tests showed considerable uniform plastic elongations at these high burnups. It was demonstrated that around 600 K a uniform plastic strain of a least 2% is reached without cladding failure. The low temperature tests at 423 K for up to 5 days revealed no cladding failure under these conditions of reduced cladding ductility. It can be concluded that the increased hydrogen content has no adverse effect on cladding performance. (Authors)

  12. Laser cladding of a Mg based Mg–Gd–Y–Zr alloy with Al–Si powders

    International Nuclear Information System (INIS)

    Chen, Erlei; Zhang, Kemin; Zou, Jianxin

    2016-01-01

    Graphical abstract: A Mg based Mg–Gd–Y–Zr alloy was treated by laser cladding with Al–Si powders at different laser scanning speeds. The laser clad layer mainly contains Mg_2Si, Mg_1_7Al_1_2 and Al_2(Gd,Y) phases distributed in the Mg matrix. After laser cladding, the corrosion resistance of the Mg alloy was significantly improved together with increased microhardness in the laser clad layers. - Highlights: • A Mg based Mg–Gd–Y–Zr alloy was laser clad with Al–Si powders. • The microstructure and morphology vary with the depth of the clad layer and the laser scanning speed. • Hardness and corrosion resistance were significantly improved after laser cladding. - Abstract: In the present work, a Mg based Mg–Gd–Y–Zr alloy was subjected to laser cladding with Al–Si powders at different laser scanning speeds in order to improve its surface properties. It is observed that the laser clad layer mainly contains Mg_2Si, Mg_1_7Al_1_2 and Al_2(Gd,Y) phases distributed in the Mg matrix. The depth of the laser clad layer increases with decreasing the scanning speed. The clad layer has graded microstructures and compositions. Both the volume fraction and size of Mg_2Si, Mg_1_7Al_1_2 and Al_2(Gd,Y) phases decreases with the increasing depth. Due to the formation of these hardening phases, the hardness of clad layer reached a maximum value of HV440 when the laser scanning speed is 2 mm/s, more than 5 times of the substrate (HV75). Besides, the corrosion properties of the untreated and laser treated samples were all measured in a NaCl (3.5 wt.%) aqueous solution. The corrosion potential was increased from −1.77 V for the untreated alloy to −1.13 V for the laser clad alloy with scanning rate of 2 mm/s, while the corrosion current density was reduced from 2.10 × 10"−"5 A cm"−"2 to 1.64 × 10"−"6 A cm"−"2. The results show that laser cladding is an efficient method to improve surface properties of Mg–Rare earth alloys.

  13. Surface improvement for inside surface of small diameter pipes by laser cladding technique

    International Nuclear Information System (INIS)

    Irisawa, Toshio; Morishige, Norio; Umemoto, Tadahiro; Ono, Kazumichi; Hamaoka, Tadashi; Tanaka, Atsushi

    1991-01-01

    A laser cladding technique has been used for surface improvement in controlling the composition of a metal surface. Recent high power YAG laser development gives an opportunity to use this laser cladding technique for various applications. A YAG laser beam can be transmitted through an optical fiber for a long distance and through narrow spaces. YAG laser cladding was studied for developing alloy steel to prevent stress corrosion cracking in austenitic stainless steel piping. In order to make a cladding layer, mixed metal powder was on the inside surface of the piping using an organic binder. Subsequently the powder beds were melted with a YAG laser beam transmitted through an optical fiber. This paper introduces the Laser cladding technique for surface improvement for the inside surface of a small diameter pipe. (author)

  14. Microscopic Analysis and Electrochemical Behavior of Fe-Based Coating Produced by Laser Cladding

    Directory of Open Access Journals (Sweden)

    Jinlin Chen

    2017-10-01

    Full Text Available The effect of laser cladding on the surface microstructure and corrosion properties of coated/uncoated specimens were investigated. Fe-based alloy coating was produced on 35CrMo steel by laser cladding. The phase composition, microstructure, interface element distribution, microhardness and corrosion resistance of the cladding coating were measured. The results show that the cladding layer is mainly composed of α-Fe phases, the microstructure presents a gradient distribution, and a good metallurgical bond is formed at the boundary with the substrate. Microhardness profiles show that the average microhardness of the cladding coating is about 2.1 times higher than that of the uncoated specimen. In addition, the electrochemical results show that the coated specimen exhibits far better corrosion resistance than to the uncoated specimen.

  15. In-reactor measurement of clad strain: effect of power history

    International Nuclear Information System (INIS)

    Fehrenbach, P.J.; Morel, P.A.

    1980-01-01

    A series of experimental irradiations has been undertaken at CRNL to measure directly the in-reactor deformation of fuel elements while they are operating at power. Power histories have been chosen to allow investigation of power, time at power and burnup on pellet-clad interaction for element linear powers to 60kW/m. Results are presented which indicate that irradiation of a fresh fuel element at high power is effective in minimizing clad hoop stresses during subsequent ramps or cycles to that power. The effectiveness of this preconditioning appears to be due primarily to fuel densification rather than stress relaxation in the clad. (auth)

  16. In-reactor cladding breach of EBR-II driver-fuel elements

    International Nuclear Information System (INIS)

    Seidel, B.R.; Einziger, R.E.

    1977-01-01

    Knowledge of performance and minimum useful element lifetime of Mark-II driver-fuel elements is required to maintain a high plant operating capacity factor with maximum fuel utilization. To obtain such knowledge, intentional cladding breach has been obtained in four run-to-cladding-breach Mark-II experimental driver-fuel subassemblies operating under normal conditions in EBR-II. Breach and subsequent fission-product release proved benign to reactor operations. The breaches originated on the outer surface of the cladding in the root of the restrainer dimples and were intergranular. The Weibull distribution of lifetime accurately predicts the observed minimum useful element lifetime of 10 at.% burnup, with breach ensuing shortly thereafter

  17. 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

  18. Progress and Challenges of Ultrasonic Testing for Stress in Remanufacturing Laser Cladding Coating

    Directory of Open Access Journals (Sweden)

    Xiao-Ling Yan

    2018-02-01

    Full Text Available 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.

  19. 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

  20. Evaluation of the potential of indigenous calcareous shale for neutralization and removal of arsenic and heavy metals from acid mine drainage in the Taxco mining area, Mexico.

    Science.gov (United States)

    Romero, F M; Núñez, L; Gutiérrez, M E; Armienta, M A; Ceniceros-Gómez, A E

    2011-02-01

    In the Taxco mining area, sulfide mineral oxidation from inactive tailings impoundments and abandoned underground mines has produced acid mine drainage (AMD; pH 2.2-2.9) enriched in dissolved concentrations (mg l⁻¹) sulfate, heavy metals, and arsenic (As): SO₄²⁻ (pH 1470-5454), zinc (Zn; 3.0-859), iron (Fe; pH 5.5-504), copper (Cu; pH 0.7-16.3), cadmium (Cd; pH 0.3-6.7), lead (Pb; pH acid-neutralizing potential of limestone decreases when surfaces of the calcite particles become less reactive as they are progressively coated by metal precipitates. This study constitutes first-stage development of passive-treatment systems for treating AMD in the Taxco mine area using indigenous calcareous shale. This geologic material consists of a mixture of calcite, quartz, muscovite, albite, and montmorillonite. Results of batch leaching test indicate that calcareous shale significantly increased the pH (to values of 6.6-7.4) and decreased heavy metal and As concentrations in treated mine leachates. Calcareous shale had maximum removal efficiency (100%) for As, Pb, Cu, and Fe. The most mobile metals ions were Cd and Zn, and their average percentage removal was 87% and 89%, respectively. In this natural system (calcareous shale), calcite provides a source of alkalinity, whereas the surfaces of quartz and aluminosilicate minerals possibly serve as a preferred locus of deposition for metals, resulting in the neutralizing agent (calcite) beings less rapidly coated with the precipitating metals and therefore able to continue its neutralizing function for a longer time.

  1. Microstructural and wear characteristics of cobalt free, nickel base intermetallic alloy deposited by laser cladding

    International Nuclear Information System (INIS)

    Awasthi, Reena; Kumar, Santosh; Viswanadham, C.S.; Srivastava, D.; Dey, G.K.; Limaye, P.K.

    2011-01-01

    This paper describes the microstructural and wear characteristics of Ni base intermetallic hardfacing alloy (Tribaloy-700) deposited on stainless steel-316 L substrate by laser cladding technique. Cobalt base hardfacing alloys have been most commonly used hardfacing alloys for application involving wear, corrosion and high temperature resistance. However, the high cost and scarcity of cobalt led to the development of cobalt free hardfacing alloys. Further, in the nuclear industry, the use of cobalt base alloys is limited due to the induced activity of long lived radioisotope 60 Co formed. These difficulties led to the development of various nickel and iron base alloys to replace cobalt base hardfacing alloys. In the present study Ni base intermetallic alloy, free of Cobalt was deposited on stainless steel- 316 L substrate by laser cladding technique. Traditionally, welding and thermal spraying are the most commonly employed hardfacing techniques. Laser cladding has been explored for the deposition of less diluted and fusion-bonded Nickel base clad layer on stainless steel substrate with a low heat input. The laser cladding parameters (Laser power density: 200 W/mm 2 , scanning speed: 430 mm/min, and powder feed rate: 14 gm/min) resulted in defect free clad with minimal dilution of the substrate. The microstructure of the clad layer was examined by Optical microscopy, Scanning electron microscopy, with energy dispersive spectroscopy. The phase analysis was performed by X-ray diffraction technique. The clad layer exhibited sharp substrate/clad interface in the order of planar, cellular, and dendritic from the interface upwards. Dilution of clad with Fe from substrate was very low passing from ∼ 15% at the interface (∼ 40 μm) to ∼ 6% in the clad layer. The clad layer was characterized by the presence of hexagonal closed packed (hcp, MgZn 2 type) intermetallic Laves phase dispersed in the eutectic of Laves and face centered cubic (fcc) gamma solid solution. The

  2. Development of advanced claddings for suppressing the hydrogen emission in accident conditions. Development of advanced claddings for suppressing the hydrogen emission in the accident condition

    International Nuclear Information System (INIS)

    Park, Jeong-Yong; KIM, Hyun-Gil; JUNG, Yang-Il; PARK, Dong-Jun; KOO, Yang-Hyun

    2013-01-01

    The development of accident-tolerant fuels can be a breakthrough to help solve the challenge facing nuclear fuels. One of the goals to be reached with accident-tolerant fuels is to reduce the hydrogen emission in the accident condition by improving the high-temperature oxidation resistance of claddings. KAERI launched a new project to develop the accident-tolerant fuel claddings with the primary objective to suppress the hydrogen emission even in severe accident conditions. Two concepts are now being considered as hydrogen-suppressed cladding. In concept 1, the surface modification technique was used to improve the oxidation resistance of Zr claddings. Like in concept 2, the metal-ceramic hybrid cladding which has a ceramic composite layer between the Zr inner layer and the outer surface coating is being developed. The high-temperature steam oxidation behaviour was investigated for several candidate materials for the surface modification of Zr claddings. From the oxidation tests carried out in 1 200 deg. C steam, it was found that the high-temperature steam oxidation resistance of Cr and Si was much higher than that of zircaloy-4. Al 3 Ti-based alloys also showed extremely low-oxidation rate compared to zircaloy-4. One important part in the surface modification is to develop the surface coating technology where the optimum process needs to be established depending on the surface layer materials. Several candidate materials were coated on the Zr alloy specimens by a laser beam scanning (LBS), a plasma spray (PS) and a PS followed by LBS and subject to the high-temperature steam oxidation test. It was found that Cr and Si coating layers were effective in protecting Zr-alloys from the oxidation. The corrosion behaviour of the candidate materials in normal reactor operation condition such as 360 deg. C water will be investigated after the screening test in the high-temperature steam. The metal-ceramic hybrid cladding consisted of three major parts; a Zr liner, a

  3. 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.

  4. Spent fuel cladding containment credit test

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1983-01-01

    As an initial step in addressing the effectiveness of breached cladding as a barrier to radionuclide release from the repository during the post-containment period, preliminary scoping tests have been initiated which compare radionuclide releases from spent fuel specimens with artificially induced cladding defects of various severities. The artificially induced defects are all more severe than the typical in-reactor type breaches which are expected to be the principal type of breach entering the repository for terminal storage. These preliminary scoping tests being conducted by Westinghouse Hanford Company for the Lawrence Livermore National Laboratory Waste Package Development Program in support of the Tuff repository project at the Nevada Test Site are described. Also included in this presentation are selected initial results from these tests. 22 figures

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-28

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

  6. 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

  7. 5  W output power from a double-clad hybrid fiber with Yb-doped phosphate core and silicate cladding.

    Science.gov (United States)

    Wang, Longfei; He, Dongbing; Zhang, Lei; Yu, Chunlei; Feng, Suya; Wang, Meng; Chen, Danping; Hu, Lili

    2017-08-01

    For the first time, to the best of our knowledge, we report on the realization of a laser from a Yb-doped phosphate core/silicate cladding double-clad hybrid fiber. 5 W output power was extracted with 14.6% slope efficiency and a laser spectrum of a 1027 nm central wavelength from a 20 cm long single-mode fiber with a ∼10  μm core diameter in a 20%-4% laser cavity. The laser efficiency can be significantly enhanced by correspondingly adjusting and optimizing the laser oscillator.

  8. 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.

  9. Analysis of clad motion observed in loss of flow accident simulation experiments

    International Nuclear Information System (INIS)

    Henkel, P.R.

    1987-01-01

    The clad motion observed in the first two STAR experiments is analysed. The movies reveal that at moderate temperatures molten cladding does not wet fresh fuel (within an argon gas atmosphere). The prevailing flow regime consists of single waves contacting the fuel pins and entrained drops. Entrainment is possible already at gas velocities of order 40-50 m/s. A multichannel clad motion model is presented that accounts for both flow modes. (author)

  10. Order of magnitude cost appraisal for selected aspects of clad waste management

    International Nuclear Information System (INIS)

    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

  11. Characteristics of Ni-Cr-Fe laser clad layers on EA4T steel

    Science.gov (United States)

    Chen, Wenjing; Chen, Hui; Wang, Yongjing; Li, Congchen; Wang, Xiaoli

    2017-07-01

    The Ni-Cr-Fe metal powder was deposited on EA4T steel by laser cladding technology. The microstructure and chemical composition of the cladding layer were analyzed by optical microscopy (OM), scanning electron microscopy (SEM) and X-ray diffraction (XRD). The bonding ability between the cladding layer and the matrix was measured. The results showed that the bonding between the cladding layer and the EA4T steel was metallurgical bonding. The microstructure of cladding layer was composed of planar crystals, columnar crystals and dendrite, which consisted of Cr2Ni3, γ phase, M23C6 and Ni3B phases. When the powder feeding speed reached 4 g/min, the upper bainite occurred in the heat affected zone (HAZ). Moreover, the tensile strength of the joint increased, while the yield strength and the ductility decreased.

  12. 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......UV-written Bragg gratings find wide spread use as wavelength selective components. In reflection high extinction ratios are routinely obtained. However, coupling to cladding modes gives excess loss on the short wavelength side of the main reflection. Different fiber-designs have been proposed......-solver based on the staircase-approximation method. The Bragg grating causes coupling between the fundamental LP01 mode and higher order LP1p modes that satisfy phase-matching. The coupling strength is determined by the overlap integral of the LP01, the LP1p mode, and the UV-induced index perturbation. For LP0...

  13. An internal conical mandrel technique for fracture toughness measurements on nuclear fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sainte Catherine, C.; Le Boulch, D.; Carassou, S. [CEA Saclay, DEN/DMN, Bldg 625 P, Gif-Sur-Yvette, F-91191 (France); Lemaignan, C. [CEA Grenoble, 17 rue des Martyrs, Grenoble, F-38054 (France); Ramasubramanian, N. [ECCATEC Inc., 92 Deburn Drive, Toronto, Ontario (Canada)

    2006-07-01

    An understanding of the limiting stress level for crack initiation and propagation in a fuel cladding material is a fundamental requirement for the development of water reactor clad materials. Conventional tests, in use to evaluate fracture properties, are of limited help, because they are adapted from ASTM standards designed for thick materials, which differ significantly from fuel cladding geometry (small diameter thin-walled tubing). The Internal Conical Mandrel (ICM) test described here is designed to simulate the effect of fuel pellet diametrical increase on a cladding with an existing axial through-wall crack. It consists in forcing a cone, having a tapered increase in diameter, inside the Zircaloy cladding with an initial axial crack. The aim of this work is to quantify the crack initiation and propagation criteria for fuel cladding material. The crack propagation is monitored by a video system for obtaining crack extension {delta}a. A finite-element (FE) simulation of the ICM test is performed in order to derive J integrals. A node release technique is applied during the FE simulation for crack propagation and the J-resistance curves (J-{delta}a) are generated. This paper presents the test methodology, the J computation validation, and results for cold-worked stress relieved Zircaloy-4 cladding at 20 deg. and 300 deg. C and also for Al 7050-T7651 aluminum alloy tubing at 20 deg. C. (authors)

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  15. Multifrequency Eddy Current Inspection of Corrosion in Clad Aluminum Riveted Lap Joints and Its Effect on Fatigue Life

    Science.gov (United States)

    Okafor, A. C.; Natarajan, S.

    2007-03-01

    Aging aircraft are prone to corrosion damage and fatigue cracks in riveted lap joints of fuselage skin panels. This can cause catastrophic failure if not detected and repaired. Hence detection of corrosion damage and monitoring its effect on structural integrity are essential. This paper presents multifrequency eddy current (EC) inspection of corrosion damage and machined material loss defect in clad A1 2024-T3 riveted lap joints and its effect on fatigue life. Results of eddy current inspection, corrosion product removal and fatigue testing are presented.

  16. Multifrequency Eddy Current Inspection of Corrosion in Clad Aluminum Riveted Lap Joints and Its Effect on Fatigue Life

    International Nuclear Information System (INIS)

    Okafor, A. C.; Natarajan, S.

    2007-01-01

    Aging aircraft are prone to corrosion damage and fatigue cracks in riveted lap joints of fuselage skin panels. This can cause catastrophic failure if not detected and repaired. Hence detection of corrosion damage and monitoring its effect on structural integrity are essential. This paper presents multifrequency eddy current (EC) inspection of corrosion damage and machined material loss defect in clad A1 2024-T3 riveted lap joints and its effect on fatigue life. Results of eddy current inspection, corrosion product removal and fatigue testing are presented

  17. 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

  18. 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

  19. Facility for in-reactor creep testing of fuel cladding

    International Nuclear Information System (INIS)

    Kohn, E.; Wright, M.G.

    1976-11-01

    A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

  20. Pump radiation distribution in multi-element first cladding laser fibres

    International Nuclear Information System (INIS)

    Mel'kumov, Mikhail A; Bufetov, Igor' A; Bubnov, M M; Shubin, Aleksei V; Semenov, S L; Dianov, Evgenii M

    2005-01-01

    Pump radiation transfer is studied experimentally in multi-element first cladding laser fibres. A model of this process is proposed, which is in good agreement with experimental results. An all-fibre single-mode cw ytterbium laser based on a three-element first cladding fibre with an output power of 100W is fabricated. (lasers)

  1. A study of friction and axial effects in pellet-clad mechanical interaction

    International Nuclear Information System (INIS)

    Harriague, Santiago; Mayer, J.E.

    1982-01-01

    An analysis is made of the effect of friction and axial forces along the fuel rod in the pellet-cladding mechanical interaction in a commercial reactor under a power-up ramp. The effect of different pellet and rod shapes on their behaviour was also determined. A linear thermoelastic computer program was used in order to obtain the stiffness matrix of a compound structure from the stiffness of its components. Pellet-cladding displacements, localized deformations of the cladding in the interfaces between pellets, as well as pellet and cladding axial deformations were determined for different power axial profiles as well as for pellets with and without dishing and with height/diameter ratios of 1.7, 1 and 0.5. (M.E.L.) [es

  2. 3D Finite Element Simulation of Pellet-Cladding Mechanical Interaction

    International Nuclear Information System (INIS)

    Seo, Sang Kyu; Lee, Sung Uk; Lee, Eun Ho; Yang, Dong Yol; Kim, Hyo Chan; Yang, Dong Yol

    2016-01-01

    In a nuclear power plant, the fuel assembly, which is composed of fuel rods, burns, and the high temperature can generate power. The fuel rod consists of pellets and a cladding that covers the pellets. It is important to understand the pellet-cladding mechanical interaction with regard to nuclear safety. This paper proposes simulation of the PCMI. The gap between the pellets and the cladding, and the contact pressure are very important for conducting thermal analysis. Since the gap conductance is not known, it has to be determined by a suitable method. This paper suggests a solution. In this study, finite element (FE) contact analysis is conducted considering thermal expansion of the pellets. As the contact causes plastic deformation, this aspect is considered in the analysis. A 3D FE module is developed to analyze the PCMI using FORTRAN 90. The plastic deformation due to the contact between the pellets and the cladding is the major physical phenomenon. The simple analytical solution of a cylinder is proposed and compared with the fuel rod performance code results

  3. Laser cladding of a Mg based Mg–Gd–Y–Zr alloy with Al–Si powders

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Erlei [School of Materials Engineering, Shanghai University of Engineering Science, Shanghai 201620 (China); Zhang, Kemin, E-mail: zhangkm@sues.edu.cn [School of Materials Engineering, Shanghai University of Engineering Science, Shanghai 201620 (China); Zou, Jianxin [National Engineering Research Center of Light Alloys Net Forming & School of Materials Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China)

    2016-03-30

    Graphical abstract: A Mg based Mg–Gd–Y–Zr alloy was treated by laser cladding with Al–Si powders at different laser scanning speeds. The laser clad layer mainly contains Mg{sub 2}Si, Mg{sub 17}Al{sub 12} and Al{sub 2}(Gd,Y) phases distributed in the Mg matrix. After laser cladding, the corrosion resistance of the Mg alloy was significantly improved together with increased microhardness in the laser clad layers. - Highlights: • A Mg based Mg–Gd–Y–Zr alloy was laser clad with Al–Si powders. • The microstructure and morphology vary with the depth of the clad layer and the laser scanning speed. • Hardness and corrosion resistance were significantly improved after laser cladding. - Abstract: In the present work, a Mg based Mg–Gd–Y–Zr alloy was subjected to laser cladding with Al–Si powders at different laser scanning speeds in order to improve its surface properties. It is observed that the laser clad layer mainly contains Mg{sub 2}Si, Mg{sub 17}Al{sub 12} and Al{sub 2}(Gd,Y) phases distributed in the Mg matrix. The depth of the laser clad layer increases with decreasing the scanning speed. The clad layer has graded microstructures and compositions. Both the volume fraction and size of Mg{sub 2}Si, Mg{sub 17}Al{sub 12} and Al{sub 2}(Gd,Y) phases decreases with the increasing depth. Due to the formation of these hardening phases, the hardness of clad layer reached a maximum value of HV440 when the laser scanning speed is 2 mm/s, more than 5 times of the substrate (HV75). Besides, the corrosion properties of the untreated and laser treated samples were all measured in a NaCl (3.5 wt.%) aqueous solution. The corrosion potential was increased from −1.77 V for the untreated alloy to −1.13 V for the laser clad alloy with scanning rate of 2 mm/s, while the corrosion current density was reduced from 2.10 × 10{sup −5} A cm{sup −2} to 1.64 × 10{sup −6} A cm{sup −2}. The results show that laser cladding is an efficient method to improve

  4. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    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.

  6. 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.)

  7. Report of the advanced neutron source (ANS) aluminum cladding corrosion workshop

    International Nuclear Information System (INIS)

    Hanson, G.H.; Gibson, G.W.; Griess, J.C.; Pawel, R.E.; Pace, N.E.; Ryskamp, J.M.

    1989-02-01

    The Advanced Neutron Source (ANS) Corrosion Workshop on aluminum cladding corrosion in reactor environments is summarized. The Workshop was held to examine the aluminum cladding oxidation studies being conducted in support of the ANS design. This report was written principally to provide a record of the ideas and judgments expressed by the workshop attendees. The ANS operating heat flux is significantly higher than that in existing reactors, and early experiments indicate that there may be an aluminum cladding oxidation problem unique to higher heat fluxes or associated cladding temperatures that, if not solved, may limit the operation of the ANS to unacceptably low power levels. A brief description of the information presented by each speaker is included along with a compilation of the most significant ideas and recommended research areas. The appendixes contain a copy of the workshop agenda and a list of attendees

  8. Adsorption and magnetic removal of neutral red dye from aqueous solution using Fe3O4 hollow nanospheres

    International Nuclear Information System (INIS)

    Iram, Mahmood; Guo, Chen; Guan Yueping; Ishfaq, Ahmad; Liu Huizhou

    2010-01-01

    Fe 3 O 4 hollow nanospheres were prepared via a simple one-pot template-free hydrothermal method and were fully characterized. These magnetic spheres have been investigated for application as an adsorbant for the removal of dye contaminants from water. Because of the high specific surface area, nano-scale particle size, and hollow porous material, Fe 3 O 4 hollow spheres showed favorable adsorption behavior for Neutral red. Factors affecting adsorption, such as, initial dye concentration, pH and contact time were evaluated. Langmuir and the Freundlich adsorption isotherms were selected to explicate the interaction of the dye and magnetic adsorbant. The characteristic parameters for each isotherm have been determined. The overall trend followed an increase of the sorption capacity with increasing dye concentration with a maximum of 90% dye removal. The monolayer adsorption capacity of magnetic hollow spheres (0.05 g) for NR in the concentration range studied, as calculated from the Langmuir isotherm model at 25 deg. C and pH 6, was found to be 105 mg g -1 . Adsorption kinetic followed pseudo-second-order reaction kinetics. Thermodynamic study showed that the adsorption processes are spontaneous and endothermic. The combination of the superior adsorption and the magnetic properties of Fe 3 O 4 nanospheres can be useful as a powerful separation tool to deal with environmental pollution.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  10. 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)

  11. Compatibility Behavior of the Ferritic-Martensitic Steel Cladding under the Liquid Sodium Environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Baek, Jong Hyuk; Kim, Sung Ho; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Fuel cladding is a component which confines uranium fuel to transport energy into the coolant as well as protect radioactive species from releasing outside. Sodium-cooled Fast Reactor (SFR) has been considered as one of the most probable next generation reactors in Korea because it can maximize uranium resource as well as reduce the amount of PWR spent fuel in conjunction with pyroprocessing. Sodium has been selected as the coolant of the SFR because of its superior fast neutron efficiency as well as thermal conductivity, which enables high power core design. However, it is reported that the fuel cladding materials like austenitic and ferritic stainless steel react sodium coolant so that the loss of the thickness, intergranular attack, and carburization or decarburization process may happen to induce the change of the mechanical property of the cladding. This study aimed to evaluate material property of the cladding material under the liquid sodium environment. Ferritic-martensitic steel (FMS) coupon and cladding tube were exposed at the flowing sodium then the microstructural and mechanical property were evaluated. mechanical property of the cladding was evaluated using the ring tension test

  12. 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.

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States)

    1977-04-01

    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.

  15. Creep collapse of TAPS fuel cladding

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Anand, A.K.

    1975-01-01

    Densification of UO 2 can cause axial gaps between fuel pelets and cladding in unsupported (internally) at these regions. An analysis is carried out regarding the possibility of creep collapse in these regions. The analysis is based on Timoshenko's theory of collapse. At various times during the residence of fuel in reactor following parameters are calculated : (1) inelastic collapse of perfectly circular tubes (2) plastic instability in oval tubes (3) effect of creep on ovality. Creep is considered to be a non-linear combination of the following : (a) thermal creep (b) intresenic creep (c) stress aided radiation enhanced (d) stress free growth (4) Critical pressure ratio. The results obtained are compared with G.E. predictions. The results do not predict collapse of TAPS fuel cladding for five year residence time. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-31

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

  18. Electro-coagulation-flotation process for algae removal

    International Nuclear Information System (INIS)

    Gao Shanshan; Yang Jixian; Tian Jiayu; Ma Fang; Tu Gang; Du Maoan

    2010-01-01

    Algae in surface water have been a long-term issue all over the world, due to their adverse influence on drinking water treatment process as well as drinking water quality. The algae removal by electro-coagulation-flotation (ECF) technology was investigated in this paper. The results indicated that aluminum was an excellent electrode material for algae removal as compared with iron. The optimal parameters determined were: current density = 1 mA/cm 2 , pH = 4-7, water temperature = 18-36 deg. C, algae density = 0.55 x 10 9 -1.55 x 10 9 cells/L. Under the optimal conditions, 100% of algae removal was achieved with the energy consumption as low as 0.4 kWh/m 3 . The ECF performed well in acid and neutral conditions. At low initial pH of 4-7, the cell density of algae was effectively removed in the ECF, mainly through the charge neutralization mechanism; while the algae removal worsened when the pH increased (7-10), and the main mechanism shifted to sweeping flocculation and enmeshment. The mechanisms for algae removal at different pH were also confirmed by atomic force microscopy (AFM) analysis. Furthermore, initial cell density and water temperature could also influence the algae removal. Overall, the results indicated that the ECF technology was effective for algae removal, from both the technical and economical points of view.

  19. Laser Cladding of Embedded Sensors for Thermal Barrier Coating Applications

    Directory of Open Access Journals (Sweden)

    Yanli Zhang

    2018-05-01

    Full Text Available The accurate real-time monitoring of surface or internal temperatures of thermal barrier coatings (TBCs in hostile environments presents significant benefits to the efficient and safe operation of gas turbines. A new method for fabricating high-temperature K-type thermocouple sensors on gas turbine engines using coaxial laser cladding technology has been developed. The deposition of the thermocouple sensors was optimized to provide minimal intrusive features to the TBC, which is beneficial for the operational reliability of the protective coatings. Notably, this avoids a melt pool on the TBC surface. Sensors were deposited onto standard yttria-stabilized zirconia (7–8 wt % YSZ coated substrates; subsequently, they were embedded with second YSZ layers by the Atmospheric Plasma Spray (APS process. Morphology of cladded thermocouples before and after embedding was optimized in terms of topography and internal homogeneity, respectively. The dimensions of the cladded thermocouple were in the order of 200 microns in thickness and width. The thermal and electrical response of the cladded thermocouple was tested before and after embedding in temperatures ranging from ambient to approximately 450 °C in a furnace. Seebeck coefficients of bared and embedded thermocouples were also calculated correspondingly, and the results were compared to that of a commercial standard K-type thermocouple, which demonstrates that laser cladding is a prospective technology for manufacturing microsensors on the surface of or even embedded into functional coatings.

  20. First results on the effect of fuel-cladding eccentricity

    International Nuclear Information System (INIS)

    Panka, I.; Kereszturi, A.

    2009-01-01

    In the traditional fuel-behaviour or hot channel calculations it is assumed that the fuel pellet is centered within the clad. However, in the real life the pellet could be positioned asymmetrically within the clad, which leads to asymmetric gap conductance and therefore it is worthwhile to investigate the magnitude of the effect on maximal fuel temperature and surface heat flux. In this paper our first experiences are presented on this topic. (Authors)

  1. Impacts of reactor. Induced cladding defects on spent fuel storage

    International Nuclear Information System (INIS)

    Johnson, A.B.

    1978-01-01

    Defects arise in the fuel cladding on a small fraction of fuel rods during irradiation in water-cooled power reactors. Defects from mechanical damage in fuel handling and shipping have been almost negligible. No commercial water reactor fuel has yet been observed to develop defects while stored in spent fuel pools. In some pools, defective fuel is placed in closed canisters as it is removed from the reactor. However, hundreds of defective fuel bundles are stored in numerous pools on the same basis as intact fuel. Radioactive species carried into the pool from the reactor coolant must be dealt with by the pool purification system. However, additional radiation releases from the defective fuel during storage appear tu be minimal, with the possible exception of fuel discharged while the reactor is operating (CANDU fuel). Over approximately two decades, defective commercial fuel has been handled, stored, shipped and reprocessed. (author)

  2. SiC interlayer by laser-cladding on WC-Co substrates for CVD diamond deposition

    Energy Technology Data Exchange (ETDEWEB)

    Contin, Andre; Fraga, Mariana Amorim; Vieira, Jose; Trava-Airoldi, Vladimir Jesus; Corat, Evaldo Jose, E-mail: andrecontin@yahoo.com.br [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil); Campos, Raonei Alves [Universidade Federal do Sul e Sudeste do Para (UNIFESSPA), Belem, PA (Brazil); Vasconcelos, Getulio [Instituto de Estudos Avancados (IEA), Sao Jose dos Campos, SP (Brazil)

    2016-07-01

    Full text: Despite their huge industrial potential and commercial interest, the direct diamond coating on cemented carbide (WC-Co) is limited, mainly because of the catalytic effect of Cobalt (Co) and the high difference in thermal expansion coefficient [1]. This results in poor adherence between diamond and WC-Co. In addition, the low diamond film adhesion to the cemented carbide useless for machining applications. Removal of Co binder from the substrate surface by superficial etching is one of the techniques used to improve the adhesion between diamond and WC-Co. For the present study, diamond films were deposited on WC-Co substrates with an intermediate barrier to block the Co diffusion to the surface substrate. The laser cladding process produced the SiC barrier, in which a powder layer is melted by a laser irradiation to create the coating on the substrate. The use of laser cladding is the novel method for an intermediate barrier for cemented carbides. The advantages of laser cladding include a faster processing speed, precision, versatility. We reported the application of pretreatment method called ESND (Electrostatic self-assembly seeding of nanocrystalline diamond). The nucleation density was around 10{sup 11}part/cm{sup 2}. Diamond films were grown by Hot Filament Chemical Vapor Deposition. Characterization of samples included Field Emission Gun-Scanning Electron Microscopy (FEG-SEM), Energy Dispersive X-ray (EDX), X-ray diffraction (XRD) and Raman Scattering Spectroscopy. Results showed that laser irradiation formed stable Co compounds in the interfacial barrier. It is because nucleation and good quality of diamond film since the cobalt are no longer free to migrate to the surface during the CVD diamond deposition. Reference: [1] Y. X. Cui, B. Shen, F. H. Sun. Diamond deposition on WC–Co substrate with amorphous SiC interlayer, Surface Engineering, 30, (2014) 237-243. (author)

  3. Charging induced emission of neutral atoms from NaCl nanocube corners

    International Nuclear Information System (INIS)

    Ceresoli, Davide; Zykova-Timan, Tatyana; Tosatti, Erio

    2008-01-01

    Detachment of neutral cations/anions from solid alkali halides can in principle be provoked by donating/subtracting electrons to the surface of alkali halide crystals, but generally constitutes a very endothermic process. However, the amount of energy required for emission is smaller for atoms located in less favorable positions, such as surface steps and kinks. For a corner ion in an alkali halide cube the binding is the weakest, so it should be easier to remove that atom, once it is neutralized. We carried out first principles density functional calculations and simulations of neutral and charged NaCl nanocubes, to establish the energetics of extraction of neutralized corner ions. Following hole donation (electron removal) we find that detachment of neutral Cl corner atoms will require a limited energy of about 0.8 eV. Conversely, following the donation of an excess electron to the cube, a neutral Na atom is extractable from the corner at the lower cost of about 0.6 eV. Since the cube electron affinity level (close to that a NaCl(100) surface state, which we also determine) is estimated to lie about 1.8 eV below vacuum, the overall energy balance upon donation to the nanocube of a zero-energy electron from vacuum will be exothermic. The atomic and electronic structure of the NaCl(100) surface, and of the nanocube Na and Cl corner vacancies are obtained and analyzed as a byproduct

  4. Investigations of chemical reactions between U-Zr alloy and FBR cladding materials

    International Nuclear Information System (INIS)

    Ishii, Tetsuya; Ukai, Shigeharu

    2005-07-01

    U-Pu-Zr alloys are candidate materials for commercial FBR fuel. However, informations about chemical reactions with cladding materials developed by JNC for commercial FBR have not been well obtained. In this work, the reaction zones formed in four diffusion couples U-10wt.%Zr/PNC-FMS, U-10wt.%Zr/9Cr-ODS, U-10wt.%Zr/12Cr-ODS, and U-10wt.%Zr/Fe at about 1013K have been examined and following results were obtained. 1) At about 1013K, in the U-10wt.%Zr/Fe couple, the liquid phase zones were obtained. In the other couples U-10wt.%Zr/PNC-FMS, U-10wt.%Zr/9Cr-ODS and U-10wt.%Zr/12Cr-ODS, no liquid phase zones were obtained. The obtained chemical reaction zones in the later 3 couples were similar to the reported ones obtained in U-Zr/Fe couples without liquid phase formation. In comparison with the reaction zones obtained in the U-10wt.%Zr/Fe couple, the reaction zones inside cladding materials obtained in the PNC-FMS, 9Cr-ODS, and 12Cr-ODS couples were thin. 2) From the investigations of relationship between the obtained depths of the chemical reaction zones inside cladding materials and composition of the cladding materials, it was considered that the depth of chemical reaction zone would depend on the Cr content of the cladding materials and the depth would decrease with increasing Cr content, resulting in prevention of liquid phase formation. 3) From the investigations of the mechanisms of chemical reactions between U-Pu-Zr/cladding materials, it was considered that the same effect of Cr obtained in the U-Zr/cladding materials would be expected in U-Pu-Zr/cladding materials. Those seemed to indicate that the threshold temperatures of liquid phase formation for U-Pu-Zr/PNC-FMS, U-Pu-Zr/9Cr-ODS, and U-Pu-Zr/12Cr-ODS might be higher than that for U-Pu-Zr/Fe. (author)

  5. Inner wall attack and its inhibition method for FBR fuel pin cladding at high burnup

    International Nuclear Information System (INIS)

    Xu Yongli; Long Bin; Li Jingang; Wan Jiaying

    1998-01-01

    The inner wall attack of the modified 316-Ti S.S. cladding tubes manufactured in China used FBR at 10at.% burnup was investigated by means of the out of pile simulation tests. The inner surface morphologies of the cladding tubes attached by fission products Cs, Te, I and Se at 700 deg. C under lower and high oxygen potentials were observed respectively, and the depth of attack was also measured. The burst strength, maximum circum expansion and the appearances of fracture were measured and observed respectively for the cladding tubes attacked by fission products. Based on the mechanism of FBR fuel cladding chemical interaction (FCCI), Cr, Zr and Nb were used as the oxygen absorbers respectively, in order to inhibit the inner wall attack of the cladding tubes. The corrosion morphologies and depth, the penetration depth of the fission products in the inner surface of the cladding tubes were detected. The inhibition effectiveness of the oxygen absorbers for the inner wall attack of the cladding tubes was evaluated. (author)

  6. Modelling of ultrasonic nondestructive testing of cracks in claddings

    Energy Technology Data Exchange (ETDEWEB)

    Bostroem, Anders; Zagbai, Theo [Calmers Univ. of Technology, Goeteborg (Sweden). Dept. of Applied Mechanics

    2006-05-15

    Nondestructive testing with ultrasound is a standard procedure in the nuclear power industry. To develop and qualify the methods extensive experimental work with test blocks is usually required. This can be very time-consuming and costly and it also requires a good physical intuition of the situation. A reliable mathematical model of the testing situation can, therefore, be very valuable and cost-effective as it can reduce experimental work significantly. A good mathematical model enhances the physical intuition and is very useful for parametric studies, as a pedagogical tool, and for the qualification of procedures and personnel. The present project has been concerned with the modelling of defects in claddings. A cladding is a layer of material that is put on for corrosion protection, in the nuclear power industry this layer is often an austenitic steel that is welded onto the surface. The cladding is usually anisotropic and to some degree it is most likely also inhomogeneous, particularly in that the direction of the anisotropy is varying. This degree of inhomogeneity is unknown but probably not very pronounced so for modelling purposes it may be a valid assumption to take the cladding to be homogeneous. However, another important complicating factor with claddings is that the interface between the cladding and the base material is often corrugated. This corrugation can have large effects on the transmission of ultrasound through the interface and can thus greatly affect the detectability of defects in the cladding. In the present project the only type of defect that is considered is a planar crack that is situated inside the cladding. The investigations are, furthermore, limited to two dimensions, and the crack is then only a straight line. The crack can be arbitrarily oriented and situated, but it must not intersect the interface to the base material. The crack can be surface-breaking, and this is often the case of most practical interest, but it should then be

  7. Heliospheric Neutral Atom Spectra Between 0.01 and 6 keV fom IBEX

    Science.gov (United States)

    Fuselier, S. A.; Allegrini, F.; Bzowski, M.; Funsten, H. O.; Ghielmetti, A. G.; Gloeckler, G.; Heirtzler, D.; Janzen, P.; Kubiak, M.; Kucharek, H.; hide

    2012-01-01

    Since 2008 December, the Interstellar Boundary Explorer (IBEX) has been making detailed observations of neutrals from the boundaries of the heliosphere using two neutral atom cameras with overlapping energy ranges. The unexpected, yet defining feature discovered by IBEX is a Ribbon that extends over the energy range from about 0.2 to 6 keV. This Ribbon is superposed on a more uniform, globally distributed heliospheric neutral population. With some important exceptions, the focus of early IBEX studies has been on neutral atoms with energies greater than approx. 0.5 keV. With nearly three years of science observations, enough low-energy neutral atom measurements have been accumulated to extend IBEX observations to energies less than approx. 0.5 keV. Using the energy overlap of the sensors to identify and remove backgrounds, energy spectra over the entire IBEX energy range are produced. However, contributions by interstellar neutrals to the energy spectrum below 0.2 keV may not be completely removed. Compared with spectra at higher energies, neutral atom spectra at lower energies do not vary much from location to location in the sky, including in the direction of the IBEX Ribbon. Neutral fluxes are used to show that low energy ions contribute approximately the same thermal pressure as higher energy ions in the heliosheath. However, contributions to the dynamic pressure are very high unless there is, for example, turbulence in the heliosheath with fluctuations of the order of 50-100 km/s.

  8. CO2 laser-fabricated cladding light strippers for high-power fiber lasers and amplifiers.

    Science.gov (United States)

    Boyd, Keiron; Simakov, Nikita; Hemming, Alexander; Daniel, Jae; Swain, Robert; Mies, Eric; Rees, Simon; Andrew Clarkson, W; Haub, John

    2016-04-10

    We present and characterize a simple CO2 laser processing technique for the fabrication of compact all-glass optical fiber cladding light strippers. We investigate the cladding light loss as a function of radiation angle of incidence and demonstrate devices in a 400 μm diameter fiber with cladding losses of greater than 20 dB for a 7 cm device length. The core losses are also measured giving a loss of cladding light stripping of a 300 W laser diode with minimal heating of the fiber coating and packaging adhesives.

  9. Modelling the gas transport and chemical processes related to clad oxidation and hydriding

    Energy Technology Data Exchange (ETDEWEB)

    Montgomery, R O; Rashid, Y R [ANATECH Research Corp., San Diego, CA (United States)

    1997-08-01

    Models are developed for the gas transport and chemical processes associated with the ingress of steam into a LWR fuel rod through a small defect. These models are used to determine the cladding regions in a defective fuel rod which are susceptible to massive hydriding and the creation of sunburst hydrides. The brittle nature of zirconium hydrides (ZrH{sub 2}) in these susceptible regions produces weak spots in the cladding which can act as initiation sites for cladding cracks under certain cladding stress conditions caused by fuel cladding mechanical interaction. The modeling of the axial gas transport is based on gaseous bimolar diffusion coupled with convective mass transport using the mass continuity equation. Hydrogen production is considered from steam reaction with cladding inner surface, fission products and internal components. Eventually, the production of hydrogen and its diffusion along the length results in high hydrogen concentration in locations remote from the primary defect. Under these conditions, the hydrogen can attack the cladding inner surface and breakdown the protective ZrO{sub 2} layer locally, initiating massive localized hydriding leading to sunburst hydride. The developed hydrogen evolution model is combined with a general purpose fuel behavior program to integrate the effects of power and burnup into the hydriding kinetics. Only in this manner can the behavior of a defected fuel rod be modeled to determine the conditions the result in fuel rod degradation. (author). 14 refs, 6 figs.

  10. FRACAS: a subcode for the analysis of fuel pellet-cladding mechanical interaction

    International Nuclear Information System (INIS)

    Bohn, M.P.

    1977-04-01

    This report describes FRACAS (Fuel Rod and Cladding Analysis Subcode), a computer code which performs the mechanical analysis in the FRAP fuel rod codes. At each loadstep, FRACAS obtains a complete elastic-plastic-creep solution for the stresses, strains, and displacements in the fuel rod cladding. The cladding is modeled as a thin cylindrical shell with prescribed temperature, pressures, and radial displacement of the inside surface. The displacement of the fuel pellets is assumed to be due to thermal gradients only. Three different regimes of pellet-cladding mechanical interaction are considered: (a) open gap, (b) closed gap, and (c) trapped stack. Both transient and steady state creep calculations are performed. The capabilities of the code are illustrated by an example problem, and comparisons are made with data obtained from two experimental fuel rods

  11. 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

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

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.

    1998-01-01

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

  13. 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

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

    Science.gov (United States)

    Lo, Wei-Yang; Yang, Yong

    2014-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-01

    Fuel cladding chemical interaction (FCCI) has been identified as one of the crucial issues for developing Ferritic/Martensitic (F/M) stainless steel claddings for metallic fuels in a fast reactor. The anticipated elevated temperature and high neutron flux can significantly aggravate the FCCI, in terms of formation of inter-diffusion and lower melting point eutectic phases. To mitigate the FCCI, vanadium carbide coating as a diffusion barrier was deposited on the HT-9 substrate using a pack cementation diffusion coating (PCDC) method, and the processing temperature was optimized down to 730 °C. A solid metallurgical bonding between the coating layer and substrate was achieved, and the coating is free from through depth cracks. The microstructural characterizations using SEM and TEM show a nanostructured grain structure. EDS/WDS and XRD analysis confirm the phase of coating layer as V{sub 2}C. Diffusion couple tests at 660 °C for 100 h demonstrate that V{sub 2}C layer with a thickness of less than 5 μm can effectively eliminate the inter-diffusion between the lanthanide cerium and HT-9 steel.

  16. 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

  17. Modelling of pellet-cladding interaction for PWRs reactors fuel rods

    International Nuclear Information System (INIS)

    Esteves, A.M.

    1991-01-01

    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 analyzes 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. Linear and non-linear material behaviors are allowed. Elastic, plastic and creep behaviors are considered for the cladding materials. The modelling is applied to Angra-II fuel rod design. The results are analyzed and compared. (author)

  18. Fuel Performance Calculations for FeCrAl Cladding in BWRs

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  20. 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.)

  1. 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.

  2. Mechanical and temperature contact in fuel rod cladding

    International Nuclear Information System (INIS)

    Fredriksson, B.E.; Rydholm, S.G.

    1977-01-01

    The paper presents results for the effect of different types of slip rules on the contact stress distribution. It is shown that the contact shear stress is smaller for the hardening model than for the ideal model. It is also shown that a crack in the fuel increases the contact stresses and that at temperature decrease high tensile stresses arise after eventual welding. It is also shown how particles between fuel and cladding influence the stresses. Also here the effect of eventual welding is studied. The present method is well suited to study cracks and crack propagation. The surfaces of the existing cracks are defined as contact surfaces and the crack extension work is calculated by releasing the nodes at the crack tip. As the crack surfaces are defined as contact surfaces eventual crack closure is automatically taken into account. Crack extension work is calculated for existing cracks in the cladding. It is shown that cracks in the fuel and particles between fuel and cladding will increase the crack extension work

  3. Cladding oxidation during air ingress. Part II: Synthesis of modelling results

    International Nuclear Information System (INIS)

    Beuzet, E.; Haurais, F.; Bals, C.; Coindreau, O.; Fernandez-Moguel, L.; Vasiliev, A.; Park, S.

    2016-01-01

    Highlights: • A state-of-the-art for air oxidation modelling in the frame of severe accident is done. • Air oxidation models from main severe accident codes are detailed. • Simulations from main severe accident codes are compared against experimental results. • Perspectives in terms of need for further model development and experiments are given. - Abstract: Air ingress is a potential risk in some low probable situations of severe accidents in a nuclear power plant. Air is a highly oxidizing atmosphere that can lead to an enhanced Zr-based cladding oxidation and core degradation affecting the release of fission products. This is particularly true speaking about ruthenium release, due to its high radiotoxicity and its ability to form highly volatile oxides in a significant manner in presence of air. The oxygen affinity is decreasing from the Zircaloy cladding, fuel and ruthenium inclusions. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues in such scenarios. In the past years, many works have been done on cladding oxidation by air under severe accident conditions. This paper with in addition the paper “Cladding oxidation during air ingress – Part I: Synthesis of experimental results” of this journal issue aim at assessing the state of the art on this phenomenon. In this paper, the modelling of air ingress phenomena in the main severe accident codes (ASTEC, ATHLET-CD, MAAP, MELCOR, RELAP/SCDAPSIM, SOCRAT) is described in details, as well as the validation against the integral experiments QUENCH-10, QUENCH-16 and PARAMETER-SF4. A full review of cladding oxidation by air is thus established.

  4. Interdiffusion between U-Zr-Mo and stainless steel cladding

    International Nuclear Information System (INIS)

    Hwang, J. Y.; Lee, B. S.; Lee, J. T.; Kang, Y. H.

    1998-01-01

    Interdiffusion investigations were carried out at 700 deg C for 200 hours for the diffusion couples assembled with the U-Zr-Mo ternary fuel versus austenitic stainless steel D9 and the U-Zr-Mo ternary fuel versus martensitic stainless steel HT9 respectively to investigate the fuel-cladding compatibility. SEM-EDS analysis was utilized to determine the composition and the penetration depths of the reaction layers. In the case of Fuel/D9 couple, (Fe, Cr, Ni) of the cladding elements formed the precipitates with the Zr, Mo and diminished the U concentration upto 800μ length from the fuel side. Composition of the precipitates was varied with the penetrated elements. In Fuel/HT9 couple, reaction layer was smaller than that of D9 couples and was less affected by cladding elements. The eutectic reaction appeared partially in the Fuel/HT9 diffusion couple

  5. In situ synthesis of hydroxyapatite coating by laser cladding.

    Science.gov (United States)

    Wang, D G; Chen, C Z; Ma, J; Zhang, G

    2008-10-15

    HA bioceramic coatings were synthesized on titanium substrate by laser cladding using cheap calcium carbonate and calcium hydrogen phosphate. The thermodynamic condition for synthesizing HA was calculated by software Matlab 5.0, the microstructure and phase analysis of laser clad HA bioceramic coatings were studied by electron probe microanalyser (EPMA), X-ray diffractometer (XRD) and transmission electron microscopy (TEM). The theoretical results show that the Gibbs free enthalpy for the synthesis of HA phase is satisfied, and the presence of HA phase in the clad coatings was then further verified by XRD and the selected area diffraction patterns. When the laser power is 600W and the scanning speed is 3.5mm/s, the compact HA bioceramic coatings were obtained, which have cellular dendritic structure and consist of the phases of HA, alpha-Ca(2)P(2)O(7), CaO and CaTiO(3).

  6. ITER neutral beam system US conceptual design

    International Nuclear Information System (INIS)

    Purgalis, P.

    1990-09-01

    In this document we present the US conceptual design of a neutral beam system for International Thermonuclear Experimental Reactor (ITER). The design incorporates a barium surface conversion D - source feeding a linear array of accelerator channels. The system uses a dc accelerator with electrostatic quadrupoles for strong focusing. A high voltage power supply that is integrated with the accelerator is presented as an attractive option. A gas neutralizer is used and residual ions exiting the neutralizer are deflected to water-cooled dumps. Cryopanels are located at the accelerator exit to pump excess gas from the source and the neutralizer, and in the ion dump cavity to pump re-neutralized ions and neutralizer gas. All the above components are packaged in compact identical, independent modules which can be removed for remote maintenance. The neutral beam system delivers 75 MW of DO at 1.3 MeV, into three ports with a total of 9 modules arranged in stacks of three modules per port . To increase reliability each module is designed to deliver up to 10 MW; this allows eight modules operating at partial capacity to deliver the required power in the event one module is out of service, and provides 20% excess capacity to improve availability. Radiation protection is provided by shielding and by locating critical components in the source and accelerator 46.5 m from the torus centerline. Neutron shielding in the drift duct and neutralizer provides the added feature of limiting conductance and thus reducing gas flow to and from the torus

  7. A comparative study on the fretting wear properties of advanced zirconium fuel cladding materials

    International Nuclear Information System (INIS)

    Lee, Young Ho; Kim, Hyung Kyu; Park, Jeong Yong; Kim, Jun Hwan

    2005-06-01

    Fretting wear tests were carried out in room and high temperature water in order to evaluate the wear properties of new zirconium nuclear fuel claddings (K2∼K6) and the commercial claddings (M5, zirlo and zircaloy-4). The objective is to compare the wear resistance of K2∼K6 claddings with that of the commercial ones at the same test condition. After the wear tests, the average wear volume and the maximum wear depth were evaluated and compared at each test condition. As a result, it is difficult to select the most wear-resistant cladding between the K2∼K6 claddings and the commercial ones. This is because the average wear volume and maximum depth of each cladding included between the scattering range of measured results. However, wear resistance of the tested claddings based on the average wear volume and maximum wear depth could be summarized as follows: K5 > zircaloy-4 > (K2,K3) > (K4,M5) > K6 > zirlo at room temperature, zircaloy-4 > K5 > (K3,K4,zirlo) > (K2,K6) > M5 at high temperature and pressure. Therefore, it is concluded that K5 cladding among the tested new zirconium alloys has relatively higher wear-resistance in room and high temperature condition. In order to examine the wear mechanism, it is necessary to systematically study with the consideration of the alloying element effect and test environment. In this report, the wear test procedure and the wear evaluation method are described in detail

  8. Shielding gas effect to diffusion activities of magnesium and copper on aluminum clad

    Science.gov (United States)

    Manurung, Charles SP; Napitupulu, Richard AM

    2017-09-01

    Aluminum is the second most metal used in many application, because of its corrosion resistance. The Aluminum will be damaged in over time if it’s not maintained in good condition. That is important to give protection to the Aluminums surface. Cladding process is one of surface protection methodes, especially for metals. Aluminum clad copper (Al/Cu) or copper clad aluminum (Cu/Al) composite metals have been widely used for many years. These mature protection method and well tested clad metal systems are used industrially in a variety application. The inherent properties and behavior of both copper and aluminum combine to provide unique performance advantages. In this paper Aluminum 2024 series will be covered with Aluminum 1100 series by hot rolling process. Observations will focus on diffusion activities of Mg and Cu that not present on Aluminum 1100 series. The differences of clad material samples is the use of shielding gas during heating before hot rolling process. The metallurgical characteristics will be examined by using optical microscopy. Transition zone from the interface cannot be observed but from Energy Dispersive Spectrometry it’s found that Mg and Cu are diffused from base metal (Al 2024) to the clad metal (Al 1100). Hardness test proved that base metals hardness to interface was decrease.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-09-01

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

  10. Method for the protection of the cladding tubes of fuel rods

    International Nuclear Information System (INIS)

    Steinberg, E.

    1978-01-01

    To present stress crack corrosion and to protect the cladding tubes of the fuel rods made of a circonium alloy from attack by iodine, the inward surfaces are provided with protective coatings. Therefore the casting tubes already filled with fuel element pellets are put under over-pressure at a temperature range between 300 and 500 0 C, until almost yield-point is reached. A small amount of H 2 O or H 2 O 2 , filled in, reacts with the cladding tube material to form the Zr-O 2 protective coating. Afterwards comes a pressure relief, and the cladding tube reaches its original dimensions. (DG) [de

  11. High power operation of cladding pumped holmium-doped silica fibre lasers.

    Science.gov (United States)

    Hemming, Alexander; Bennetts, Shayne; Simakov, Nikita; Davidson, Alan; Haub, John; Carter, Adrian

    2013-02-25

    We report the highest power operation of a resonantly cladding-pumped, holmium-doped silica fibre laser. The cladding pumped all-glass fibre utilises a fluorine doped glass layer to provide low loss cladding guidance of the 1.95 µm pump radiation. The operation of both single mode and large-mode area fibre lasers was demonstrated, with up to 140 W of output power achieved. A slope efficiency of 59% versus launched pump power was demonstrated. The free running emission was measured to be 2.12-2.15 µm demonstrating the potential of this architecture to address the long wavelength operation of silica based fibre lasers with high efficiency.

  12. A systematic approach for development of a PWR cladding corrosion model

    International Nuclear Information System (INIS)

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

    2001-01-01

    A new model for the in-reactor corrosion of Improved (low-tin) Zircaloy-4 cladding irradiated in commercial pressurized water reactors (PWRs) is described. The model is based on an extensive database of PWR fuel cladding corrosion data from fuel irradiated in commercial reactors, with a range of fuel duty and coolant chemistry control strategies which bracket current PWR fuel management practices. The fuel thermal duty with these current fuel management practices is characterized by a significant amount of sub-cooled nucleate boiling (SNB) during the fuel's residence in-core, and the cladding corrosion model is very sensitive to the coolant heat transfer models used to calculate the coolant temperature at the oxide surface. The systematic approach to developing the new corrosion model therefore began with a review and evaluation of several alternative models for the forced convection and SNB coolant heat transfer. The heat transfer literature is not sufficient to determine which of these heat transfer models is most appropriate for PWR fuel rod operating conditions, and the selection of the coolant heat transfer model used in the new cladding corrosion model has been coupled with a statistical analysis of the in-reactor corrosion enhancement factors and their impact on obtaining the best fit to the cladding corrosion data. The in-reactor corrosion enhancement factors considered in this statistical analysis are based on a review of the current literature for PWR cladding corrosion phenomenology and models. Fuel operating condition factors which this literature review indicated could have a significant effect on the cladding corrosion performance were also evaluated in detail in developing the corrosion model. An iterative least squares fitting procedure was used to obtain the model coefficients and select the coolant heat transfer models and in-reactor corrosion enhancement factors. This statistical procedure was completed with an exhaustive analysis of the model

  13. Evaluation of titanium carbide metal matrix composites deposited via laser cladding

    Science.gov (United States)

    Cavanaugh, Daniel Thomas

    Metal matrix composites have been widely studied in terms of abrasion resistance, but a particular material system may behave differently as particle size, morphology, composition, and distribution of the hardening phase varies. The purpose of this thesis was to understand the mechanical and microstructural effects of combining titanium carbide with 431 series stainless steel to create a unique composite via laser cladding, particularly regarding wear properties. The most predominant effect in increasing abrasion resistance, measured via ASTM G65, was confirmed to be volume fraction of titanium carbide addition. Macrohardness was directly proportional to the amount of carbide, though there was an overall reduction in individual particle microhardness after cladding. The reduction in particle hardness was obscured by the effect of volume fraction carbide and did not substantially contribute to the wear resistance changes. A model evaluating effective mean free path of the titanium carbide particles was created and correlated to the measured data. The model proved successful in linking theoretical mean free path to overall abrasion resistance. The effects of the titanium carbide particle distributions were limited, while differences in particle size were noticeable. The mean free path model did not correlate well with the particle size, but it was shown that the fine carbides were completely removed by the coarse abrasive particles in the ASTM G65 test. The particle morphology showed indications of influencing the wear mode, but no statistical reduction was observed in the volume loss figures. Future studies may more specifically focus on particle morphology or compositional effects of the carbide particles.

  14. 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

  15. 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.

  16. Transmission of laser pulses with high output beam quality using step-index fibers having large cladding

    Science.gov (United States)

    Yalin, Azer P; Joshi, Sachin

    2014-06-03

    An apparatus and method for transmission of laser pulses with high output beam quality using large core step-index silica optical fibers having thick cladding, are described. The thick cladding suppresses diffusion of modal power to higher order modes at the core-cladding interface, thereby enabling higher beam quality, M.sup.2, than are observed for large core, thin cladding optical fibers. For a given NA and core size, the thicker the cladding, the better the output beam quality. Mode coupling coefficients, D, has been found to scale approximately as the inverse square of the cladding dimension and the inverse square root of the wavelength. Output from a 2 m long silica optical fiber having a 100 .mu.m core and a 660 .mu.m cladding was found to be close to single mode, with an M.sup.2=1.6. Another thick cladding fiber (400 .mu.m core and 720 .mu.m clad) was used to transmit 1064 nm pulses of nanosecond duration with high beam quality to form gas sparks at the focused output (focused intensity of >100 GW/cm.sup.2), wherein the energy in the core was laser pulses was about 6 ns. Extending the pulse duration provided the ability to increase the delivered pulse energy (>20 mJ delivered for 50 ns pulses) without damaging the silica fiber.

  17. Research on microstructure properties of the TiC/Ni-Fe-Al coating prepared by laser cladding technology

    Science.gov (United States)

    Jiao, Junke; Xu, Zifa; Zan, Shaoping; Zhang, Wenwu; Sheng, Liyuan

    2017-10-01

    In this paper, the laser cladding method was used to preparation the TiC reinforced Ni-Fe-Al coating on the Ni base superalloy. The Ti/Ni-Fe-Al powder was preset on the Ni base superalloy and the powder layer thickness is 0.5mm. A fiber laser was used the melting Ti/Ni-Fe-Al powder in an inert gas environment. The shape of the cladding layer was tested using laser scanning confocal microscope (LSCM) under different cladding parameters such as the laser power, the melting velocity and the defocused amount. The microstructure, the micro-hardness was tested by LSCM, SEM, Vickers hardness tester. The test result showed that the TiC particles was distributed uniformly in the cladding layer and hardness of the cladding layer was improved from 180HV to 320HV compared with the Ni-Fe-Al cladding layer without TiC powder reinforced, and a metallurgical bonding was produced between the cladding layer and the base metal. The TiC powder could make the Ni-Fe-Al cladding layer grain refining, and the more TiC powder added in the Ni-Fe-Al powder, the smaller grain size was in the cladding layer.

  18. Transparent conducting oxide clad limited area epitaxy semipolar III-nitride laser diodes

    KAUST Repository

    Myzaferi, A.; Reading, A. H.; Cohen, D. A.; Farrell, R. M.; Nakamura, S.; Speck, J. S.; DenBaars, S. P.

    2016-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-01

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

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

    International Nuclear Information System (INIS)

    Perez, Emmanuel; Keiser Jr, Dennis D.; Forsmann, Bryan; Janney, Dawn E.; Henley, Jody; Woolstenhulme, Eric C.

    2016-01-01

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

  1. 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

  2. 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

  3. 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

  4. Flux Density through Guides with Microstructured Twisted Clad DB Medium

    Directory of Open Access Journals (Sweden)

    M. A. Baqir

    2014-01-01

    Full Text Available The paper deals with the study of flux density through a newly proposed twisted clad guide containing DB medium. The inner core and the outer clad sections are usual dielectrics, and the introduced twisted windings at the core-clad interface are treated under DB boundary conditions. The pitch angle of twist is supposed to greatly contribute towards the control over the dispersion characteristics of the guide. The eigenvalue equation for the guiding structure is deduced, and the analytical investigations are made to explore the propagation patterns of flux densities corresponding to the sustained low-order hybrid modes under the situation of varying pitch angles. The emphasis has been put on the effects due to the DB twisted pitch on the propagation of energy flux density through the guide.

  5. 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

  6. 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

  7. 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

  8. Heparin removal by ecteola-cellulose pre-treatment enables the use of plasma samples for accurate measurement of anti-Yellow fever virus neutralizing antibodies.

    Science.gov (United States)

    Campi-Azevedo, Ana Carolina; Peruhype-Magalhães, Vanessa; Coelho-Dos-Reis, Jordana Grazziela; Costa-Pereira, Christiane; Yamamura, Anna Yoshida; Lima, Sheila Maria Barbosa de; Simões, Marisol; Campos, Fernanda Magalhães Freire; de Castro Zacche Tonini, Aline; Lemos, Elenice Moreira; Brum, Ricardo Cristiano; de Noronha, Tatiana Guimarães; Freire, Marcos Silva; Maia, Maria de Lourdes Sousa; Camacho, Luiz Antônio Bastos; Rios, Maria; Chancey, Caren; Romano, Alessandro; Domingues, Carla Magda; Teixeira-Carvalho, Andréa; Martins-Filho, Olindo Assis

    2017-09-01

    Technological innovations in vaccinology have recently contributed to bring about novel insights for the vaccine-induced immune response. While the current protocols that use peripheral blood samples may provide abundant data, a range of distinct components of whole blood samples are required and the different anticoagulant systems employed may impair some properties of the biological sample and interfere with functional assays. Although the interference of heparin in functional assays for viral neutralizing antibodies such as the functional plaque-reduction neutralization test (PRNT), considered the gold-standard method to assess and monitor the protective immunity induced by the Yellow fever virus (YFV) vaccine, has been well characterized, the development of pre-analytical treatments is still required for the establishment of optimized protocols. The present study intended to optimize and evaluate the performance of pre-analytical treatment of heparin-collected blood samples with ecteola-cellulose (ECT) to provide accurate measurement of anti-YFV neutralizing antibodies, by PRNT. The study was designed in three steps, including: I. Problem statement; II. Pre-analytical steps; III. Analytical steps. Data confirmed the interference of heparin on PRNT reactivity in a dose-responsive fashion. Distinct sets of conditions for ECT pre-treatment were tested to optimize the heparin removal. The optimized protocol was pre-validated to determine the effectiveness of heparin plasma:ECT treatment to restore the PRNT titers as compared to serum samples. The validation and comparative performance was carried out by using a large range of serum vs heparin plasma:ECT 1:2 paired samples obtained from unvaccinated and 17DD-YFV primary vaccinated subjects. Altogether, the findings support the use of heparin plasma:ECT samples for accurate measurement of anti-YFV neutralizing antibodies. Copyright © 2017 Elsevier B.V. All rights reserved.

  9. Assessment of thin-walled cladding tube mechanical properties by segmented expanding Mandrel test

    International Nuclear Information System (INIS)

    Nilsson, Karl-Fredrik

    2015-01-01

    This paper presents the principles of the segmented expanding mandrel test for thin-walled cladding tubes, which can be used as a basic material characterisation test to determine stress-strain curves and ductility or as a test to simulate mechanical pellet-cladding interaction. The paper discusses the strengths and weaknesses of the test method and it illustrates how the test can be used to simulate hydride reorientations in zirconium claddings and quantify how hydride reorientation affects ductility. (authors)

  10. 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.

  11. An example of coupling behaviour-damage-environment in polycrystals. Application to Pellet-Cladding Interaction

    International Nuclear Information System (INIS)

    Diard, Olivier

    2001-01-01

    Zircaloy-4 cladding is the first containment barrier for fission products, and its integrity must therefore be ensured in nominal and accidental situations. However, stress corrosion induced cracks may appear due to a strong pellet-cladding interaction. It is therefore important to model this interaction and crack growth and propagation to establish non-damage criteria. Thus, this research thesis aims at developing a modelling covering both issues (pellet-cladding interaction, and stress corrosion cracking) and allowing macroscopic and microscopic scales to be coupled. After a bibliographical synthesis on iodine-induced stress corrosion cracking and similar phenomena, the author presents the model proposed for the pellet-cladding interaction: phenomena to be taken into account, phenomenological and macroscopic behaviour laws used respectively for pellet and cladding. An extended version of an existing cladding viscoplastic model is proposed. Stress and strain fields in the cladding are obtained, notably in the contact zone. In the next part, the author presents various numerical tools developed or used to model multi-crystalline aggregates, and the model of crystalline plasticity used to simulate cladding behaviour at the microstructure scale. Effects of mesh density, element types and anisotropic elasticity are also discussed. The next chapter addresses the mechanical-chemical coupling. Some coupling formulas are presented for simple cases in order to define the effective diffusion coefficient. The last part reports the modelling of intergranular damage: definition of a damage criterion at the granular scale, assessment of stresses at grain boundaries, and effect of crystallographic neighbouring. A model of grain boundary damage is also proposed. This model is assessed on Failure Mechanics test samples and on simple microstructures. The application of the whole numerical model is reported [fr

  12. 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.

  13. Experimental study of residual stresses in laser clad AISI P20 tool steel on pre-hardened wrought P20 substrate

    International Nuclear Information System (INIS)

    Chen, J.-Y.; Conlon, K.; Xue, L.; Rogge, R.

    2010-01-01

    Research highlights: → Laser cladding of P20 tool steel. → Residual stress analysis of laser clad P20 tool steel. → Microstructure of laser clad P20 tool steel. → Tooling Repair using laser cladding. → Stress reliving treatment of laser clad P20 tool steel. - Abstract: Laser cladding is to deposit desired material onto the surface of a base material (or substrate) with a relatively low heat input to form a metallurgically sound and dense clad. This process has been successfully applied for repairing damaged high-value tooling to reduce their through-life cost. However, laser cladding, which needs to melt a small amount of a substrate along with cladding material, inevitably introduces residual stresses in both clad and substrate. The tensile residual stresses in the clad could adversely affect mechanical performance of the substrate being deposited. This paper presents an experimental study on process-induced residual stresses in laser clad AISI P20 tool steel onto pre-hardened wrought P20 base material and the correlation with microstructures using hole-drilling and neutron diffraction methods. Combined with X-ray diffraction and scanning electron microscopic analyses, the roles of solid-state phase transformations in the clad and heat-affected zone (HAZ) of the substrate during cladding and post-cladding heat treatments on the development and controllability of residual stresses in the P20 clad have been investigated, and the results could be beneficial to more effective repair of damaged plastic injection molds made by P20 tool steel.

  14. Cladding nuclear steels - the application of plasma-arc hot wire surfacing

    International Nuclear Information System (INIS)

    Trarbach, K.O.

    1981-01-01

    The effect of one and two layer plasma-arc hot wire cladding on the HAZ microstructure of the fine grained structural steel 22 NiMoCr 3 7, which is similar to ASTM A 508, class 2, and steel 20 MnMoNi 5 5, similar to ASTM A 533, grade B, class 1 is determined. Attention is directed particularly to the behaviour of the susceptible region, and the consumables considered are cladding materials X 2 CrNiNb 19 9, similar to ER 347 Elc, and S-NiCr 20 Nb, similar to ER NiCr-3 (Inconel 82). Results of corrosion resistance tests show that this cladding technique can be recommended for manufacture of equipment for the chemical industry to avoid corrosion failure. Plasma-arc hot wire surfacing is also shown to be capable of depositing single or double clad layers to meet the highest safety requirements and could be applied to nuclear power plants for the special manufacture of wear resistant parts and for protection of equipment subject to a variety of corrosive environments. (U.K.)

  15. Investigation of Hot Rolling Influence on the Explosive-Welded Clad Plate

    Directory of Open Access Journals (Sweden)

    Guanghui ZHAO

    2016-11-01

    Full Text Available The microstructure, the shear strength and tensile strength of stainless steel explosive-welded clad plate at different rolling reduction were studied. The mechanical properties of the explosive-welded and explosive-rolled clad plates were experimentally measured. Simultaneously, the microstructures of the clad plate were investigated by the Ultra deep microscope and the tensile fracture surface were observed by the scan electron microscope (SEM. It was observed that the tensile strength has been increased considerably, whereas the elongation percentage has been reduced with the increase of hot rolling reduction. In the tensile shear test, the bond strength is higher than the strength of the ferritic stainless steel layer and meets the relevant known standard criterion. Microstructural evaluations showed that the grain of the stainless steel and steel refined with the increase of thickness reduction. Examination of the tensile fracture surfaces reveal that, after hot rolling, the fracture in the low alloy steel and ferritic stainless steel clad plates is of the ductile type.DOI: http://dx.doi.org/10.5755/j01.ms.22.4.12409

  16. A preliminary study of laser cladding of AISI 316 stainless steel using preplaced NiTi wire

    International Nuclear Information System (INIS)

    Cheng, F.T.; Lo, K.H.; Man, H.C.

    2004-01-01

    NiTi wire of diameter 1 mm was preplaced on AISI 316 stainless steel samples by using a binder. Melting of the NiTi wire to form a clad track on the steel substrate was achieved by means of a high-power CW Nd:YAG laser using different processing parameters. The geometry and microstructure of the clad deposit were studied by optical microscopy and scanning electron microscopy (SEM), respectively. The hardness and compositional profiles along the depth of the deposit were acquired by microhardness testing and energy-dispersive spectroscopy (EDS), respectively. The elastic behavior of the deposit was analyzed using nanoindentation, and compared with that of the NiTi wire. The dilution of the NiTi clad by the substrate material beneath was substantial in single clad tracks, but could be successively reduced in multiple clad layers. A strong fusion bonding with tough interface could be obtained as evidenced by the integrity of Vickers indentations in the interfacial region. In comparison with the NiTi cladding on AISI 316 using the tungsten inert gas (TIG) process, the laser process was capable of producing a much less defective cladding with a more homogeneous microstructure, which is an essential cladding quality with respect to cavitation erosion and corrosion resistance. Thus, the present preliminary study shows that laser cladding using preplaced wire is a feasible method to obtain a thick and homogeneous NiTi-based alloy layer on AISI 316 stainless steel substrate

  17. Prevention of microcracking by REM addition to alloy 690 filler metal in laser clad welds

    International Nuclear Information System (INIS)

    Okauchi, Hironori; Saida, Kazuyoshi; Nishimoto, Kazutoshi

    2011-01-01

    Effect of REM addition to alloy 690 filler metal on microcracking prevention was verified in laser clad welding. Laser clad welding on alloy 132 weld metal or type 316L stainless steel was conducted using the five different filler metals of alloy 690 varying the La content. Ductility-dip crack occurred in laser clad welding when La-free alloy 690 filler metal was applied. Solidification and liquation cracks occurred contrarily in the laser cladding weld metal when the 0.07mass%La containing filler metal was applied. In case of laser clad welding on alloy 132 weld metal and type 316L stainless steel, the ductility-dip cracking susceptibility decreased, and solidification/liquation cracking susceptibilities increased with increasing the La content in the weld metal. The relation among the microcracking susceptibility, the (P+S) and La contents in every weld pass of the laser clad welding was investigated. Ductility-dip cracks occurred in the compositional range (atomic ratio) of La/(P+S) 0.99(on alloy 132 weld metal), >0.90 (on type 316L stainless steel), while any cracks did not occur at La/(P+S) being between 0.21-0.99 (on alloy 132 weld metal) 0.10-0.90 (on type 316L stainless steel). Laser clad welding test on type 316L stainless steel using alloy 690 filler metal containing the optimum La content verified that any microcracks did not occurred in the laser clad welding metal. (author)

  18. Efficient single-mode operation of a cladding-pumped ytterbium-doped helical-core fiber laser.

    Science.gov (United States)

    Wang, P; Cooper, L J; Sahu, J K; Clarkson, W A

    2006-01-15

    A novel approach to achieving robust single-spatial-mode operation of cladding-pumped fiber lasers with multimode cores is reported. The approach is based on the use of a fiber geometry in which the core has a helical trajectory within the inner cladding to suppress laser oscillation on higher-order modes. In a preliminary proof-of-principle study, efficient single-mode operation of a cladding-pumped ytterbium-doped helical-core fiber laser with a 30 microm diameter core and a numerical aperture of 0.087 has been demonstrated. The laser yielded 60.4 W of output at 1043 nm in a beam with M2 clad fiber lasers.

  19. Effects of Synchronous Rolling on Microstructure, Hardness, and Wear Resistance of Laser Multilayer Cladding

    Science.gov (United States)

    Zhao, W.; Zha, G. C.; Xi, M. Z.; Gao, S. Y.

    2018-03-01

    A synchronous rolling method was proposed to assist laser multilayer cladding, and the effects of this method on microstructure, microhardness, and wear resistance were studied. Results show that the microstructure and mechanical properties of the traditional cladding layer exhibit periodic inhomogeneity. Synchronous rolling breaks the columnar dendrite crystals to improve the uniformity of the organization, and the residual plastic energy promotes the precipitation of strengthening phases, as CrB, M7C3, etc. The hardness and wear resistance of the extruded cladding layer increase significantly because of the grain refinement, formation of dislocations, and dispersion strengthening. These positive significances of synchronous rolling provide a new direction for laser cladding technology.

  20. Femtosecond-laser inscribed double-cladding waveguides in Nd:YAG crystal: a promising prototype for integrated lasers.

    Science.gov (United States)

    Liu, Hongliang; Chen, Feng; Vázquez de Aldana, Javier R; Jaque, D

    2013-09-01

    We report on the design and implementation of a prototype of optical waveguides fabricated in Nd:YAG crystals by using femtosecond-laser irradiation. In this prototype, two concentric tubular structures with nearly circular cross sections of different diameters have been inscribed in the Nd:YAG crystals, generating double-cladding waveguides. Under 808 nm optical pumping, waveguide lasers have been realized in the double-cladding structures. Compared with single-cladding waveguides, the concentric tubular structures, benefiting from the large pump area of the outermost cladding, possess both superior laser performance and nearly single-mode beam profile in the inner cladding. Double-cladding waveguides of the same size were fabricated and coated by a thin optical film, and a maximum output power of 384 mW and a slope efficiency of 46.1% were obtained. Since the large diameters of the outer claddings are comparable with those of the optical fibers, this prototype paves a way to construct an integrated single-mode laser system with a direct fiber-waveguide configuration.

  1. Chemical dissolution of spent fuel and cladding using complexed fluoride species

    International Nuclear Information System (INIS)

    Rance, P.J.W.; Freeman, G.A.; Mishin, V.; Issoupov, V.

    2001-01-01

    The dissolution of LWR fuel cladding using two fluoride ion donors, HBF 4 and K 2 ZrF 6 , in combination with nitric acid has been investigated as a potential reprocessing head-end process suitable for chemical decladding and fuel dissolution in a single process step. Maximum zirconium concentrations in the order of 0,75 to 1 molar have been achieved and dissolution found to continue to low F:Zr ratios albeit at ever decreasing rates. Dissolution rates of un-oxidised zirconium based fuel claddings are fast, whereas oxidised materials exhibit an induction period prior to dissolution. Data is presented relating to the rates of dissolution of cladding and UO 2 fuels under various conditions. (author)

  2. Development of laser cladding technology for maintenance of pipe wall thinning

    International Nuclear Information System (INIS)

    Terada, Takaya; Nishimura, Akihiko; Oka, Kiyoshi

    2011-01-01

    We are developing the laser welding and cladding device for the maintenance of heat exchanger pipes. In the case of flow accelerated corrosion where pipe wall thinning occurred after a long time operation, laser cladding is mostly expected. A laser processing head was proposed in order to access the pipe wall. A composite-type optical fiber scope was used for real time observation and laser processing. An air-cooled compact fiber laser was used for spot heating. We present the concept of the laser cladding device which have the following features: 1) Wire feeding modules, 2) Module capable of laser irradiation in the vertical heat exchanger pipe, 3) Assist gas injection module. (author)

  3. Zinc Cadmium Selenide Cladded Quantum Dot Based Electroluminescent and Nonvolatile Memory Devices

    Science.gov (United States)

    Al-Amody, Fuad H.

    This dissertation presents electroluminescent (EL) and nonvolatile memory devices fabricated using pseudomorphic ZnCdSe-based cladded quantum dots (QDs). These dots were grown using our own in-school built novel reactor. The EL device was fabricated on a substrate of ITO (indium tin oxide) coated glass with the quantum dots sandwiched between anode and cathode contacts with a small barrier layer on top of the QDs. The importance of these cladded dots is to increase the quantum yield of device. This device is unique as they utilize quantum dots that are pseudomorphic (nearly lattice-matched core and the shell of the dot). In the case of floating quantum dot gate nonvolatile memory, cladded ZnCdSe quantum dots are deposited on single crystalline gate insulator (ZnMgS/ZnMgSe), which is grown using metal-organic chemical vapor deposition (MOCVD). The control gate dielectric layer of the nonvolatile memory is Si3N4 or SiO2 and is grown using plasma enhanced chemical vapor deposition (PECVD). The cladded dots are grown using an improved methodology of photo-assisted microwave plasma metal-organic chemical vapor deposition (PMP-MOCVD) enhanced reactor. The cladding composition of the core and shell of the dots was engineered by the help of ultraviolet light which changed the incorporation of zinc (and hence composition of ZnCdSe). This makes ZnxCd1--xSe-ZnyCd1--y Se QDs to have a low composition of zinc in the core than the cladding (x

  4. Fuel chemistry and pellet-clad interaction related to high burnup fuel. Proceedings of the technical committee

    International Nuclear Information System (INIS)

    2000-10-01

    The purpose of the meeting was to review new developments in clad failures. Major findings regarding the causes of clad failures are presented in this publication, with the main topics being fuel chemistry and fission product behaviour, swelling and pellet-cladding mechanical interaction, cladding failure mechanism at high burnup, thermal properties and fuel behaviour in off-normal conditions. This publication contains 17 individual presentations delivered at the meeting; each of them was indexed separately

  5. Laser cladding of austenitic stainless steel using NiTi strips for resisting cavitation erosion

    International Nuclear Information System (INIS)

    Chiu, K.Y.; Cheng, F.T.; Man, H.C.

    2005-01-01

    Being part of a larger project on using different forms of nickel titanium (NiTi) in the surface modification of stainless steel for enhancing cavitation erosion resistance, the present study employs NiTi strips as the cladding material. Our previous study shows that laser surfacing using NiTi powder can significantly increase the cavitation erosion resistance of AISI 316 L stainless steel [K.Y. Chiu, F.T. Cheng, H.C. Man, Mater. Sci. Eng. A 392 (2005) 348-358]. However, from an engineering point of view, NiTi strips are more attractive than powder because NiTi powder is very expensive due to high production cost. In the present study, NiTi strips were preplaced on AISI 316 L samples and remelted using a high-power CW Nd:YAG laser to form a clad layer. To lower the dilution due to the substrate material, samples doubly clad with NiTi were prepared. The volume dilution ratio in the singly clad sample was high, being in the range of 13-30% depending on the processing parameters, while that of the doubly clad sample was reduced to below 10%. Analysis by scanning electron microscopy (SEM), energy-dispersive spectroscopy (EDS) and X-ray diffractometry (XRD) reveals that the clad layer is composed of a NiTi B2 based matrix together with fine precipitates of a tetragonal structure. Vickers indentation shows a tough cladding/substrate interface. The microhardness of the clad layer is increased from 200 HV of the substrate to about 750 HV due to the dissolution of elements like Fe, Cr and N in the matrix. Nanoindentation tests record a recovery ratio near to that of bulk NiTi, a result attributable to a relatively low dilution. The cavitation erosion resistance of the doubly clad samples is higher than that of 316-NiTi-powder (samples laser-surfaced with NiTi powder) and approaches that of NiTi plate. The high erosion resistance is attributed to a high hardness, high indentation recovery ratio and the absence of cracks or pores

  6. A preliminary study of cladding steel with NiTi by microwave-assisted brazing

    International Nuclear Information System (INIS)

    Chiu, K.Y.; Cheng, F.T.; Man, H.C.

    2005-01-01

    Nickel titanium (NiTi) plate of 1.2 mm thickness was successfully clad on AISI 316L stainless steel substrate by a microwave-assisted brazing process. Brazing was conducted in a multimode microwave oven in air using a copper-based brazing material in tape form. The brazing material was melted in a few minutes by microwave-induced plasma initiated by conducting wires surrounding the brazing assembly. Metallographic study by scanning-electron microscopy (SEM) and compositional analysis by energy-dispersive spectroscopy (EDS) of the brazed joint revealed metallurgical bonding formed via inter-diffusion between the brazing filler and the adjacent materials. A shear bonding strength in the range of 100-150 MPa was recorded in shear tests of the brazed joint. SEM and X-ray diffractometry (XRD) analysis for the surface of as-received NiTi plate and NiTi cladding showed similar microstructure and phase composition. Nanoindentation tests also indicated that the superelastic properties of NiTi were essentially retained. The cavitation erosion resistance of the NiTi cladding was essentially the same as that of as-received NiTi plate, and higher than that obtained in laser or TIG (tungsten-inert gas) surfacing. The high resistance could be attributed to avoidance of dilution and defect formation in the NiTi clad since the cladding did not undergo melting and solidification in the brazing process. Electrochemical tests also recorded similar corrosion resistance in both as-received NiTi and NiTi cladding. Thus, the present study indicates that microwave-assisted brazing is a simple, economical, and feasible process for cladding NiTi on 316L stainless steel for enhancing cavitation erosion resistance

  7. 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

  8. The VULKIN code used for evaluation of the cladding tube's performance

    International Nuclear Information System (INIS)

    Marbach, G.

    1979-01-01

    Full text: 1 - Introduction. The French approach for fast subassembly project is to analyse each component part of the subassembly and each basic phenomenon to estimate the total behaviour. The VULKIN code describes the mechanical behaviour of a clad alone. A cladding damage parameter is calculated from the observed deformations. When it is greater than a fixed value we consider that the rupture probability is not negligible. But this function is not the only limit for the irradiation project. Other limits are bound to other problems: no fuel melting bundle, interaction behaviour. 2 - VULKIN code - Presentation. The VULKIN code gives the evolution of stresses and strains distribution in the thickness of the clad with the hypothesis of revolution symmetry. This program takes into account temperature dilatation and radial thermal gradient, fission gas pressure and steel swelling due to neutron flux. The fuel clad mechanical interaction is not described by this model. Experimental results show that its influence is negligible for the most unusual subassemblies but, if it is necessary, a special calculation is obtained using a specific code like TUREN, described in another paper. This model does not consider the stresses and strains resulting from interaction between bundle and wrapper. Another model describes the bundle behaviour and determines diametral deformation limit from the subassembly geometrical characteristics. The clad is considered as an elasto-plastic element. Plastic flows instantaneous, thermal creep or irradiation creep are determined at each time. The data of this code are the geometry, the irradiation parameters (temperature, dose), the fission gas pressure evolution, the swelling law and the experimental relations for thermal and irradiation creep. The mechanical resolution is classical: the clad is divided into concentric rings. At each time the equations resulting from the equilibrium of strengths and compatibility of displacements are resolved

  9. Effect of Temperature and Sheet Temper on Isothermal Solidification Kinetics in Clad Aluminum Brazing Sheet

    Science.gov (United States)

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

    2016-09-01

    Isothermal solidification (IS) is a phenomenon observed in clad aluminum brazing sheets, wherein the amount of liquid clad metal is reduced by penetration of the liquid clad into the core. The objective of the current investigation is to quantify the rate of IS through the use of a previously derived parameter, the Interface Rate Constant (IRC). The effect of peak temperature and initial sheet temper on IS kinetics were investigated. The results demonstrated that IS is due to the diffusion of silicon (Si) from the liquid clad layer into the solid core. Reduced amounts of liquid clad at long liquid duration times, a roughened sheet surface, and differences in resolidified clad layer morphology between sheet tempers were observed. Increased IS kinetics were predicted at higher temperatures by an IRC model as well as by experimentally determined IRC values; however, the magnitudes of these values are not in good agreement due to deficiencies in the model when applied to alloys. IS kinetics were found to be higher for sheets in the fully annealed condition when compared with work-hardened sheets, due to the influence of core grain boundaries providing high diffusivity pathways for Si diffusion, resulting in more rapid liquid clad penetration.

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

    International Nuclear Information System (INIS)

    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

  11. Effect of annealing temperature on the mechanical properties of Zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Beauregard, R.J.; Clevinger, G.S.; Murty, K.L.

    1977-01-01

    The mechanical properties of Zircaloy cladding materials are sensitive to those fabrication variables which have an effect on the preferred crystallographic orientation or texture of the finished tube. The effect of one such variable, the final annealing temperature, on various mechanical properties is examined using tube reduced Zircaloy-4 fuel rod cladding annealed at temperatures from 905F to 1060F. This temperature range provides cladding with varying degrees of recrystallization including full recrystallization. The burst strength of the cladding at 650F decreased with the annealing temperature reaching a saturation value at approximately 1000F. The total circumferential elongation increased with the annealing temperature reaching a maximum at approximately 1000F and decreasing at higher temperatures. Hoop creep characteristics of Zircaloy cladding were studied as a function of the annealing temperature using closed-end internal pressurization tests at 750F and hoop stresses of 10, 15, 20 and 25 ksi. The effect of annealing temperature on the room temperature mechanical anisotropy parameters, R and P, was studied. The R-parameter was essentially independent of the annealing temperature while the P-parameter increased with annealing temperature. The mechanical anisotropy parameters were also studied as a function of the test temperature from ambient to approximately 800F using continuously monitored high precision extensometry. (Auth.)

  12. Laser Powder Cladding of Ti-6Al-4V α/β Alloy

    Science.gov (United States)

    Al-Sayed Ali, Samar Reda; Hussein, Abdel Hamid Ahmed; Nofal, Adel Abdel Menam Saleh; Elgazzar, Haytham Abdelrafea; Sabour, Hassan Abdel

    2017-01-01

    Laser cladding process was performed on a commercial Ti-6Al-4V (α + β) titanium alloy by means of tungsten carbide-nickel based alloy powder blend. Nd:YAG laser with a 2.2-KW continuous wave was used with coaxial jet nozzle coupled with a standard powder feeding system. Four-track deposition of a blended powder consisting of 60 wt % tungsten carbide (WC) and 40 wt % NiCrBSi was successfully made on the alloy. The high content of the hard WC particles is intended to enhance the abrasion resistance of the titanium alloy. The goal was to create a uniform distribution of hard WC particles that is crack-free and nonporous to enhance the wear resistance of such alloy. This was achieved by changing the laser cladding parameters to reach the optimum conditions for favorable mechanical properties. The laser cladding samples were subjected to thorough microstructure examinations, microhardness and abrasion tests. Phase identification was obtained by X-ray diffraction (XRD). The obtained results revealed that the best clad layers were achieved at a specific heat input value of 59.5 J·mm−2. An increase by more than three folds in the microhardness values of the clad layers was achieved and the wear resistance was improved by values reaching 400 times. PMID:29036935

  13. Laser Powder Cladding of Ti-6Al-4V α/β Alloy.

    Science.gov (United States)

    Al-Sayed Ali, Samar Reda; Hussein, Abdel Hamid Ahmed; Nofal, Adel Abdel Menam Saleh; Hasseb Elnaby, Salah Elden Ibrahim; Elgazzar, Haytham Abdelrafea; Sabour, Hassan Abdel

    2017-10-15

    Laser cladding process was performed on a commercial Ti-6Al-4V (α + β) titanium alloy by means of tungsten carbide-nickel based alloy powder blend. Nd:YAG laser with a 2.2-KW continuous wave was used with coaxial jet nozzle coupled with a standard powder feeding system. Four-track deposition of a blended powder consisting of 60 wt % tungsten carbide (WC) and 40 wt % NiCrBSi was successfully made on the alloy. The high content of the hard WC particles is intended to enhance the abrasion resistance of the titanium alloy. The goal was to create a uniform distribution of hard WC particles that is crack-free and nonporous to enhance the wear resistance of such alloy. This was achieved by changing the laser cladding parameters to reach the optimum conditions for favorable mechanical properties. The laser cladding samples were subjected to thorough microstructure examinations, microhardness and abrasion tests. Phase identification was obtained by X-ray diffraction (XRD). The obtained results revealed that the best clad layers were achieved at a specific heat input value of 59.5 J·mm -2 . An increase by more than three folds in the microhardness values of the clad layers was achieved and the wear resistance was improved by values reaching 400 times.

  14. Thermal and mechanical behavior of APWR-claddings under critical heat flux conditions

    International Nuclear Information System (INIS)

    Diegele, E.; Rust, K.

    1986-10-01

    Helical grid spacers, such as three or six helical fins as integral part of the claddings, are regarded as a more convenient design for the very tight lattice of an advanced pressurized water reactor (APWR) than grid spacers usually used. Furthermore, it is expected that this spacer design allows an increased safety margin against the critical heat flux (CHF), the knowledge of which is important for design, licensing, and operation of water cooled reactors. To address the distribution of the heat flux density at the outer circumference of the cladding geometry under investigation, the temperature fields in claddings without as well with fins were calculated taking into consideration nuclear and electrically heated rods. Besides the thermal behavior of the claddings, the magnitude and distribution of thermal stresses were determined additionally. A locally increased surface heat flux up to about 40 percent was calculated for the fin bases of nuclear as well as indirect electrically heated claddings with six such helical fins. For all investigated cases, the VON MISES stresses are clearly lower than 200 MPa, implying that no plastic deformations are to be expected. The aim of this theoretical analysis is to allow a qualitative assessment of the finned tube conception and to support experimental investigations concerning the critical heat flux. (orig.) [de

  15. 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

  16. Out-of-pile experiments of fuel-cladding chemical interaction, (2)

    International Nuclear Information System (INIS)

    Konashi, Kenji; Yato, Tadao; Kaneko, Hiromitsu; Honda, Yutaka

    1980-01-01

    Cesium seems to be one of the most important fission products in the fuel-cladding chemical interaction of fuel pins for LMFBRs. However the FCCI under irradiation cannot always be explained by considering only cesium-oxygen system as the corrosive, since attack does not occur in the cesium-oxygen system unless oxygen potential is sufficiently high. Cesium-tellurium-oxygen system has been proposed to account for heavy cladding attack which was sometimes found in hypostoichiometric mixed oxide fuel pins. In this paper, the experiment on the reaction of liquid tellurium with stainless steel is reported. The type 316 stainless steel claddings for Monju type fuel pins were used as the test specimens. Tellurium was contained into the cladding tubes with end plugs. The temperature dependence of the attack by tellurium was examined in the range from 450 to 900 deg C for 30 min, and the heating time dependence was examined from 5 min to 200 hr at 725 deg C. An infrared lamp furnace was used for the experiment within 7 hr, and a resistance furnace for longer experiment. The character of corrosion was matrix attack, and the reaction products on the stainless steel surfaces consisted of chrome rich inner phase and iron and nickel rich outer phase. The results are reported. (Kako, I.)

  17. Cladding tube of fuel rod for a BWR type reactor

    International Nuclear Information System (INIS)

    Nakayama, Hitoshi; Fujie, Kunio; Kuwahara, Heikichi; Hirai, Tadamasa; Kakizaki, Kimio.

    1976-01-01

    Object: To form a cladding tube wall with tunnels in communication with the exterior through a number of small-diameter openings to rapidly disperse a large quantity of heat thereby providing high density of the fuel rod. Structure: Tunnels adjacent to each other are provided under the skin in contact with cooling liquid of a cladding tube, and a number of openings through which said tunnels and the periphery of the cladding tube are placed in communication are formed, said openings each having its section smaller than that of said tunnel. With this arrangement, the cooling water entered the tunnel through some of small diameter openings absorbs heat of the fuel rod to be vaporized, which is flown out into the cooling water through the other small diameter openings and formed into vapor bubbles which move up for release of heat. (Taniai, N.)

  18. Finite element modeling of pellet-clad mechanical interaction with ABAQUS

    International Nuclear Information System (INIS)

    Cheon, C. S.; Lee, B. H.; Koo, Y. H.; Oh, J. Y.; Son, D. S.

    2002-01-01

    Pellet-clad mechanical interaction (PCMI) was modelled by an axisymmetric finite element method. Thermomechanical models of pellet and clad materials and a contact model for their interaction have been implemented in addition to the application of appropriate boundary conditions so that the FE model was configured. Temperature and displacement were evaluated through a coupled analysis using a general purposed FE code, ABAQUS. Also, a batch program has been developed to efficiently deal with a series of jobs such as making an interface with a fuel performance code, the generation of an input deck for ABAQUS code and its execution, and an interpretation of the output. Under various conditions, results from the present FE model were analyzed. Preliminary verification was conducted by comparing the clad elongation measured during an in-pile PCMI experiment with that calculated by means of the developed FE model

  19. Development of a laser multi-layer cladding technology for damage mitigation of fuel spacers in Hanaro reactor

    International Nuclear Information System (INIS)

    Kim, J. S.; Lee, D. H.; Hwang, S. S.; Suh, J. H.

    2002-01-01

    A laser multi-layer cladding technology was developed to mitigate the fretting wear damages occurred at fuel spacers in Hanaro reactor. The detailed experimental results are as follows. 1) Analyses of fretting wear damages and fabrication process of fuel spacers 2) Development and analysis of spherical Al 6061 T-6 alloy powders for the laser cladding 3) Analysis of parameter effects on laser cladding process for clad bids, and optimization of laser cladding process 4) Analysis on the changes of cladding layers due to overlapping factor change 5) Microstructural observation and phase analysis 6) Characterization of materials properties (hardness and wear tests) 7) Manufacture of prototype fuel spacers 8) Development of a vision system and revision of its related softwares

  20. 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.

  1. Failure probabilities of SiC clad fuel during a LOCA in public acceptable simple SMR (PASS)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Ho Sik, E-mail: hskim25@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

    2015-10-15

    Highlights: • Graceful operating conditions of SMRs markedly lower SiC cladding stress. • Steady-state fracture probabilities of SiC cladding is below 10{sup −7} in SMRs. • PASS demonstrates fuel coolability (T < 1300 °C) with sole radiation in LOCA. • SiC cladding failure probabilities of PASS are ∼10{sup −2} in LOCA. • Cold gas gap pressure controls SiC cladding tensile stress level in LOCA. - Abstract: Structural integrity of SiC clad fuels in reference Small Modular Reactors (SMRs) (NuScale, SMART, IRIS) and a commercial pressurized water reactor (PWR) are assessed with a multi-layered SiC cladding structural analysis code. Featured with low fuel pin power and temperature, SMRs demonstrate markedly reduced incore-residence fracture probabilities below ∼10{sup −7}, compared to those of commercial PWRs ∼10{sup −6}–10{sup −1}. This demonstrates that SMRs can serve as a near-term deployment fit to SiC cladding with a sound management of its statistical brittle fracture. We proposed a novel SMR named Public Acceptable Simple SMR (PASS), which is featured with 14 × 14 assemblies of SiC clad fuels arranged in a square ring layout. PASS aims to rely on radiative cooling of fuel rods during a loss of coolant accident (LOCA) by fully leveraging high temperature tolerance of SiC cladding. An overarching assessment of SiC clad fuel performance in PASS was conducted with a combined methodology—(1) FRAPCON-SiC for steady-state performance analysis of PASS fuel rods, (2) computational fluid dynamics code FLUENT for radiative cooling rate of fuel rods during a LOCA, and (3) multi-layered SiC cladding structural analysis code with previously developed SiC recession correlations under steam environments for both steady-state and LOCA. The results show that PASS simultaneously maintains desirable fuel cooling rate with the sole radiation and sound structural integrity of fuel rods for over 36 days of a LOCA without water supply. The stress level of

  2. FUMAC-a new model for light water reactor fuel relocation and pellet-cladding interaction

    International Nuclear Information System (INIS)

    Walton, L.A.; Matheson, J.E.

    1984-01-01

    An improved approach to the mechanical modeling of fuel rod performance is presented. Previous computer modeling has centered around a unified finite element approach with both fuel pellets and cladding being represented by ring elements. The fuel mechanical analysis code (FUMAC) departs from these approaches in two areas. The pellet model is an empirically based deterministic algorithm, while the cladding model uses both plane stress and plane strain finite elements. The work describes a semiempirical fuel cracking and fragment relocation model, which is burnup and power-level dependent. The interaction of the pellet with the cladding is treated classically. The resulting thick cylinder stresses are used in conjunction with an orthotropic creep model to predict cladding ridging. The resulting ridging compares well with experimental data for both steady-state and transient operating conditions. Future work planned includes the integration of the finite element cladding model with the pellet model and refinement of the pellet relocation and thermal models. Transient performance predictions will be emphasized

  3. An analytical model to predict and minimize the residual stress of laser cladding process

    Science.gov (United States)

    Tamanna, N.; Crouch, R.; Kabir, I. R.; Naher, S.

    2018-02-01

    Laser cladding is one of the advanced thermal techniques used to repair or modify the surface properties of high-value components such as tools, military and aerospace parts. Unfortunately, tensile residual stresses generate in the thermally treated area of this process. This work focuses on to investigate the key factors for the formation of tensile residual stress and how to minimize it in the clad when using dissimilar substrate and clad materials. To predict the tensile residual stress, a one-dimensional analytical model has been adopted. Four cladding materials (Al2O3, TiC, TiO2, ZrO2) on the H13 tool steel substrate and a range of preheating temperatures of the substrate, from 300 to 1200 K, have been investigated. Thermal strain and Young's modulus are found to be the key factors of formation of tensile residual stresses. Additionally, it is found that using a preheating temperature of the substrate immediately before laser cladding showed the reduction of residual stress.

  4. 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...

  5. Chemical inhomogeneity populations in various zircaloy claddings and their association with SCC and corrosion resistance

    International Nuclear Information System (INIS)

    Tasooji, A.; Miller, A.K.; Cheung, T.Y.; Brooks, M.; Santucci, J.

    1987-01-01

    A technique has been developed that permits detection and characterization of sparsely distributed chemical inhomogeneities in Zircaloy. These inhomogeneities have previously been observed at the origins of iodine stress-corrosion cracks but are not detectable by, for example, simple scanning electron microscopy (SEM) examination. The technique uses radioactive iodine to ''label'' the chemical inhomogeneities, autoradiography to detect their locations, and SEM and energy-dispersive X-ray analysis (EDAX) to further characterize them. Large areas of surface have been surveyed and statistically meaningful populations of chemical inhomogeneities measured for five different lots of Zircaloy cladding. Inner surfaces and cladding cross-sectional surfaces have been studied. There are clear differences in chemical inhomogeneity size distribution and composition between the various claddings. For three of the claddings characterized in this work, the previously measured stress-corrosion cracking (SCC) threshold stresses correlate well (inversely) with the new data on their average chemical inhomogeneity sizes. Of special interest is the fact that the most SCC-resistant cladding contains far fewer iron-bearing inhomogeneities than the other claddings

  6. Application of laser cladding method to small-diameter stainless steel pipes in actual nuclear plant

    International Nuclear Information System (INIS)

    Atago, Y.; Yamadera, M.; Tsuji, H.; Shiraiwa, T.; Kanno, M.

    1995-01-01

    Recently, to prevent stress corrosion cracking (SCC) the material of stainless steel (Type 304), a laser cladding method which produces a highly corrosion-resisting coating (cladding) to be formed on the surface of the material was developed. This is applicable to a long distance and narrow space, because of the good accessibility of the YAG (Yttrium-Aluminum Garnet) laser beam that can be transmitted through an optical fiber. In this method, a paste mixed metallic powder and heating resistive organic solvent is firstly placed on the inner surface of a small pipe and then a YAG laser beam transmitted through an optical fiber is irradiated to the paste, which will be melted and formed a clad subsequently, which is excellent in corrosion resistance. Finally, it can be achieved further resistance against the SCC due to the clad layer formed thus on the surface of the material. Recently, this Laser Cladding method was practically and successfully applied to the actual BWR Nuclear Power Plant in Japan. This report introduces the laser cladding technique, the equipments developed for practical application in the field

  7. Design of absorber assemblies with intentional pellet-cladding mechanical interaction

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Birney, K.R.; Pitner, A.L.; Basmajian, J.A.

    1980-04-01

    A number of improvements in absorber assembly performance characteristics can be achieved through implementation of absorber cladding mechanical interaction (ACMI). Benefits include lower operating temperatures, less potential for material relocation, longer lifetime, and increased reactivity worth. Analyses indicate that substantial cladding strains may be attainable without significant risk of breach. However, actual in-reactor testing of ACMI in absorber elements will be required before design criteria can be revised to accept ACMI

  8. 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

  9. 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.

  10. A Eutectic Melting Study of Double Wall Cladding Tubes of FeCrAl and Zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Woojin; Son, Seongmin; Lee, You Ho; Lee, Jeong Ik; Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Jeong, Eun [Kyunghee University, Yongin (Korea, Republic of)

    2015-10-15

    The eutectic melting behavior of FeCrAl/Zircaloy-4 double wall cladding tubes was investigated by annealing at various temperatures ranging from 900 .deg. C to 1300 .deg. C. It was found that significant eutectic melting occurred after annealing at temperatures equal to or higher than 1150 .deg. C. It means that an additional diffusion barrier layer is necessary to limit the eutectic melting between FeCrAl and Zircaloy-4 alloy cladding tubes. Coating of FeCrAl layers on the Zr alloy cladding tube is being investigated for the development of accident tolerant fuel by exploiting of both the oxidation resistance of FeCrAl alloys and the neutronic advantages of Zr alloys. Coating of FeCrAl alloys on Zr alloy cladding tubes can be performed by various techniques including thermal spray, laser cladding, and co-extrusion. Son et al. also reported the fabrication of FeCrAl/Zr ally double wall cladding by the shrink fit method. For the double layered cladding tubes, the thermal expansion mismatch between the dissimilar materials, severe deformation or mechanical failure due to the evolution of thermal stresses can occur when there is a thermal cycling. In addition to the thermal stress problems, chemical compatibilities between the two different alloys should be investigated in order to check the stability and thermal margin of the double wall cladding at a high temperature. Generally, it is considered that Zr alloy cladding will maintain its mechanical integrity up to 1204 .deg. C (2200 .deg. F) to satisfy the acceptance criteria for emergency core cooling systems.

  11. Tangential neutral-beam-driven instabilities in the princeton beta experiment

    OpenAIRE

    Heidbrink, WW; Bol, K; Buchenauer, D; Fonck, R; Gammel, G; Ida, K; Kaita, R; Kaye, S; Kugel, H; LeBlanc, B; Morris, W; Okabayashi, M; Powell, E; Sesnic, S; Takahashi, H

    1986-01-01

    During tangential neutral-beam injection into the PBX tokamak, bursts of two types of instabilities are observed. One instability occurs in the frequency range 120-210 kHz and the other oscillates predominantly near the frequency of bulk plasma rotation (20-30 kHz). Both instabilities correlate with drops in neutron emission and bursts in charge-exchange neutral flux, indicating that beam ions are removed from the center of the plasma by the instabilities. The central losses are comparable to...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-03-02

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

  13. 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

  14. 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

  15. In-cell facility for performing mechanical-property tests on irradiated cladding

    International Nuclear Information System (INIS)

    Yaggee, F.L.; Haglund, R.C.; Mattas, R.F.

    1978-11-01

    A new facility was developed for testing cladding sections of LWR fuel rods. This facility and the accompanying test procedures have improved the level of in-cell mechanical-testing capabilities, making them comparable to existing capabilities for unirradiated cladding. The new facility is currently being used to study the susceptibility of irradiated Zircaloy cladding from LWR fuel rods to iodine stress-corrosion cracking. Preliminary testing results indicate a systematic effect of temperature, stress and irradiation on the susceptibility of annealed and stress-relieved Zircaloy-2. Experimental data obtained to date are being used to develop a stress-corrosion cracking model for LWR fuel rod failure. SEM examination of the undisturbed fracture surface of specimens that failed by pinhole leakage provides useful information on crack propagation and morphology

  16. Effects of irradiation on strength and toughness of commercial LWR vessel cladding

    International Nuclear Information System (INIS)

    Haggag, F.M.; Corwin, W.R.; Alexander, D.J.; Nanstad, R.K.

    1987-01-01

    The potential for stainless steel cladding to improve the fracture behavior of an operating nuclear reactor pressure vessel, particularly during certain overcooling transients, may depend greatly on the properties of the irradiated cladding. Therefore, weld overlay cladding irradiated at temperatures and to fluences relevant to power reactor operation was examined. The cladding was applied to a pressure vessel steel plate by the three-wire series-arc commercial method. Cladding was applied in three layers to provide adequate thickness for the fabrication of test specimens. The three-wire series-arc procedure, developed by Combustion Engineering, Inc., Chattanooga, Tennessee, produced a highly controlled weld chemistry, microstructure, and fracture properties in all three layers of the weld. Charpy V-notch and tensile specimens were irradiated at 288 0 C to fluence levels of 2 and 5 x 10 19 neutrons/cm 2 (>1 MeV). Postirradiation testing results show that, in the test temperature range from -125 to 288 0 C, the yield strength increased by 8 to 30%, ductility insignificantly increased, while there was almost no change in ultimate tensile strength. All cladding exhibited ductile-to-brittle transition behavior during Charpy impact testing, due to the dominance of delta-ferrite failures at low temperatures. On the upper shelf, energy was reduced, due to irradiation exposure, 15 and 20%, while the lateral expansion was reduced 43 and 41%, at 2 and 5 x 10 19 neutrons/cm 2 (>1 MeV), respectively. In addition, radiation damage resulted in 13 and 28 0 C shifts of the Charpy impact transition temperature at the 41-J level for the low and high fluences, respectively

  17. Modeling of Heat Transfer and Fluid Flow in the Laser Multilayered Cladding Process

    Science.gov (United States)

    Kong, Fanrong; Kovacevic, Radovan

    2010-12-01

    The current work examines the heat-and-mass transfer process in the laser multilayered cladding of H13 tool steel powder by numerical modeling and experimental validation. A multiphase transient model is developed to investigate the evolution of the temperature field and flow velocity of the liquid phase in the molten pool. The solid region of the substrate and solidified clad, the liquid region of the melted clad material, and the gas region of the surrounding air are included. In this model, a level-set method is used to track the free surface motion of the molten pool with the powder material feeding and scanning of the laser beam. An enthalpy-porosity approach is applied to deal with the solidification and melting that occurs in the cladding process. Moreover, the laser heat input and heat losses from the forced convection and heat radiation that occurs on the top surface of the deposited layer are incorporated into the source term of the governing equations. The effects of the laser power, scanning speed, and powder-feed rate on the dilution and height of the multilayered clad are investigated based on the numerical model and experimental measurements. The results show that an increase of the laser power and powder feed rate, or a reduction of the scanning speed, can increase the clad height and directly influence the remelted depth of each layer of deposition. The numerical results have a qualitative agreement with the experimental measurements.

  18. Microstructure and phase evolution in laser clad chromium carbide-NiCrMoNb

    International Nuclear Information System (INIS)

    Venkatesh, L.; Samajdar, I.; Tak, Manish; Doherty, Roger D.; Gundakaram, Ravi C.; Prasad, K. Satya; Joshi, S.V.

    2015-01-01

    Highlights: • Microstructural development during laser cladding has been studied. • In this multi component system Cr 7 C 3 is found to be the stable carbide phase. • Phases were identified by EBSD since XRD results were not conclusive. • Increase in laser power and/or scanning speed reduced the carbide content. • Hardness seems to depend on phase content as well as microstructure. - Abstract: Microstructural development in laser clad layers of Chromium carbide (Cr x C y )-NiCrMoNb on SA 516 steel has been investigated. Although the starting powder contained both Cr 3 C 2 and Cr 7 C 3 , the clad layers showed only the presence of Cr 7 C 3 . Microtexture measurements by electron back scattered diffraction (EBSD) revealed primary dendritic Cr 7 C 3 with Ni rich FCC metallic phase being present in the interdendritic spaces. Further annealing of the laser clad layers and furnace melting of the starting powder confirmed that Cr 7 C 3 is the primary as well as stable carbide phase in this multi component system. Increase in laser power and scanning speed progressively reduced carbide content in the laser clad layers. Increased scanning speed, which enhances the cooling rate, also led to reduction in the secondary arm spacing (λ 2 ) of the Cr 7 C 3 dendrites. The clad layer hardness increased with carbide content and with decreased dendrite arm spacing.

  19. Evaluation of thermocouple fin effect in cladding surface temperature measurement during film boiling

    International Nuclear Information System (INIS)

    Tsuruta, Takaharu; Fujishiro, Toshio

    1984-01-01

    Thermocouple fin effect on surface temperature measurement of a fuel rod has been studied at elevated wall temperatures under film boiling condition in a reactivity initiated accident (RIA) situation. This paper presents an analytical equation to evaluate temperature drops caused by the thermocouple wires attached to cladding surface. The equation yielded the local temperature drop at measuring point depending on thermocouple diameter, cladding temperature, coolant flow condition and vapor film thickness. The temperature drops by the evaluating equation were shown in cases of free and forced convection conditions. The analytical results were compared with the measured data for various thermocouple sizes, and also with the estimated maximum cladding temperature based on the oxidation layer thickness in the cladding outer surface. It was concluded that the temperature drops at above 1,000 0 C in cladding temperature were around 120 and 150 0 C for 0.2 and 0.3 mm diameter Pt-Pt.Rh thermocouples, respectively, under a stagnant coolant condition. The fin effect increases with the decrease of vapor film thickness such as under forced flow cooling or at near the quenching point. (author)

  20. 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